Behavior of mixed-oxide fuel elements during an overpower transient

International Nuclear Information System (INIS)

Tsai, H.; Shikakura, S.

1993-01-01

A slow-ramp (0.1%/s), extended overpower (∼90%) transient test was conducted in EBR-II on 19 mixed-oxide fuel elements with conservative, moderate, and aggressive designs. Claddings for the elements were Type 316, D9, or PNC-316 stainless steel. Before the transient, the elements were preirradiated under steady-state or steady-state plus duty-cycle (periodic 15% overpower transient) conditions to burnups of 2.5-9.7 at%. Cladding integrity during the transient test was maintained by all fuel elements except one, which had experienced substantial overtemperature in the earlier stedy-state irradiation. Extensive centerline fuel melting occurred in all test elements. Significantly, this melting did not cause any elements to breach, although it did have a strong effect on the other aspects of fuel element behavior. (orig.)

Mixed waste paper to ethanol fuel

Energy Technology Data Exchange (ETDEWEB)

1991-01-01

The objectives of this study were to evaluate the use of mixed waste paper for the production of ethanol fuels and to review the available conversion technologies, and assess developmental status, current and future cost of production and economics, and the market potential. This report is based on the results of literature reviews, telephone conversations, and interviews. Mixed waste paper samples from residential and commercial recycling programs and pulp mill sludge provided by Weyerhauser were analyzed to determine the potential ethanol yields. The markets for ethanol fuel and the economics of converting paper into ethanol were investigated.

Fuel assembly for FBR type reactor and reactor core thereof

International Nuclear Information System (INIS)

Kobayashi, Kaoru.

1998-01-01

The present invention provides a fuel assembly to be loaded to a reactor core of a large sized FBR type reactor, in which a coolant density coefficient can be reduced without causing power peaking in the peripheral region of neutron moderators loaded in the reactor core. Namely, the fuel assembly for the FBR type reactor comprises a plurality of fission product-loaded fuel rods and a plurality of fertile material-loaded fuel rods and one or more rods loading neutron moderators. In this case, the plurality of fertile material-loaded fuel rods are disposed to the peripheral region of the neutron moderator-loaded rods. The plurality of fission product-loaded fuel rods are disposed surrounding the peripheral region of the plurality of fertile material-loaded fuel rods. The neutron moderator comprises zirconium hydride, yttrium hydride and calcium hydride. The fission products are mixed oxide fuels. The fertile material comprises depleted uranium or natural uranium. (I.S.)

HETEROGENEOUS REBURNING BY MIXED FUELS

Energy Technology Data Exchange (ETDEWEB)

Wei-Yin Chen; Benson B. Gathitu

2005-01-14

Recent studies of heterogeneous reburning, i.e., reburning involving a coal-derived char, have elucidated its variables, kinetics and mechanisms that are valuable to the development of a highly efficient reburning process. Young lignite chars contain catalysts that not only reduce NO, but they also reduce HCN that is an important intermediate that recycles to NO in the burnout zone. Gaseous CO scavenges the surface oxides that are formed during NO reduction, regenerating the active sites on the char surface. Based on this mechanistic information, cost-effective mixed fuels containing these multiple features has been designed and tested in a simulated reburning apparatus. Remarkably high reduction of NO and HCN has been observed and it is anticipated that mixed fuel will remove 85% of NO in a three-stage reburning process.

Mixing enhancement in a scramjet combustor using fuel jet injection swirl

Science.gov (United States)

Flesberg, Sonja M.

The scramjet engine has proven to be a viable means of powering a hypersonic vehicle, especially after successful flights of the X-51 WaveRider and various Hy-SHOT test vehicles. The major challenge associated with operating a scramjet engine is the short residence time of the fuel and oxidizer in the combustor. The fuel and oxidizer have only milliseconds to mix, ignite and combust in the combustion chamber. Combustion cannot occur until the fuel and oxidizer are mixed on a molecular level. Therefore the improvement of mixing is of utmost interest since this can increase combustion efficiency. This study investigated mixing enhancement of fuel and oxidizer within the combustion chamber of a scramjet by introducing swirl to the fuel jet. The investigation was accomplished with numerical simulations using STAR-CCM+ computational fluid dynamic software. The geometry of the University of Virginia Supersonic Combustion Facility was used to model the isolator, combustor and nozzle of a scramjet engine for simulation purposes. Experimental data from previous research at the facility was used to verify the simulation model before investigating the effect of fuel jet swirl on mixing. The model used coaxial fuel jet with a swirling annular jet. Single coaxial fuel jet and dual coaxial fuel jet configurations were simulated for the investigation. The coaxial fuel jets were modelled with a swirling annular jet and non-swirling core jet. Numerical analysis showed that fuel jet swirl not only increased mixing and entrainment of the fuel with the oxidizer but the mixing occurred further upstream than without fuel jet swirl. The burning efficiency was calculated for the all the configurations. An increase in burning efficiency indicated an increase in the mixing of H2 with O2. In the case of the single fuel jet models, the maximum burning efficiency increase due to fuel injection jet swirl was 23.3%. The research also investigated the possibility that interaction between two

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

International Nuclear Information System (INIS)

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

1983-01-01

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

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

Fuel type characterization based on coarse resolution MODIS satellite data

Directory of Open Access Journals (Sweden)

Lanorte A

2007-01-01

Full Text Available Fuel types is one of the most important factors that should be taken into consideration for computing spatial fire hazard and risk and simulating fire growth and intensity across a landscape. In the present study, forest fuel mapping is considered from a remote sensing perspective. The purpose is to delineate forest types by exploring the use of coarse resolution satellite remote sensing MODIS imagery. In order to ascertain how well MODIS data can provide an exhaustive classification of fuel properties a sample area characterized by mixed vegetation covers and complex topography was analysed. The study area is located in the South of Italy. Fieldwork fuel type recognitions, performed before, after and during the acquisition of remote sensing MODIS data, were used as ground-truth dataset to assess the obtained results. The method comprised the following three steps: (I adaptation of Prometheus fuel types for obtaining a standardization system useful for remotely sensed classification of fuel types and properties in the considered Mediterranean ecosystems; (II model construction for the spectral characterization and mapping of fuel types based on two different approach, maximum likelihood (ML classification algorithm and spectral Mixture Analysis (MTMF; (III accuracy assessment for the performance evaluation based on the comparison of MODIS-based results with ground-truth. Results from our analyses showed that the use of remotely sensed MODIS data provided a valuable characterization and mapping of fuel types being that the achieved classification accuracy was higher than 73% for ML classifier and higher than 83% for MTMF.

Microbial Fuel Cells using Mixed Cultures of Wastewater for Electricity Generation

International Nuclear Information System (INIS)

Zain, S.M; Roslani, N.S.; Hashim, R.; Anuar, N.; Suja, F.; Basi, N.E.A.; Anuar, N.; Daud, W.R.W.

2011-01-01

Fossil fuels (petroleum, natural gas and coal) are the main resources for generating electricity. However, they have been major contributors to environmental problems. One potential alternative to explore is the use of microbial fuel cells (MFCs), which generate electricity using microorganisms. MFCs uses catalytic reactions activated by microorganisms to convert energy preserved in the chemical bonds between organic molecules into electrical energy. MFC has the ability to generate electricity during the wastewater treatment process while simultaneously treating the pollutants. This study investigated the potential of using different types of mixed cultures (raw sewage, mixed liquor from the aeration tank and return waste activated sludge) from an activated sludge treatment plant in MFCs for electricity generation and pollutant removals (COD and total kjeldahl nitrogen, TKN). The MFC in this study was designed as a dual-chambered system, in which the chambers were separated by a Nafion TM membrane using a mixed culture of wastewater as a bio catalyst. The maximum power density generated using activated sludge was 9.053 mW/ cm 2 , with 26.8 % COD removal and 40 % TKN removal. It is demonstrated that MFC offers great potential to optimize power generation using mixed cultures of wastewater. (author)

Simulated physical inventory verification exercise at a mixed-oxide fuel fabrication facility

International Nuclear Information System (INIS)

Reilly, D.; Augustson, R.

1985-01-01

A physical inventory verification (PIV) was simulated at a mixed-oxide fuel fabrication facility. Safeguards inspectors from the International Atomic Energy Agency (IAEA) conducted the PIV exercise to test inspection procedures under ''realistic but relaxed'' conditions. Nondestructive assay instrumentation was used to verify the plutonium content of samples covering the range of material types from input powders to final fuel assemblies. This paper describes the activities included in the exercise and discusses the results obtained. 5 refs., 1 fig., 6 tabs

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

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

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

Research on calculation of mixing fraction for natural uranium equivalent fuel

International Nuclear Information System (INIS)

Huang Shien; Wang Lianjie; Wei Yanqin; Li Qing; Zheng Jiye

2013-01-01

Based on the first-order perturbation theory and reasonable approximations, the calculation method of recycled uranium (RU) and depleted uranium (DU) mixing fraction for natural uranium equivalent (NUE) fuel was studied, so the equivalence between NUE fuel and natural uranium (NU) fuel was assured. The adopted calculation method accurately takes the variation of micro cross sections alone with fuel depletion into account. A computer code named ALPHA was programmed to execute the calculation procedure. Then the ALPHA code and the WIMS-AECL code compose a processing system, which is applicable to the mixing fraction calculation for heavy water reactor NUE fuel. The validation shows that the processing system can accurately calculate the mixing fraction for NUE fuel. (authors)

Ashes from biofuels and mixed fuels - amount and qualities

International Nuclear Information System (INIS)

Bjurstroem, Henrik; Ilskog, Elisabeth; Berg, Magnus

2003-04-01

In this study, ashes from biofuels used in the energy utilities, the pulp and paper industry and the wood-working industries have been inventoried. The selection of plants to which enquiries were addressed consists of about 50 utilities, all pulp and paper plants and about 20 wood-working industries (e.g. sawmills). The purpose of the study was to estimate the quantities of bio ashes that are recycled to the forests and those that could be recycled. The background to this study is that logging slash is harvested from ca 30,000 ha per year, while ash is recycled only to 2 to 4,000 ha per year. A working hypothesis has been that logging slash or clean wooden fuels are mixed with other fuels to such an extent that the ash is too contaminated to be recycled. The consequence would be that there is a shortage of suitable ash. Therefore, it was desirable that motives for mixing fuels be chartered. In Sweden, approximately one million ton ashes are produced each year and the share of the three industries that have been studied is estimated as: 200 - 340,000 tons from utilities about 275,000 tons from the pulp and paper industry and 100,000 tons from the woodworking industry. These quantities include unburned carbon, water added when the ash is extracted from the boilers etc. Additional quantities of ash are those produced by waste combustion (447,000 tons), wood-burning in residential buildings (50 - 100,000 tons) etc. In all, ash that may be recycled should total about 300,000 tons (Recyclable ash in t/a: Utilities - 80,000; Pulp and Paper Industry - 100-130,000; Woodworking Industry 100,000). Logging slash is seldom burned alone in the boilers at the utilities, but are almost always mixed with other wood fuel fractions such as waste from sawmills. The mixtures can be very complex. Clean mixtures of wood fuel fractions represent ca 4,500 GWh of the ca 7,800 GWh in this study. Other fuels that are often used in mixtures are peat and Salix, which does not necessarily lead

Porous fuel air mixing enhancing nozzle (PFAMEN)

NARCIS (Netherlands)

Reijnders, J.J.E.; Boot, M.D.; Luijten, C.C.M.; Frijters, P.J.M.; Goey, de L.P.H.

2009-01-01

One of the challenges with conventional diesel engines is the emission of soot. To reduce soot emission whilst maintaining fuel efficiency, an important pathway is to improve the fuel-air mixing process. This can be achieved by creating small droplets in order to enhance evaporation. Furthermore,

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

Fabrication and characterization of MX-type fuels and fuel pins

International Nuclear Information System (INIS)

Richter, K.; Bartscher, W.; Benedict, U.; Gueugnon, J.F.; Kutter, H.; Sari, C.; Schmidt, H.E.

1978-01-01

This paper summarizes the most important fabrication parameters and characterization of fuel and fuel pins obtained during the investigation of uranium-plutonium carbides, oxicarbides, carbonitrides and nitrides in the past years at the European Institute for Transuranium Elements at Karlsruhe. All preparation methods discussed are based on carbothermic reduction of a mechanical blend of uranium-plutonium oxide and carbon powder. General data for carbothermic reduction processes are discussed (influence of starting material, homogeneity, control of degree of reaction, etc). A survey of different preparation methods investigated is given. Limitations with respect to temperature and atmosphere for both carbothermic reduction processes and sintering conditions for the different compounds are summarized. A special preparation process for mixed carbonitrides with low nitrogen content (U,Pu)sub(1-x)Nsub(x) in the range 0.1 0 C to 1400 0 C by means of a modulated electron beam technique. A scheme is proposed, which allows to predict the thermal properties of MX fuels on the basis of their chemical composition and porosity. Preparation, preirradiation characterization and final controls of fuel test pins for pellet and vibrocompacted type of pins are described and the most important data summarized for all advanced fuels irradiated at Dounreay (DN1) and Rapsodie Fast Reactor (DN2) within the TU irradiation programme

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

The influence of droplet evaporation on fuel-air mixing rate in a burner

Science.gov (United States)

Komiyama, K.; Flagan, R. C.; Heywood, J. B.

1977-01-01

Experiments involving combustion of a variety of hydrocarbon fuels in a simple atmospheric pressure burner were used to evaluate the role of droplet evaporation in the fuel/air mixing process in liquid fuel spray flames. Both air-assist atomization and pressure atomization processes were studied; fuel/air mixing rates were determined on the basis of cross-section average oxygen concentrations for stoichiometric overall operation. In general, it is concluded that droplets act as point sources of fuel vapor until evaporation, when the fuel jet length scale may become important in determining nonuniformities of the fuel vapor concentration. In addition, air-assist atomizers are found to have short droplet evaporation times with respect to the duration of the fuel/air mixing process, while for the pressure jet atomizer the characteristic evaporation and mixing times are similar.

Powering Profits. Profits, Investments and Fuel Type Mixes in the Dutch Power Sector

International Nuclear Information System (INIS)

Wilde-Ramsing, J.; Steinweg, T.

2007-06-01

This report addresses the Dutch power sector, identifying the major corporate players in the market, types of fuel used to generate electricity, the profits being made, and investments in both renewable and non-renewable generation capacity. For the purposes of this report, the power sector is understood to encompass production (i.e. generation) and supply of electricity. Some discussion and figures on heat and gas, which are also essential energy services, are provided, but the focus is primarily on electricity. Section 2 of the report provides an overview of the Dutch power sector, breaking the market down into production and supply. Major players, markets shares, and recent trends and developments are given for each of these activities. Sections 3 - 7 go into detail on the five major corporate players active in the Dutch power sector: ENECO, Essent, Nuon, Electrabel, and E.ON Energie. For each company, information is provided on profits and earnings, the fuel mix used to generate and supply electricity, the CO2 emissions associated with these activities, installed capacity in the Netherlands, and recent investments in renewable and non-renewable generation capacity in the Netherlands. For the Dutch companies, ENECO, Essent and Nuon, additional information on the ownership structure of the company, shareholders and dividends paid and received is given. A section on RWE (Section 8) is also included in the study because, although RWE is not currently active in generating electricity in the Netherlands, RWE Energy does currently supply electricity generated by producers in the Netherlands. In addition, RWE Power is currently planning to invest significantly in power generation capacity in the Netherlands. The final section of the report compares the companies activities in the Netherlands and draws conclusions based on the companies' respective performance

Thermal aspects of mixed oxide fuel in application to supercritical water-cooled nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Grande, L.; Peiman, W.; Rodriguez-Prado, A.; Villamere, B.; Mikhael, S.; Allison, L.; Pioro, I., E-mail: lisa.grande@mycampus.uoit.ca, E-mail: igor.pioro@uoit.ca [Univ. of Ontario Inst. of Tech., Faculty of Energy Systems and Nuclear Science, Oshawa, Ontario (Canada)

2010-07-01

SuperCritical Water-cooled nuclear Reactors (SCWRs) are a renewed technology being developed as one of the Generation IV reactor concepts. This reactor type uses a light water coolant at temperatures and pressures above its critical point. These elevated operating conditions will improve Nuclear Power Plant (NPP) thermal efficiencies by 10 - 15% compared to those of current NPPs. Also, SCWRs will have the ability to utilize a direct cycle, thus decreasing NPP capital and operational costs. The SCWR core has 2 configurations: 1) Pressure Vessel (PV) -type enclosing a fuel assembly and 2) Pressure Tube (PT) -type consisting of individual pressurized channels containing fuel bundles. Canada and Russia are developing PT-type SCWRs. In particular, the Canadian SCWR reactor has an output of 1200 MW{sub el} and will operate at a pressure of 25 MPa with inlet and outlet fuel-channel temperatures of 350 and 625°C, respectively. These extreme operating conditions require alternative fuels and materials to be investigated. Current CANadian Deuterium Uranium (CANDU) nuclear reactor fuel-channel design is based on the use of uranium dioxide (UO{sub 2}) fuel; zirconium alloy sheath (clad) bundle, pressure and calandria tubes. Alternative fuels should be considered to supplement depleting world uranium reserves. This paper studies general thermal aspects of using Mixed OXide (MOX) fuel in an Inconel-600 sheath in a generic PT-type SCWR. The bulk fluid, sheath and fuel centerline temperatures along with the Heat Transfer Coefficient (HTC) profiles were calculated at uniform and non-uniform Axial Heat Flux Profiles (AHFPs). (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

Performance simulation of planar SOFC using mixed hydrogen and carbon monoxide gases as fuel

Energy Technology Data Exchange (ETDEWEB)

Inui, Y. [Department of Electrical and Electronic Engineering, Toyohashi University of Technology, Tempaku-cho, Toyohashi 441-8580 (Japan)]. E-mail: inui@eee.tut.ac.jp; Urata, A. [Department of Electrical and Electronic Engineering, Toyohashi University of Technology, Tempaku-cho, Toyohashi 441-8580 (Japan); Ito, N. [Department of Electrical and Electronic Engineering, Toyohashi University of Technology, Tempaku-cho, Toyohashi 441-8580 (Japan); Nakajima, T. [Department of Electrical and Electronic Engineering, Toyohashi University of Technology, Tempaku-cho, Toyohashi 441-8580 (Japan); Tanaka, T. [Department of Electrical and Electronic Engineering, Toyohashi University of Technology, Tempaku-cho, Toyohashi 441-8580 (Japan)

2006-08-15

The authors investigate in detail the influence of the mixing ratio of hydrogen and carbon monoxide in the fuel on the cell performance of the SOFC through numerical simulations for a single cell plate of the co-flow type planar cell. It is made clear that the cell performance is almost the same and excellent, independent of the mixing ratio of hydrogen and carbon monoxide under the nominal operating condition. The electromotive force of the hydrogen rich fuel gas is a little higher than that of the carbon monoxide rich fuel gas. The internal voltage drop in the cell decreases as the fraction of carbon monoxide becomes high. Since the value of the single cell voltage is determined by the balance of these two phenomena, the lowering of the electromotive force is dominant and the single cell voltage of the hydrogen rich fuel gas is higher when the inlet gas temperature is high, whereas the voltage drop reduction is dominant and the single cell voltage of the carbon monoxide rich fuel gas is higher when the temperature is low. The effect of the additional gases of water vapor and carbon dioxide is restricted to the single cell voltage shift, and the qualitative dependence of the single cell voltage on the inlet gas temperature is determined by the mixing ratio of hydrogen and carbon monoxide.

Performance simulation of planar SOFC using mixed hydrogen and carbon monoxide gases as fuel

International Nuclear Information System (INIS)

Inui, Y.; Urata, A.; Ito, N.; Nakajima, T.; Tanaka, T.

2006-01-01

The authors investigate in detail the influence of the mixing ratio of hydrogen and carbon monoxide in the fuel on the cell performance of the SOFC through numerical simulations for a single cell plate of the co-flow type planar cell. It is made clear that the cell performance is almost the same and excellent, independent of the mixing ratio of hydrogen and carbon monoxide under the nominal operating condition. The electromotive force of the hydrogen rich fuel gas is a little higher than that of the carbon monoxide rich fuel gas. The internal voltage drop in the cell decreases as the fraction of carbon monoxide becomes high. Since the value of the single cell voltage is determined by the balance of these two phenomena, the lowering of the electromotive force is dominant and the single cell voltage of the hydrogen rich fuel gas is higher when the inlet gas temperature is high, whereas the voltage drop reduction is dominant and the single cell voltage of the carbon monoxide rich fuel gas is higher when the temperature is low. The effect of the additional gases of water vapor and carbon dioxide is restricted to the single cell voltage shift, and the qualitative dependence of the single cell voltage on the inlet gas temperature is determined by the mixing ratio of hydrogen and carbon monoxide

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)

Biological behavior of mixed LMFBR-fuel-sodium aerosols

International Nuclear Information System (INIS)

Mahlum, D.D.; Hackett, P.L.; Hess, J.O.; Allen, M.D.

1979-01-01

Immediately after exposure of rats to mixed aerosols of sodium-LMFBR fuel, about 80 to 90% of the body burden of 239 Pu is in the gastrointestinal tract; 1.5 to 4% is in the lungs. With fuel-only aerosols, less of the body burden was in the GI tract and more in the lung and the head. Blood and urine values suggest an increased absorption of 239 Pu from sodium-fuel than from fuel-only aerosols

Used mixed oxide fuel reprocessing at RT-1 plant

Energy Technology Data Exchange (ETDEWEB)

Kolupaev, D.; Logunov, M.; Mashkin, A.; Bugrov, K.; Korchenkin, K. [FSUE PA ' Mayak' , 30, Lenins str, Ozersk, 460065 (Russian Federation); Shadrin, A.; Dvoeglazov, K. [ITCP ' PRORYV' , 2/8 Malaya Krasmoselskay str, Moscow, 107140 (Russian Federation)

2016-07-01

Reprocessing of the mixed uranium-plutonium spent nuclear fuel of the BN-600 reactor was performed at the RT-1 plant twice, in 2012 and 2014. In total, 8 fuel assemblies with a burn-up from 73 to 89 GW day/t and the cooling time from 17 to 21 years were reprocessed. The reprocessing included the stages of dissolution, clarification, extraction separation of U and Pu with purification from the fission products, refining of uranium and plutonium at the relevant refining cycles. Dissolution of the fuel composition of MOX used nuclear fuel (UNF) in nitric acid solutions in the presence of fluoride ion has occurred with the full transfer of actinides into solution. Due to the high content of Pu extraction separation of U and Pu was carried out on a nuclear-safe equipment designed for the reprocessing of highly enriched U spent nuclear fuel and Pu refining. Technological processes of extraction, separation and refining of actinides proceeded without deviations from the normal mode. The output flow of the extraction outlets in their compositions corresponded to the regulatory norms and remained at the level of the compositions of the streams resulting from the reprocessing of fuel types typical for the RT-1 plant. No increased losses of Pu into waste have been registered during the reprocessing of BN-600 MOX UNF an compare with VVER-440 uranium UNF reprocessing. (authors)

Fuel and nuclear fuel cycle

International Nuclear Information System (INIS)

Prunier, C.

1998-01-01

The nuclear fuel is studied in detail, the best choice and why in relation with the type of reactor, the properties of the fuel cans, the choice of fuel materials. An important part is granted to the fuel assembly of PWR type reactor and the performances of nuclear fuels are tackled. The different subjects for research and development are discussed and this article ends with the particular situation of mixed oxide fuels ( materials, behavior, efficiency). (N.C.)

Intermediate flow mixing nonsupport grid for BWR fuel assembly

International Nuclear Information System (INIS)

Taleyarkhan, R.P.

1987-01-01

An intermediate flow mixing nonsupport grid is described for use in a nuclear reactor fuel assembly containing an array of elongated fuel rods. The grid comprises: (a) interleaved inner straps arranged in an egg-crate configuration to define inner cell openings for receiving respective ones of the fuel rods. The inner straps have outer terminal end portions; (b) an outer peripheral strap attached to the respective terminal end portions of the inner straps to define perimeter cell openings for receiving other ones of the fuel rods. The inner straps and outer strap together have opposite upstream and downstream sides; (c) a first group of mixing vanes disposed at the downstream side and being attached on portions of the outer strap and on respective portions of the inner straps. Together with the outer strap portions, they define the perimeter cell openings. Each of the mixing vanes of the first group extend generally in a downstream direction and inwardly toward the perimeter cell openings for deflecting coolant flowing; and (d) a second group of mixing vanes disposed at the downstream side and being attached on other portions of the inner straps. Together with the respective portions, they define the inner cell openings. Each of the mixing vanes of the second group extend generally in a downstream direction and inwardly toward the inner cell openings for deflecting coolant flowing therethrough; (e) the mixing vanes of the second group are substantially smaller in size than the mixing vanes of the first group so as to generate substantially less turbulence in the portions of the coolant flowing through the inner cell openings than in the portions of the coolant flowing through the perimeter cell openings

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

Proposed model for fuel-coolant mixing during a core-melt accident

International Nuclear Information System (INIS)

Corradini, M.L.

1983-01-01

If complete failure of normal and emergency coolant flow occurs in a light water reactor, fission product decay heat would eventually cause melting of the reactor fuel and cladding. The core melt may then slump into the lower plenum and later into the reactor cavity and contact residual liquid water. A model is proposed to describe the fuel-coolant mixing process upon contact. The model is compared to intermediate scale experiments being conducted at Sandia. The modelling of this mixing process will aid in understanding three important processes: (1) fuel debris sizes upon quenching in water, (2) the hydrogen source term during fuel quench, and (3) the rate of steam production. Additional observations of Sandia data indicate that the steam explosion is affected by this mixing process

Survey of post-irradiation examinations made of mixed carbide fuels

International Nuclear Information System (INIS)

Coquerelle, M.

1997-01-01

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

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

Mixed core conversion study with HEU and LEU fuels

International Nuclear Information System (INIS)

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

1984-01-01

The results of a mixed core study are presented for gradual replacement of HEU fuel with LEU fuel using the IAEA generic 10 MW reactor as an example. The key parameters show that the transition can be accomplished safely and economically

Emission testing of jatropha and pongamia mixed bio diesel fuel in a diesel engine

International Nuclear Information System (INIS)

Ali, M.; Shaikh, A.A.

2012-01-01

The present investigation is based on the emission characteristics of mixed bio diesel fuel in a four stroke single cylinder compression ignition engine at constant speed. Refined oils of jatropha and pongamia are converted into bio diesel by acid catalyzed esterification and base catalyzed transesterification reactions. The jatropha and pongamia bio diesel were mixed in equal proportions with conventional mineral diesel fuel. Four samples of fuel were tested namely, diesel fuel, B10, B20 and B40. The emission analysis showed B20 mixed bio diesel fuel blend having better results as compared to other samples. There is 60% and 35% lower emission of carbon monoxide and in sulphur dioxide observed while consuming B20 blended fuel respectively. The test result showed NOx emissions were 10% higher from bio diesel fuel, as compared to conventional diesel fuel. However, these emissions may be reduced by EGR (Exhaust Gas Recirculation) technology. Present research also revealed that that B20 mixed bio diesel fuel can be used, without any modification in a CI engine. (author)

Nondestructive characterization of mixed oxide pellets in welded nuclear fuel pins by neutron radiography and gamma-autoradiography

International Nuclear Information System (INIS)

Panakkal, J.P.; Ghosh, J.K.; Roy, P.R.

1989-01-01

Nondestructive evaluation of nuclear fuel pellets after the welding of fuel pins plays a vital role in assuring a safe and reliable operation of reactors. Some of the important characteristics to be monitored in low plutonium enriched mixed oxide fuel pellets are plutonium enrichment, size of plutonium dioxide agglomerates, incorrect loading and geometric shape. Experiments were carried out at Bhabha Atomic Research Centre, Bombay on experimental fuel pins containing mixed oxide pellets of different geometry (solid and annular), of different plutonium enrichment (0-6 w% of plutonium dioxide) and containing PuO 2 agglomerates of size 125-2000 microns to evaluate these characteristics nondestructively. Neutron radiography of these fuel pins was carried out using a swimming pool type reactor 'APSARA'. Results of quantitative evaluation of the neutron radiographs and a simple model correlating neutron interaction probability and the optical density are presented. Gamma autoradiography of these fuel pins showed that these parameters could be evaluated with a few limitations. This paper presents the experimental details, quantitative analysis of the radiographs by microdensitometry and merits and demerits of neutron radiography and gamma autoradiography for nondestructive charcterisation of nuclear fuel pellets. (orig.)

Prices versus policy: An analysis of the drivers of the primary fossil fuel mix

International Nuclear Information System (INIS)

Atalla, Tarek; Blazquez, Jorge; Hunt, Lester C.; Manzano, Baltasar

2017-01-01

Energy policymakers often attempt to shape their countries' energy mix, rather than leave it purely to market forces. By calibrating and simulating a Dynamic Stochastic General Equilibrium (DSGE) model, this paper analyzes the primary fossil fuel mix in the USA and compares it to Germany and the UK, given the different evolution of the mixes and the different roles played by relative prices and policy in North America and Europe. It is found that the model explains well the evolution of the primary fossil fuel mix in the USA for the period 1980–2014, suggesting that relative fossil fuel prices generally dominated in determining the mix during this time. However, this is not the case for Germany and the UK. For both countries, the model performs well only for the period after the market-oriented reforms in the 1990s. Additionally, the volatility of private consumption and output for the pre- and post-reform periods is evaluated for Germany and the UK and it is found that the liberalized energy markets brought about a transition from coal to natural gas, but with increased macroeconomic volatility. - Highlights: • Macroeconomic analysis of the importance of prices vs policy in driving the primary fossil fuel mix. • USA primary fossil fuel mix chiefly driven by relative prices since the early 1980s. • Germany and UK primary fossil fuel mix chiefly driven by policy until 1990s. • Germany and UK primary fossil fuel mix chiefly driven by relative prices since early to mid-1990s. • Transition from coal to natural gas in Germany and UK increased macroeconomic volatility.

The optimal fuel mix and redistribution of social surplus in the Korean power market

International Nuclear Information System (INIS)

Kim, Hyunsook; Kim, Sung-Soo

2010-01-01

This paper investigates the difference between the optimal fuel mix incorporating a pre-installed generation capacity constraint and the actual fuel mix in the Korean power market. Since the restructuring of the market, the fuel mix has been determined partly by investors and partly by the Basic Plan for Long-Term Electricity Supply and Demand (BPE). Both the system marginal price (SMP), and the capacity payment (CP), which has been based on the fixed cost of a specific gas turbine generator, were intended to provide an investment incentive in the market; however, they did not bring about an optimal fuel mix in Korea. Under the circumstances of a shortage of base load generators, these generators may garner excessive profits due to a high SMP level. However, the adjustment scheme of profit between KEPCO and Gencos left scant profit for generators. This paper suggests that a contract is needed to create the appropriate profit and tax levels for these base load generators. The redistribution of profit improves equality between consumers and generators, and the proper margin creates incentives for base load technology investment in Korea. - Research highlights: → This paper aims to determine the optimal fuel mix in Korea and shows the difference between the optimal and actual fuel mix; → We discovered that the optimal fuel mix ratio should be 63.5%:20.5%:16.0%for nuclear energy, coal, and LNG, respectively; → Because the pre-installed capacity is taken as given, the optimal fuel mix under the given installed capacity is an addition of a new nuclear unit to the current fuel mix; → The additional profit above the profit margin in the BPE should be collected and redistributed to consumers; → While we provide an incentive to build nuclear and coal units in Korea, we also suggest that a contract is needed to guarantee the profit level for generators based on a government regulation constraint, the BPE, to achieve a fair treatment.

Mixed Reload Design Using MOX and UOX Fuel Assemblies

International Nuclear Information System (INIS)

Ramon, Ramirez Sanchez J.; Perry, R.T.

2002-01-01

As part of the studies involved in plutonium utilization assessment for a Boiling Water Reactor, a conceptual design of MOX fuel was developed, this design is mechanically the same design of 10 X 10 BWR fuel assemblies but different fissile material. Several plutonium and gadolinium concentrations were tested to match the 18 months cycle length which is the current cycle length of LVNPP, a reference UO 2 assembly was modeled to have a full cycle length to compare results, an effective value of 0.97 for the multiplication factor was set as target for 470 Effective Full Power days for both cycles, here the gadolinium concentration was a key to find an average fissile plutonium content of 6.55% in the assembly. A reload of 124 fuel assemblies was assumed to simulate the complete core, several load fractions of MOX fuel mixed with UO 2 fresh fuel were tested to verify the shutdown margin, the UO 2 fuel meets the shutdown margin when 124 fuel assemblies are loaded into the core, but it does not happen when those 124 assemblies are replaced with MOX fuel assemblies, so the fraction of MOX was reduced step by step up to find a mixed load that meets both length cycle and shutdown margin. Finally the conclusion is that control rods losses some of their worth in presence of plutonium due to a more hardened neutron spectrum in MOX fuel and this fact limits the load of MOX fuel assemblies in the core, this results are shown in this paper. (authors)

Gas cooled fast breeder reactors using mixed carbide fuel

International Nuclear Information System (INIS)

Kypreos, S.

1976-09-01

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

The effects of alternative fuel types on the organoleptic qualities of ...

African Journals Online (AJOL)

This study was carried out to assess the effects of alternative smoking fuel types on the organoleptic qualities of coarse pork sausages. The sausages were produced with lean pork (2.5 kg) and pork fat (0.5 kg), minced, mixed with spices and stuffed into natural casings. They were grouped into four and each group was ...

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

Conversion of Mixed Plastic Wastes (High Density Polyethylene and Polypropylene) into Liquid Fuel

International Nuclear Information System (INIS)

Chaw Su Su Hmwe; Tint Tint Kywe; Moe Moe Kyaw

2010-12-01

In this study, mixed plastic wastes were converted into liquid fuels. Mixed plastic wastes used were high density polyethylene (HDPE) and polypropylene (PP). The pyrolysis of mixed plastic waste to liquid fuel was carried out with and without prepared zeolite catalyst.The catalyst was characterized by X-ray Diffraction (XRD). This catalyst was pre-treated for activation. The experiments were carried out at temperature range of 350-410C.Physical properties (density, kinematic, viscosity,refractive index)of prepared liquid fuel samples were measured. From this study, yields of liquid fuel and gas fuel were found to be 41-64% and 15-35% respectively. As for by products, char was obtained as the yield percentages from 9 to 14% and wax (yield% - 1 to 14) was formed during pyrolysis.

Detailed analysis of coolant mixing in WWER-440 fuel assembly heads

International Nuclear Information System (INIS)

Toth, S.; Aszodi, A.

2008-01-01

Based on experiences of former validation and sensitivity studies, a CFD model for head part of real working fuel assemblies with pitch of 12.3 mm has been developed with the code ANSYS CFX. Calculations were performed for typical fuel assemblies used in the Paks NPP. Differences between the outlet average temperatures and thermocouple signals were determined. Effect of the mixing grid position on thermocouple signal was investigated also. Mixing was analyzed in details with using so-called mixing scalars and weight factors of the central tube and rod bundle regions for in-core thermocouple signal were determined. Sensitivity of the weight factors for pin power distribution and mixing grid position were investigated. (Authors)

Passive measurements of mixed-oxide fuel for nuclear nonproliferation

International Nuclear Information System (INIS)

Dolan, Jennifer L.; Flaska, Marek; Pozzi, Sara A.; Chichester, David L.

2013-01-01

We present new results on passive measurements and simulations of mixed-oxide fuel-pin assemblies. Potential tools for mixed-oxide fuel pin characterization are discussed for future nuclear-nonproliferation applications. Four EJ-309 liquid scintillation detectors coupled with an accurate pulse timing and digital, offline and optimized pulse-shape discrimination method were used. Measurement analysis included pulse-height distributions to distinguish between purely fission neutron sources and alpha-n plus fission neutrons sources. Time-dependent cross-correlation functions were analyzed to measure the fission neutron contribution to the measured sample's neutron source. The use of Monte Carlo particle transport code MCNPX-PoliMi is discussed in conjunction with the measurements

A mixed core conversion study with HEU and LEU fuels

International Nuclear Information System (INIS)

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

1985-01-01

The results of a mixed core study are presented for gradual replacement of HEU fuel with LEU fuel using the IAEA generic 10 MW reactor as an example. The key parameters show that the transition can be accomplished safely and economically. (author)

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.)

Determination of fissile fraction in MOX (mixed U + Pu oxides) fuels for different burnup values

International Nuclear Information System (INIS)

Ozdemir, Levent; Acar, Banu Bulut; Zabunoglu, Okan H.

2011-01-01

When spent Light Water Reactor fuels are processed by the standard Purex method of reprocessing, plutonium (Pu) and uranium (U) in spent fuel are obtained as pure and separate streams. The recovered Pu has a fissile content (consisting of 239 Pu and 241 Pu) greater than 60% typically (although it mainly depends on discharge burnup of spent fuel). The recovered Pu can be recycled as mixed-oxide (MOX) fuel after being blended with a fertile U makeup in a MOX fabrication plant. The burnup that can be obtained from MOX fuel depends on: (1) isotopic composition of Pu, which is closely related to the discharge burnup of spent fuel from which Pu is recovered; (2) the type of fertile U makeup material used (depleted U, natural U, or recovered U); and (3) fraction of makeup material in the mix (blending ratio), which in turn determines the total fissile fraction of MOX. Using the Non-linear Reactivity Model and the code MONTEBURNS, a step-by-step procedure for computing the total fissile content of MOX is introduced. As was intended, the resulting expression is simple enough for quick/hand calculations of total fissile content of MOX required to reach a desired burnup for a given discharge burnup of spent fuel and for a specified fertile U makeup. In any case, due to non-fissile (parasitic) content of recovered Pu, a greater fissile fraction in MOX than that in fresh U is required to obtain the same burnup as can be obtained by the fresh U fuel.

Thermal performance of fresh mixed-oxide fuel in a fast flux LMR [liquid metal reactor

International Nuclear Information System (INIS)

Ethridge, J.L.; Baker, R.B.

1985-01-01

A test was designed and irradiated to provide power-to-melt (heat generation rate necessary to initiate centerline fuel melting) data for fresh mixed-oxide UO 2 -PuO 2 fuel irradiated in a fast neutron flux under prototypic liquid metal reactor (LMR) conditions. The fuel pin parameters were selected to envelope allowable fabrication ranges and address mass production of LMR fuel using sintered-to-size techniques. The test included fuel pins with variations in fabrication technique, pellet density, fuel-to-cladding gap, Pu concentration, and fuel oxygen-to-metal ratios. The resulting data base has reestablished the expected power-to-melt in mixed-oxide fuels during initial reactor startup when the fuel temperatures are expected to be the highest. Calibration of heat transfer models of fuel pin performance codes with these data are providing more accurate capability for predicting steady-state thermal behavior of current and future mixed-oxide LMR fuels

Fuel loads and fuel type mapping

Science.gov (United States)

Chuvieco, Emilio; Riaño, David; Van Wagtendonk, Jan W.; Morsdof, Felix; Chuvieco, Emilio

2003-01-01

Correct description of fuel properties is critical to improve fire danger assessment and fire behaviour modeling, since they guide both fire ignition and fire propagation. This chapter deals with properties of fuel that can be considered static in short periods of time: biomass loads, plant geometry, compactness, etc. Mapping these properties require a detail knowledge of vegetation vertical and horizontal structure. Several systems to classify the great diversity of vegetation characteristics in few fuel types are described, as well as methods for mapping them with special emphasis on those based on remote sensing images.

New type fuel exchange system

International Nuclear Information System (INIS)

Meshii, Toshio; Maita, Yasushi; Hirota, Koichi; Kamishima, Yoshio.

1988-01-01

When the reduction of the construction cost of FBRs is considered from the standpoint of the machinery and equipment, to make the size small and to heighten the efficiency are the assigned mission. In order to make a reactor vessel small, it is indispensable to decrease the size of the equipment for fuel exchange installed on the upper part of a core. Mitsubishi Heavy Industries Ltd. carried out the research on the development of a new type fuel exchange system. As for the fuel exchange system for FBRs, it is necessary to change the mode of fuel exchange from that of LWRs, such as handling in the presence of chemically active sodium and inert argon atmosphere covering it and handling under heavy shielding against high radiation. The fuel exchange system for FBRs is composed of a fuel exchanger which inserts, pulls out and transfers fuel and rotary plugs. The mechanism adopted for the new type fuel exchange system that Mitsubishi is developing is explained. The feasibility of the mechanism on the upper part of a core was investigated by water flow test, vibration test and buckling test. The design of the mechanism on the upper part of the core of a demonstration FBR was examined, and the new type fuel exchange system was sufficiently applicable. (Kako, I.)

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

International Nuclear Information System (INIS)

Knight, Travis; Anghaie, Samim

2002-01-01

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

Behavior of mixed-oxide fuel subjected to multiple thermal transients

International Nuclear Information System (INIS)

Fenske, G.R.; Hofman, G.L.; Neimark, L.A.; Poeppel, R.B.

1983-11-01

The microstructural behavior of irradiated mixed-oxide fuel subjected to multiple, mild thermal transients was investigated using direct electrical heating. The results demonstrate that significant intergranular porosity, accompanied by large-scale (>90%) release of the retained fission gas, developed as a result of the cyclic heating. Microstructural examination of the fuel indicated that thermal-shock-induced cracking of the fuel contributed significantly to the increased swelling and gas release

Fuel Mix Impacts from Transportation Fuel Carbon Intensity Standards in Multiple Jurisdictions

Science.gov (United States)

Witcover, J.

2017-12-01

Fuel carbon intensity standards have emerged as an important policy in jurisdictions looking to target transportation greenhouse gas (GHG) emissions for reduction. A carbon intensity standard rates transportation fuels based on analysis of lifecycle GHG emissions, and uses a system of deficits and tradable, bankable credits to reward increased use of fuels with lower carbon intensity ratings while disincentivizing use of fuels with higher carbon intensity ratings such as conventional fossil fuels. Jurisdictions with carbon intensity standards now in effect include California, Oregon, and British Columbia, all requiring 10% reductions in carbon intensity of the transport fuel pool over a 10-year period. The states and province have committed to grow demand for low carbon fuels in the region as part of collaboration on climate change policies. Canada is developing a carbon intensity standard with broader coverage, for fuels used in transport, industry, and buildings. This study shows a changing fuel mix in affected jurisdictions under the policy in terms of shifting contribution of transportation energy from alternative fuels and trends in shares of particular fuel pathways. It contrasts program designs across the jurisdictions with the policy, highlights the opportunities and challenges these pose for the alternative fuel market, and discusses the impact of having multiple policies alongside federal renewable fuel standards and sometimes local carbon pricing regimes. The results show how the market has responded thus far to a policy that incentivizes carbon saving anywhere along the supply chain at lowest cost, in ways that diverged from a priori policy expectations. Lessons for the policies moving forward are discussed.

Fluid flow and fuel-air mixing in a motored two-dimensional Wankel rotary engine

Science.gov (United States)

Shih, T. I.-P.; Nguyen, H. L.; Stegeman, J.

1986-01-01

The implicit-factored method of Beam and Warming was employed to obtain numerical solutions to the conservation equations of mass, species, momentum, and energy to study the unsteady, multidimensional flow and mixing of fuel and air inside the combustion chambers of a two-dimensional Wankel rotary engine under motored conditions. The effects of the following engine design and operating parameters on fluid flow and fuel-air mixing during the intake and compression cycles were studied: engine speed, angle of gaseous fuel injection during compression cycle, and speed of the fuel leaving fuel injector.

Description of a reference mixed oxide fuel fabrication plant (MOFFP)

International Nuclear Information System (INIS)

1978-01-01

In order to evaluate the environment impact, due to the Mixed Oxide Fuel Fabrication Plants, work has been initiated to describe the general design and operating conditions of a reference Mixed Oxide Fuel Fabrication Plant (MOFFP) for the 1990 time frame. The various reference data and basic assumptions for the reference MOFFP plant have been defined after discussion with experts. The data reported in this document are only made available to allow an evaluation of the environmental impact due to a reference MOFFP plant. These data have therefore not to be used as recommandation, standards, regulatory guides or requirements

Burn-up calculations for a thorium HTR with one and with two types of fuel particle

Energy Technology Data Exchange (ETDEWEB)

Griggs, C. F.

1975-06-15

Cell burn-up calculations have been made on a thorium pin-cell operating with one or with two types of particle. With one particle, the input thorium and uranium are mixed prior to irradiation and all discharged uranium is recycled. With two particles, the fuel is kept in two streams and only the uranium generated from thorium is recycled. The two models are found to give similar power generations from a given initial U-235 input. The choice between the two types of particle is probably not determined by reactor physics considerations but by the value of the fuel credits and by the cost of fuel fabrication and reprocessing.

Increase in VVER type reactor critical heat fluxes due to placing the mixing grids

International Nuclear Information System (INIS)

Bezrukov, Y.; Lisenkov, E.; Vasilchenko, I.

2011-01-01

The report deals with the results of studies of critical heat fluxes (CHF) on the models of VVER type reactor fuel assembly models equipped with the 'Vihr' intensifiers-grids. The models are the seven-rod bundles with the uniform and non-uniform axial power that correspond to two periods of FA operation i.e. beginning of cycle and end of cycle. The experiments performed showed that the mixing grids of this type are capable of increasing the FA burnout power. The power ascension rate depends on both coolant pressure and steam quality value in the CHF point. Placing the mixing grids in the bundle upper spans results in shifting the point of DNB occurrence downward along the FA height. The experimental data obtained will be used to develop the correlations for determining the CHF in the FA equipped with the mixing grids. (authors)

Behavior of mixed-oxide fuel subjected to multiple thermal transients

International Nuclear Information System (INIS)

Fenske, G.R.; Neimark, L.A.; Poeppel, R.B.; Hofman, G.L.

1985-01-01

The microstructural behavior of irradiated mixed-oxide fuel subjected to multiple, mild thermal transients was investigated using direct electrical heating. The results demonstrate that significant intergranular porosity, accompanied by large-scale (>90%) release of the retained fission gas, developed as a result of the cyclic heating. Microstructural examination of the fuel indicated that thermal-shock-induced cracking of the fuel contributed significantly to the increased swelling and gas release. 29 refs., 12 figs

DEM simulation of particle mixing for optimizing the overcoating drum in HTR fuel fabrication

Science.gov (United States)

Liu, Malin; Lu, Zhengming; Liu, Bing; Shao, Youlin

2013-06-01

The rotating drum was used for overcoating coated fuel particles in HTR fuel fabrication process. All the coated particles should be adhered to equal amount of graphite powder, which means that the particle should be mixed quickly in both radial and axial directions. This paper investigated the particle flow dynamics and mixing behavior in different regimes using the discrete element method (DEM). By varying the rotation speed, different flow regimes such as slumping, rolling, cascading, cataracting, centrifuging were produced. The mixing entropy based on radial and axial grid was introduced to describe the radial and axial mixing behaviors. From simulation results, it was found that the radial mixing can be achieved in the cascading regime more quickly than the slumping, rolling and centrifuging regimes, but the traditional rotating drum without internal components can not achieve the requirements of axial mixing and should be improved. Three different structures of internal components are proposed and simulated. The new V-shaped deflectors were found to achieve a quick axial mixing behavior and uniform axial distribution in the rotating drum based on simulation results. At last, the superiority was validated by experimental results, and the new V-shaped deflectors were used in the industrial production of the overcoating coated fuel particles in HTR fuel fabrication process.

Performance of direct alcohol fuel cells fed with mixed methanol/ethanol solutions

Energy Technology Data Exchange (ETDEWEB)

Wongyao, N. [The Joint Graduate School of Energy and Environment, King Mongkut' s University of Technology Thonburi, 126 Pracha-Uthit Rd., Bang Mod, Thung Khru, Bangkok 10140 (Thailand); Therdthianwong, A., E-mail: apichai.the@kmutt.ac.t [Fuel Cell and Hydrogen Research and Engineering Center, Clean Energy System Group, PDTI, King Mongkut' s University of Technology Thonburi, 126 Pracha-Uthit Rd., Bang Mod, Thung Khru, Bangkok 10140 (Thailand); Therdthianwong, S. [Department of Chemical Engineering, Faculty of Engineering, King Mongkut' s University of Technology Thonburi, 126 Pracha-Uthit Rd., Bang Mod, Thung Khru, Bangkok 10140 (Thailand)

2011-07-15

Research highlights: {yields} We examined the performance of direct alcohol fuel cells fed with mixed alcohol. {yields} PtRu-PtSn/C and PtRu/C as catalysts for mixed alcohol electrooxidation reaction. {yields} Misplace adsorption of ethanol on PtRu/C caused the cell performance drop. {yields} PtRu/C showed higher performance than PtRu-PtSn/C for mixed alcohol fuel. -- Abstract: In combining the advantages of both methanol and ethanol, direct alcohol fuel cells fed with mixed alcohol solutions (1 M methanol and 1 M ethanol in varying volume ratios) were tested for performance. Employing a PtRu-PtSn/C catalyst as anode, cell performance was found to diminish rapidly even at 2.5% by volume ethanol mixture. Further increase of ethanol exceeded 10%, the cell performance gradually decreased and finally approached that of direct ethanol fuel cells. The causes of the decrease in the cell performance were the slow electro-oxidation of ethanol and the misplaced adsorption of ethanol on PtRu/C. By comparing the PtRu-PtSn/C cell with the PtRu/C cell operated with mixed alcohol solutions, the cell using PtRu/C as an anode catalyst provided higher power density since more PtRu/C surface was available for methanol oxidation reaction and less ohmic resistance of PtRu/C than that of PtRu-PtSn/C. In order to reach optimization of DAFC performance fed with mixed alcohol, the electrocatalyst used for the anode must selectively adsorb an alcohol, especially ethanol.

Performance of direct alcohol fuel cells fed with mixed methanol/ethanol solutions

International Nuclear Information System (INIS)

Wongyao, N.; Therdthianwong, A.; Therdthianwong, S.

2011-01-01

Research highlights: → We examined the performance of direct alcohol fuel cells fed with mixed alcohol. → PtRu-PtSn/C and PtRu/C as catalysts for mixed alcohol electrooxidation reaction. → Misplace adsorption of ethanol on PtRu/C caused the cell performance drop. → PtRu/C showed higher performance than PtRu-PtSn/C for mixed alcohol fuel. -- Abstract: In combining the advantages of both methanol and ethanol, direct alcohol fuel cells fed with mixed alcohol solutions (1 M methanol and 1 M ethanol in varying volume ratios) were tested for performance. Employing a PtRu-PtSn/C catalyst as anode, cell performance was found to diminish rapidly even at 2.5% by volume ethanol mixture. Further increase of ethanol exceeded 10%, the cell performance gradually decreased and finally approached that of direct ethanol fuel cells. The causes of the decrease in the cell performance were the slow electro-oxidation of ethanol and the misplaced adsorption of ethanol on PtRu/C. By comparing the PtRu-PtSn/C cell with the PtRu/C cell operated with mixed alcohol solutions, the cell using PtRu/C as an anode catalyst provided higher power density since more PtRu/C surface was available for methanol oxidation reaction and less ohmic resistance of PtRu/C than that of PtRu-PtSn/C. In order to reach optimization of DAFC performance fed with mixed alcohol, the electrocatalyst used for the anode must selectively adsorb an alcohol, especially ethanol.

A binary mixed integer coded genetic algorithm for multi-objective optimization of nuclear research reactor fuel reloading

Energy Technology Data Exchange (ETDEWEB)

Binh, Do Quang [University of Technical Education Ho Chi Minh City (Viet Nam); Huy, Ngo Quang [University of Industry Ho Chi Minh City (Viet Nam); Hai, Nguyen Hoang [Centre for Research and Development of Radiation Technology, Ho Chi Minh City (Viet Nam)

2014-12-15

This paper presents a new approach based on a binary mixed integer coded genetic algorithm in conjunction with the weighted sum method for multi-objective optimization of fuel loading patterns for nuclear research reactors. The proposed genetic algorithm works with two types of chromosomes: binary and integer chromosomes, and consists of two types of genetic operators: one working on binary chromosomes and the other working on integer chromosomes. The algorithm automatically searches for the most suitable weighting factors of the weighting function and the optimal fuel loading patterns in the search process. Illustrative calculations are implemented for a research reactor type TRIGA MARK II loaded with the Russian VVR-M2 fuels. Results show that the proposed genetic algorithm can successfully search for both the best weighting factors and a set of approximate optimal loading patterns that maximize the effective multiplication factor and minimize the power peaking factor while satisfying operational and safety constraints for the research reactor.

A binary mixed integer coded genetic algorithm for multi-objective optimization of nuclear research reactor fuel reloading

International Nuclear Information System (INIS)

Binh, Do Quang; Huy, Ngo Quang; Hai, Nguyen Hoang

2014-01-01

This paper presents a new approach based on a binary mixed integer coded genetic algorithm in conjunction with the weighted sum method for multi-objective optimization of fuel loading patterns for nuclear research reactors. The proposed genetic algorithm works with two types of chromosomes: binary and integer chromosomes, and consists of two types of genetic operators: one working on binary chromosomes and the other working on integer chromosomes. The algorithm automatically searches for the most suitable weighting factors of the weighting function and the optimal fuel loading patterns in the search process. Illustrative calculations are implemented for a research reactor type TRIGA MARK II loaded with the Russian VVR-M2 fuels. Results show that the proposed genetic algorithm can successfully search for both the best weighting factors and a set of approximate optimal loading patterns that maximize the effective multiplication factor and minimize the power peaking factor while satisfying operational and safety constraints for the research reactor.

Melting temperature of uranium - plutonium mixed oxide fuel

Energy Technology Data Exchange (ETDEWEB)

Ishii, Tetsuya; Hirosawa, Takashi [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

1997-08-01

Fuel melting temperature is one of the major thermodynamical properties that is used for determining the design criteria on fuel temperature during irradiation in FBR. In general, it is necessary to evaluate the correlation of fuel melting temperature to confirm that the fuel temperature must be kept below the fuel melting temperature during irradiation at any conditions. The correlations of the melting temperature of uranium-plutonium mixed oxide (MOX) fuel, typical FBR fuel, used to be estimated and formulized based on the measured values reported in 1960`s and has been applied to the design. At present, some experiments have been accumulated with improved experimental techniques. And it reveals that the recent measured melting temperatures does not agree well to the data reported in 1960`s and that some of the 1960`s data should be modified by taking into account of the recent measurements. In this study, the experience of melting temperature up to now are summarized and evaluated in order to make the fuel pin design more reliable. The effect of plutonium content, oxygen to metal ratio and burnup on MOX fuel melting was examined based on the recent data under the UO{sub 2} - PuO{sub 2} - PuO{sub 1.61} ideal solution model, and then formulized. (J.P.N.)

Melting temperature of uranium - plutonium mixed oxide fuel

International Nuclear Information System (INIS)

Ishii, Tetsuya; Hirosawa, Takashi

1997-08-01

Fuel melting temperature is one of the major thermodynamical properties that is used for determining the design criteria on fuel temperature during irradiation in FBR. In general, it is necessary to evaluate the correlation of fuel melting temperature to confirm that the fuel temperature must be kept below the fuel melting temperature during irradiation at any conditions. The correlations of the melting temperature of uranium-plutonium mixed oxide (MOX) fuel, typical FBR fuel, used to be estimated and formulized based on the measured values reported in 1960's and has been applied to the design. At present, some experiments have been accumulated with improved experimental techniques. And it reveals that the recent measured melting temperatures does not agree well to the data reported in 1960's and that some of the 1960's data should be modified by taking into account of the recent measurements. In this study, the experience of melting temperature up to now are summarized and evaluated in order to make the fuel pin design more reliable. The effect of plutonium content, oxygen to metal ratio and burnup on MOX fuel melting was examined based on the recent data under the UO 2 - PuO 2 - PuO 1.61 ideal solution model, and then formulized. (J.P.N.)

Influence of “whirlwind” mixing grids on the critical power of WWER fuel assembly

International Nuclear Information System (INIS)

Selivanov, Yu.F.; Pomet'ko, R.S.; Volkov, S.E.

2014-01-01

The problem of optimizing the number and placement of lattices in different types assemblies is discussed. The effect of the amount of mixing lattices and their locations in assemblies on the conditions of occurrence of boiling crisis in the fuel assembly on its critical power (power of assembly in case of boiling crisis) is studied. Experiments were carried out with the use of freon as a coolant. It is recommended simultaneous use in the assembly of lattices of “whirlwind” type, well-intensifying heat exchange, and cell lattices of “pass” type (or lattices with deflectors) affecting on moving flow, provided the optimal location of lattices in the assembly [ru

Dry low NOx combustion system with pre-mixed direct-injection secondary fuel nozzle

Science.gov (United States)

Zuo, Baifang; Johnson, Thomas; Ziminsky, Willy; Khan, Abdul

2013-12-17

A combustion system includes a first combustion chamber and a second combustion chamber. The second combustion chamber is positioned downstream of the first combustion chamber. The combustion system also includes a pre-mixed, direct-injection secondary fuel nozzle. The pre-mixed, direct-injection secondary fuel nozzle extends through the first combustion chamber into the second combustion chamber.

Performance analysis of a mixed nitride fuel system for an advanced liquid metal reactor

International Nuclear Information System (INIS)

Lyon, W.F.; Baker, R.B.; Leggett, R.D.

1991-01-01

In this paper, the conceptual development and analysis of a proposed mixed nitride driver and blanket fuel system for a prototypic advanced liquid metal reactor design is performed. As a first step, an intensive literature survey is completed on the development and testing of nitride fuel systems. Based on the results of this survey, prototypic mixed nitride fuel and blanket pins is designed and analyzed using the SIEX computer code. The analysis predicts that the nitride fuel consistently operated at peak temperatures and cladding strain levels that compared quite favorably with competing fuel designs. These results, along with data available in the literature on nitride fuel performance, indicate that a nitride fuel system should offer enhanced capabilities for advanced liquid metal reactors

Performance analysis of a mixed nitride fuel system for an advanced liquid metal reactor

International Nuclear Information System (INIS)

Lyon, W.F.; Baker, R.B.; Leggett, R.D.

1990-11-01

The conceptual development and analysis of a proposed mixed nitride driver and blanket fuel system for a prototypic advanced liquid metal reactor design has been performed. As a first step, an intensive literature survey was completed on the development and testing of nitride fuel systems. Based on the results of this survey, prototypic mixed nitride fuel and blanket pins were designed and analyzed using the SIEX computer code. The analysis predicted that the nitride fuel consistently operated at peak temperatures and cladding strain levels that compared quite favorably with competing fuel designs. These results, along with data available in the literature on nitride fuel performance, indicate that a nitride fuel system should offer enhanced capabilities for advanced liquid metal reactors. 13 refs., 10 figs., 2 tabs

Fuel pellet relocation behavior in fast reactor uranium-plutonium mixed oxide fuel pin at beginning-of-life

International Nuclear Information System (INIS)

Inoue, Masaki; Ukai, Shigeharu; Asaga, Takeo

1999-08-01

The effects of fabrication parameters, irradiation conditions and fuel microstructural feature on fuel pellet relocation behavior in fast reactor fuel pins were investigated. This work focused only on beginning-of-life conditions, when fuel centerline temperature depends largely on the behavior. Fuel pellet relocation behavior in Joyo Mk-II driver could not be characterized because of the lack of data. And the behavior in FFTF driver and its larger diameter type fuel pins could not be characterized because of the extensive lot-by-lot scatters. The behavior both in Monju type and in Joyo power-to-melt type fuel pins were similar to each other, and depends largely on the as-fabricated gap width while the effects of linear heat rate and the extent of microstructural evolution were negligible. And fuel pellet centerline melting seems to affect slightly the behavior. The correlation, which describes the extent of relocation both in Monju type and in Joyo power-to-melt type fuel pins, were newly formulated and extrapolated for Joyo Mk-II driver, FFTF driver and its larger diameter type fuel pins. And the behavior in Joyo Mk-II driver seemed to be similar. On the contrary, the similarity with JNC fuel pins was observed case-by-case in FFTF driver and its larger diameter type fuel pins. (author)

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

Science.gov (United States)

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

1992-06-01

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

Mixed U/Pu oxide fabrication facility for gel-sphere-pac fuel

International Nuclear Information System (INIS)

1978-09-01

This paper describes a conceptual plant which uses the gel-sphere-pac process to fabricate mixed oxide (MOX) fuel and covers (1) fabrication of co-processed MOX fuel and (2) fabrication of co-processed spiked MOX fuel, using 60 Co. The report describes: the fuel fabrication process and plant layout, including scrap and waste processing; and maintenance safety and ventilation measures. A description of the conversion of U and Pu nitrate using a gel sphere process is given in Appendix A

Laser induced fluorescence measurements of the mixing of fuel oil with air

Energy Technology Data Exchange (ETDEWEB)

Arnold, A; Bombach, R; Hubschmid, W; Kaeppeli, B [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

1999-08-01

We report on measurements of the mixing of fuel oil with air at atmospheric pressure in an industrial premixed gas turbine burner. The concentration of the vaporized fuel oil was measured with laser induced fluorescence. We reason that the fuel oil concentration can be considered with good accuracy as proportional to the fluorescence intensity. (author) 6 fig., 3 refs.

Light water reactor mixed-oxide fuel irradiation experiment

International Nuclear Information System (INIS)

Hodge, S.A.; Cowell, B.S.; Chang, G.S.; Ryskamp, J.M.

1998-01-01

The United States Department of Energy Office of Fissile Materials Disposition is sponsoring and Oak Ridge National Laboratory (ORNL) is leading an irradiation experiment to test mixed uranium-plutonium oxide (MOX) fuel made from weapons-grade (WG) plutonium. In this multiyear program, sealed capsules containing MOX fuel pellets fabricated at Los Alamos National Laboratory (LANL) are being irradiated in the Advanced Test Reactor (ATR) at the Idaho National Engineering and Environmental Laboratory (INEEL). The planned experiments will investigate the utilization of dry-processed plutonium, the effects of WG plutonium isotopics on MOX performance, and any material interactions of gallium with Zircaloy cladding

The prospect of uranium nitride (UN) and mixed nitride fuel (UN-PuN) for pressurized water reactor

International Nuclear Information System (INIS)

Syarifah, Ratna Dewi; Suud, Zaki

2015-01-01

Design study of small Pressurized Water Reactors (PWRs) core loaded with uranium nitride fuel (UN) and mixed nitride fuel (UN-PuN), Pa-231 as burnable poison, and Americium has been performed. Pa-231 known as actinide material, have large capture cross section and can be converted into fissile material that can be utilized to reduce excess reactivity. Americium is one of minor actinides with long half life. The objective of adding americium is to decrease nuclear spent fuel in the world. The neutronic analysis results show that mixed nitride fuel have k-inf greater than uranium nitride fuel. It is caused by the addition of Pu-239 in mixed nitride fuel. In fuel fraction analysis, for uranium nitride fuel, the optimum volume fractions are 45% fuel fraction, 10% cladding and 45% moderator. In case of UN-PuN fuel, the optimum volume fractions are 30% fuel fraction, 10% cladding and 60% coolant/ moderator. The addition of Pa-231 as burnable poison for UN fuel, enrichment U-235 5%, with Pa-231 1.6% has k-inf more than one and excess reactivity of 14.45%. And for mixed nitride fuel, the lowest value of reactivity swing is when enrichment (U-235+Pu) 8% with Pa-231 0.4%, the excess reactivity value 13,76%. The fuel pin analyze for the addition of Americium, the excess reactivity value is lower than before, because Americium absorb the neutron. For UN fuel, enrichment U-235 8%, Pa-231 1.6% and Am 0.5%, the excess reactivity is 4.86%. And for mixed nitride fuel, when enrichment (U-235+Pu) 13%, Pa-231 0.4% and Am 0.1%, the excess reactivity is 11.94%. For core configuration, it is better to use heterogeneous than homogeneous core configuration, because the radial power distribution is better

The prospect of uranium nitride (UN) and mixed nitride fuel (UN-PuN) for pressurized water reactor

Science.gov (United States)

Syarifah, Ratna Dewi; Suud, Zaki

2015-09-01

Design study of small Pressurized Water Reactors (PWRs) core loaded with uranium nitride fuel (UN) and mixed nitride fuel (UN-PuN), Pa-231 as burnable poison, and Americium has been performed. Pa-231 known as actinide material, have large capture cross section and can be converted into fissile material that can be utilized to reduce excess reactivity. Americium is one of minor actinides with long half life. The objective of adding americium is to decrease nuclear spent fuel in the world. The neutronic analysis results show that mixed nitride fuel have k-inf greater than uranium nitride fuel. It is caused by the addition of Pu-239 in mixed nitride fuel. In fuel fraction analysis, for uranium nitride fuel, the optimum volume fractions are 45% fuel fraction, 10% cladding and 45% moderator. In case of UN-PuN fuel, the optimum volume fractions are 30% fuel fraction, 10% cladding and 60% coolant/ moderator. The addition of Pa-231 as burnable poison for UN fuel, enrichment U-235 5%, with Pa-231 1.6% has k-inf more than one and excess reactivity of 14.45%. And for mixed nitride fuel, the lowest value of reactivity swing is when enrichment (U-235+Pu) 8% with Pa-231 0.4%, the excess reactivity value 13,76%. The fuel pin analyze for the addition of Americium, the excess reactivity value is lower than before, because Americium absorb the neutron. For UN fuel, enrichment U-235 8%, Pa-231 1.6% and Am 0.5%, the excess reactivity is 4.86%. And for mixed nitride fuel, when enrichment (U-235+Pu) 13%, Pa-231 0.4% and Am 0.1%, the excess reactivity is 11.94%. For core configuration, it is better to use heterogeneous than homogeneous core configuration, because the radial power distribution is better.

Platinum- and membrane-free swiss-roll mixed-reactant alkaline fuel cell.

Science.gov (United States)

Aziznia, Amin; Oloman, Colin W; Gyenge, Előd L

2013-05-01

Eliminating the expensive and failure-prone proton exchange membrane (PEM) together with the platinum-based anode and cathode catalysts would significantly reduce the high capital and operating costs of low-temperature (<373 K) fuel cells. We recently introduced the Swiss-roll mixed-reactant fuel cell (SR-MRFC) concept for borohydride-oxygen alkaline fuel cells. We now present advances in anode electrocatalysis for borohydride electrooxidation through the development of osmium nanoparticulate catalysts supported on porous monolithic carbon fiber materials (referred to as an osmium 3D anode). The borohydride-oxygen SR-MRFC operates at 323 K and near atmospheric pressure, generating a peak power density of 1880 W m(-2) in a single-cell configuration by using an osmium-based anode (with an osmium loading of 0.32 mg cm(-2)) and a manganese dioxide gas-diffusion cathode. To the best of our knowledge, 1880 W m(-2) is the highest power density ever reported for a mixed-reactant fuel cell operating under similar conditions. Furthermore, the performance matches the highest reported power densities for conventional dual chamber PEM direct borohydride fuel cells. Copyright © 2013 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Determination of melting point of mixed-oxide fuel irradiated in a fast breeder reactor

International Nuclear Information System (INIS)

Tachibana, Toshimichi

1985-01-01

The melting point of fuel is important to set its in-reactor maximum temperature in fuel design. The fuel melting point measuring methods are broadly the filament method and the capsule sealing method. The only instance of measuring the melting point of irradiated mixed oxide (U, Pu)O 2 fuel by the filament method is by GE in the United States. The capsule sealing method, while the excellent means, is difficult in weld sealing the irradiated fuel in a capsule within the cell. In the fast reactor development program, the remotely operated melting point measuring apparatus in capsule sealing the mixed (U, Pu)O 2 fuel irradiated in the experimental FBR Joyo was set in the cell and the melting point was measured, for the first time in the world. (Mori, K.)

Mixed fuel strategy for carbon deposition mitigation in solid oxide fuel cells at intermediate temperatures.

Science.gov (United States)

Su, Chao; Chen, Yubo; Wang, Wei; Ran, Ran; Shao, Zongping; Diniz da Costa, João C; Liu, Shaomin

2014-06-17

In this study, we propose and experimentally verified that methane and formic acid mixed fuel can be employed to sustain solid oxide fuel cells (SOFCs) to deliver high power outputs at intermediate temperatures and simultaneously reduce the coke formation over the anode catalyst. In this SOFC system, methane itself was one part of the fuel, but it also played as the carrier gas to deliver the formic acid to reach the anode chamber. On the other hand, the products from the thermal decomposition of formic acid helped to reduce the carbon deposition from methane cracking. In order to clarify the reaction pathways for carbon formation and elimination occurring in the anode chamber during the SOFC operation, O2-TPO and SEM analysis were carried out together with the theoretical calculation. Electrochemical tests demonstrated that stable and high power output at an intermediate temperature range was well-maintained with a peak power density of 1061 mW cm(-2) at 750 °C. With the synergic functions provided by the mixed fuel, the SOFC was running for 3 days without any sign of cell performance decay. In sharp contrast, fuelled by pure methane and tested at similar conditions, the SOFC immediately failed after running for only 30 min due to significant carbon deposition. This work opens a new way for SOFC to conquer the annoying problem of carbon deposition just by properly selecting the fuel components to realize their synergic effects.

Determination of mixing factors for VVER-440 fuel assembly head

Energy Technology Data Exchange (ETDEWEB)

Tóth, S., E-mail: toth@reak.bme.hu [Institute of Nuclear Techniques, Budapest University of Technology and Economics, Műegyetem rkp. 9, H-1111 Budapest (Hungary); Aszódi, A. [Institute of Nuclear Techniques, Budapest University of Technology and Economics, Műegyetem rkp. 9, H-1111 Budapest (Hungary)

2013-11-15

CFD models have been developed for the heads of the old, the present and the new type VVER-440 fuel assemblies using the experience of a former validation process. With these models temperature distributions are investigated in the heads of some typical assemblies and the in-core thermocouple signals are calculated. The analyses show that the coolant mixing is intensive but not-perfect in the assembly heads. The difference between the thermocouple signal and the cross-sectional average temperature at the measurement level depends on the assembly type. Using the results of these CFD calculations the weight factors of the rod bundle regions for the in-core thermocouple have been determined. With these factors the thermocouple signals are estimated and the results are statistically tested using the registered data of the Hungarian nuclear power plant. This test shows that the deviations between the measured and the calculated temperatures can be significantly decreased and consequently monitoring uncertainties can be reduced with using the weight factors.

Molten carbonate fuel cell cathode with mixed oxide coating

Science.gov (United States)

Hilmi, Abdelkader; Yuh, Chao-Yi

2013-05-07

A molten carbonate fuel cell cathode having a cathode body and a coating of a mixed oxygen ion conductor materials. The mixed oxygen ion conductor materials are formed from ceria or doped ceria, such as gadolinium doped ceria or yttrium doped ceria. The coating is deposited on the cathode body using a sol-gel process, which utilizes as precursors organometallic compounds, organic and inorganic salts, hydroxides or alkoxides and which uses as the solvent water, organic solvent or a mixture of same.

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.

Vehicle type choice under the influence of a tax reform and rising fuel prices

DEFF Research Database (Denmark)

Mabit, Stefan Lindhard

2014-01-01

change in new vehicle purchases toward more diesel vehicles and more fuel-efficient vehicles. The paper analyses to what extent a vehicle tax reform similar to the Danish 2007 reform may explain changes in purchasing behaviour. The paper investigates the effects of a tax reform, fuel price changes......, and technological development on vehicle type choice using a mixed logit model. The model allows a simulation of the effect of car price changes that resemble those induced by the tax reform. This effect is compared to the effects of fuel price changes and technology improvements. The simulations show...... that the effect of the tax reform on fuel efficiency is similar to the effect of rising fuel prices while the effect of technological development is much larger. The conclusion is that while the tax reform appeared in the same year as a large increase in fuel efficiency, it seems likely that it only explains...

Operating method of molten carbonate type fuel cell

Energy Technology Data Exchange (ETDEWEB)

Nakanishi, Tsuneo

1988-12-06

Molten carbonate type fuel cell involves a problem of oxidation of anode while the unit is stopped. Although there is a method proposed wherein an inactive gas is supplied to anode during the stoppage, the market-available inactive gas contains a slight amount of oxygen which makes it difficult to prevent the deterioration of the anode. In this invention, at the start and the stop other than the normal operation, a protective gas mixture of an inactive gas with a small amount of hydrogen is supplied to the anode. The inactive gas is a commercial type nitrogen, argon or helium; hydrogen is mixed in amount 0.5 - 2.0% of the inactive gas. By this method, oxygen in air which comes in from the gas-sealed portion of the cell is reduced by hydrogen in the protective gas and is discharged in the form of water. 2 figs.

On the mixing model for calculating the temperature fields in nuclear reactor fuel assemblies

International Nuclear Information System (INIS)

Mikhin, V.I.; Zhukov, A.V.

1985-01-01

One of the alternatives of the mixing model applied for calculating temperature fields in nuclear reactor fuel assemblies,including the fuel assemblies with nonequilibrium energy-release in fuel element cross section, is consistently described. The equations for both constant and variable values of coolant density and heat capacity are obtained. The mixing model is based on a set of mass, heat and longitudinal momentum balance equations. This set is closed by the ratios connecting the unknown values for gaps between fuel elements with the averaged values for neighbouring channels. The ratios to close momentum and heat balance equations, explaining, in particular, the nonequivalent heat and mass, momentum and mass transfer coefficients, are suggested. The balance equations with variable coolant density and heat capacity are reduced to the form coinciding with those of the similar equations with constant values of these parameters. Application of one of the main ratios of the mixing model relating the coolant transverse overflow in the gaps between fuel elements to the averaged coolant rates (flow rates) in the neighbouring channels is mainly limited by the coolant stabilized flow in the fuel assemblies with regular symmetrical arrangement of elements. Mass transfer coefficients for these elements are experimentally determined. The ratio in the paper is also applicable for calculation of fuel assembly temperature fields with a small relative shift of elements

Dissolution of mixed oxide spent fuel from FBR

International Nuclear Information System (INIS)

Sanyoshi, H.; Nishina, H.; Toyota, O.; Yamamoto, R.; Nemoto, S.; Okamoto, F.; Togashi, A.; Kawata, T.; Hayashi, S.

1991-01-01

At the Tokai Works of the Power Reactor and Nuclear Fuel Development Corporation (PNC), the Chemical Processing Facility (CPF) has been continuing operation since 1982 for laboratory scale hot experiments on reprocessing of FBR mixed oxide fuel. As a part of these experiments, dissolution experiments have been performed to define the key parameters affecting dissolution rates such as concentration of nitric acid, temperature and burnup and also to confirm the amount of insoluble residue. The dissolution rate of the irradiated fuel was determined to be in proportion to the 1.7 power of the nitric acid concentration. The activation energy determined from the experiments varied from 6 to 11 kcal/mol depending on the method of dissolution. The dissolution rate decreased as the fuel burnup increased in low nitric acid media below 5 mol/l. However, it was found that the effect of the burnup became negligible in a high concentration of nitric acid media. The amount of insoluble residue and its constituents were evaluated by changing the dissolution condition. (author)

Prediction for the flow distribution and the pressure drop of a plate type fuel assembly

International Nuclear Information System (INIS)

Park, Jong Hark; Jo, Dea Sung; Chae, Hee Taek; Lee, Byung Chul

2011-01-01

A plate type fuel assembly widely used in many research reactors does not allow the coolant to mix with neighboring fuel channels due to the completely separated flow channels. If there is a serious inequality of coolant distribution among channels, it can reduce thermal-hydraulic safety margin, as well as it can cause a deformation of fuel plates by the pressure difference between neighboring channels, thus the flow uniformity in the fuel assembly should be confirmed. When designing a primary cooling system (PCS), the pressure drop through a reactor core is a dominant value to determine the PCS pump size. The major portion of reactor core pressure drop is caused by the fuel assemblies. However it is not easy to get a reasonable estimation of pressure drop due to the geometric complexity of the fuel assembly and the thin gaps between fuel assemblies. The flow rate through the gap is important part to determine the total flow rate of PCS, so it should be estimated as reasonable as possible. It requires complex and difficult jobs to get useful data. In this study CFD analysis to predict the flow distribution and the pressure drop were conducted on the plate type fuel assembly, which results would be used to be preliminary data to determine the PCS flow rate and to improve the design of a fuel assembly

Development of a new technique for experimental evaluation of the fuel element's subchannel mixing

International Nuclear Information System (INIS)

Silin, Nicolas; Delmastro, Dario; Juanico, Luis

2004-01-01

In this work, the development of a new experimental method for the measurement of mixing between the cooling subchannels of nuclear fuel elements by using thermal traces, is presented.The method has been proved on a reduced test section with very positive results, having demonstrated its simplicity and low cost.Because it is suitable for heterogeneous and compact subchannels (asArgentinean fuels) with high water flows in simple and affordable tests at atmospheric pressure, this new method is specially well suited for the design of fuel elements, while it offers advantages over other methods of mixing measurement [es

Fuel-steel mixing and radial mesh effects in power excursion simulations

International Nuclear Information System (INIS)

Chen, X.-N.; Rineiski, A.; Gabrielli, F.; Andriolo, L.; Vezzoni, B.; Li, R.; Maschek, W.; Kiefhaber, E.

2016-01-01

Highlights: • Fuel-steel mixing and radial mesh effects are significant on power excursion. • The earliest power peak is reduced and retarded by these two effects. • Unprotected loss of coolant transients in ESFR core are calculated. - Abstract: This paper deals with SIMMER-III once-through simulations of the earliest power excursion initiated by an unprotected loss of flow (ULOF) in the Working Horse design of the European Sodium Cooled Fast Reactor (ESFR). Since the sodium void effect is strictly positive in this core and dominant in the transient, a power excursion is initiated by sodium boiling in the ULOF case. Two major effects, namely (1) reactivity effects due to fuel-steel mixing after melting and (2) the radial mesh size, which were not considered originally in SIMMER simulations for ESFR, are studied. The first effect concerns the reactivity difference between the heterogeneous fuel/clad/wrapper configuration and the homogeneous mixture of steel and fuel. The full core homogenization (due to melting) effect is −2 $, though a smaller effect takes place in case of partial core melting. The second effect is due to the SIMMER sub-assembly (SA) coarse mesh treatment, where a simultaneous sodium boiling onset in all SAs belonging to one ring leads to an overestimated reactivity ramp. For investigating the influence of fuel/steel mixing effects, a lumped “homogenization” reactivity feedback has been introduced, being proportional to the molten steel mass. For improving the coarse mesh treatment, we employ finer radial meshes to take the subchannel effects into account, where the side and interior channels have different coolant velocities and temperatures. The simulation results show that these two effects have significant impacts on the earliest power excursion after the sodium boiling.

Neutronic analysis concerning the utilization of mixed U N-Pu N nitride fuel for fast reactors

International Nuclear Information System (INIS)

Renke, C.A.C.; Batista, J.L.; Waintraub, M.; Santos Bastos, W. dos; Brito Aghina, L.O. de.

1991-08-01

Neutronic behavior of mixed UN-PuN nitride fuel in substitution of the mixed oxide U O 2 - Pu O 2 for fast reactors is discussed with focus on Super Phenix I. Characteristics parameters of both cores are calculated and compared and the results presented show a great advantage for the nitride fuel, pointing out a larger performance of fuel elements in the core and an effective reduction of reactivity loss during the cycle. (author)

Experimental investigation of effect of spacer on two phase turbulent mixing rate in subchannels of pressure tube type BWR

Energy Technology Data Exchange (ETDEWEB)

Verma, Shashi Kant; Sinha, S.L. [National Institute of Technology, Raipur (India). Mechanical Engineering Dept.; Chandraker, D.K. [Bhabha Atomic Research Centre, Mumbai (India). Reactor Design and Development Group

2017-11-15

Turbulent mixing rate between adjacent subchannels in a two-phase flow has been known to be strongly dependent on the flow pattern. The most important aspect of turbulent motion is that the velocity and pressure at a fixed point do not remain constant with time even in steady state but go through very irregular high frequency fluctuations. These fluctuations influence the diffusion of scalar and vector quantities. The Advanced Heavy Water Reactor (AHWR) is a vertical pressure tube type, heavy water moderated and boiling light water cooled natural circulation based reactor. The fuel bundle of AHWR contains 54 fuel rods set in three concentric rings of 12, 18 and 24 fuel rods. This fuel bundle is divided into number of imaginary interacting flow channel called subchannels. Alteration from single phase to two phase flow situation occurs in reactor rod bundle with raise in power. The two phase flow regimes like bubbly, slug-churn, and annular flow are generally encountered in reactor rod bundle. Prediction of thermal margin of the reactor has necessitated the investigation of turbulent mixing rate of coolant between these subchannels under these flow regimes. Thus, it is fundamental to estimate the effect of spacer grids on turbulent mixing between subchannels of AHWR rod bundle.

Emission computer tomography on a Dodewaard mixed oxide fuel pin

International Nuclear Information System (INIS)

Buurveld, H.A.; Dassel, G.

1993-12-01

A nondestructive technique as well as a destructive PIE technique have been used to verify the results obtained with a newly 8-e computer tomography (GECT) system. Multi isotope Scanning (MIS), electron probe micro analysis (EPMA) and GECT were used on a mixed oxide (MOX) fuel rod from the Dodewaard reactor with an average burnup of 24 MWd/kg fuel. GECT shows migration of Cs to the periphery of fuel pellets and to radial cracks and pores in the fuel, whereas MIS shows Cs migration to pellet interfaces. The EPMA technique appeared not to be useful to show migration of Cs but, it shows the distribution of fission products from Pu. EPMA clearly shows the distribution of fission products from Pu, but did not reveal the Cs-migration. (orig./HP)

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

International Nuclear Information System (INIS)

Harry, G.R.

1983-01-01

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

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

International Nuclear Information System (INIS)

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

1980-10-01

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

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

International Nuclear Information System (INIS)

Huwyler, S.

1977-09-01

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

Understanding fuel magnetization and mix using secondary nuclear reactions in magneto-inertial fusion.

Science.gov (United States)

Schmit, P F; Knapp, P F; Hansen, S B; Gomez, M R; Hahn, K D; Sinars, D B; Peterson, K J; Slutz, S A; Sefkow, A B; Awe, T J; Harding, E; Jennings, C A; Chandler, G A; Cooper, G W; Cuneo, M E; Geissel, M; Harvey-Thompson, A J; Herrmann, M C; Hess, M H; Johns, O; Lamppa, D C; Martin, M R; McBride, R D; Porter, J L; Robertson, G K; Rochau, G A; Rovang, D C; Ruiz, C L; Savage, M E; Smith, I C; Stygar, W A; Vesey, R A

2014-10-10

Magnetizing the fuel in inertial confinement fusion relaxes ignition requirements by reducing thermal conductivity and changing the physics of burn product confinement. Diagnosing the level of fuel magnetization during burn is critical to understanding target performance in magneto-inertial fusion (MIF) implosions. In pure deuterium fusion plasma, 1.01 MeV tritons are emitted during deuterium-deuterium fusion and can undergo secondary deuterium-tritium reactions before exiting the fuel. Increasing the fuel magnetization elongates the path lengths through the fuel of some of the tritons, enhancing their probability of reaction. Based on this feature, a method to diagnose fuel magnetization using the ratio of overall deuterium-tritium to deuterium-deuterium neutron yields is developed. Analysis of anisotropies in the secondary neutron energy spectra further constrain the measurement. Secondary reactions also are shown to provide an upper bound for the volumetric fuel-pusher mix in MIF. The analysis is applied to recent MIF experiments [M. R. Gomez et al., Phys. Rev. Lett. 113, 155003 (2014)] on the Z Pulsed Power Facility, indicating that significant magnetic confinement of charged burn products was achieved and suggesting a relatively low-mix environment. Both of these are essential features of future ignition-scale MIF designs.

Fuel-Coolant Interactions: Visualization and Mixing Measurements

International Nuclear Information System (INIS)

Loewen, Eric P.; Bonazza, Riccardo; Corradini, Michael L.; Johannesen, Robert E.

2002-01-01

Dynamic X-ray imaging of fuel-coolant interactions (FCI), including quantitative measurement of fuel-coolant volume fractions and length scales, has been accomplished with a novel imaging system at the Nuclear Safety Research Center at the University of Wisconsin, Madison. The imaging system consists of visible-light high-speed digital video, low-energy X-ray digital imaging, and high-energy X-ray digital imaging subsystems. The data provide information concerning the melt jet velocity, melt jet configuration, melt volume fractions, void fractions, and spatial and temporal quantification of premixing length scales for a model fuel-coolant system of molten lead poured into a water pool (fuel temperatures 500 to 1000 K; jet diameters 10 to 30 mm; coolant temperatures 20 to 90 deg. C). Overall results indicate that the FCI has three general regions of behavior, with the high fuel-coolant temperature region similar to what might be expected under severe accident conditions. It was observed that the melt jet leading edge has the highest void fraction and readily fragments into discrete masses, which then subsequently subdivide into smaller masses of length scales <10 mm. The intact jet penetrates <3 to 5 jet length/jet diameter before this breakup occurs into discrete masses, which continue to subdivide. Hydrodynamic instabilities can be visually identified at the leading edge and along the jet column with an interfacial region that consists of melt, vapor, and water. This interface region was observed to grow in size as the water pool temperature was increased, indicating mixing enhancement by boiling processes

Study of reactions between fuel (mixed oxide (UPu)Osub(2-x)) and cladding (stainless-steel) in reactors: influence of iodine compounds

International Nuclear Information System (INIS)

Aubert, Michel.

1976-03-01

The influence of iodine compounds on the development of the oxide-cladding reaction was examined. The action of iodine, cesium and cesium iodide on type 316 stainless was determined in the presence or absence of uranium oxide or mixed uranium-plutonium oxide type fuel in a closed system, isothermal or with a temperature gradient. The study of the stainless steel iodine reactions was developed in particular. These experiments showed that cesium combines with uranium oxide to give cesium uranate Cs 2 U 2 O 7 ; it is not unreasonable to suppose that cesium urano-plutonate Cs 2 (U,Pu) 2 O 7 could be formed inside the pile. It was then shown that cesium iodide in the presence of sufficiently non-stoichiometric mixed oxide could contribute towards the degradation of the stainless steel cladding. Under these conditions the reaction is accompained by a transport of manganese, chromium and iron into the hot parts of the fuel by a Van-Arkel type mechanism. This might explain the presence of metallic precipitates in the fuel, but the role assigned to molybdenum iodide in the same phenomenon is considered unlikely. Finally it is proposed to deposit a thin layer of manganese metal on the inner surface of the cladding in order to minimize the action of fission products (CsI, Te) [fr

Material control in nuclear fuel fabrication facilities. Part I. Fuel descriptions and fabrication processes, P.O. 1236909 Final report

International Nuclear Information System (INIS)

Borgonovi, G.M.; McCartin, T.J.; Miller, C.L.

1978-12-01

The report presents information on foreign nuclear fuel fabrication facilities. Fuel descriptions and fuel fabrication information for three basic reactor types are presented: The information presented for LWRs assumes that Pu--U Mixed Oxide Fuel (MOX) will be used as fuel

Reliability of fast reactor mixed-oxide fuel during operational transients

International Nuclear Information System (INIS)

Boltax, A.; Neimark, L.A.; Tsai, Hanchung; Katsuragawa, M.; Shikakura, S.

1991-07-01

Results are presented from the cooperative DOE and PNC Phase 1 and 2 operational transient testing programs conducted in the EBR-2 reactor. The program includes second (D9 and PNC 316 cladding) and third (FSM, AST and ODS cladding) generation mixed-oxide fuel pins. The irradiation tests include duty cycle operation and extended overpower tests. the results demonstrate the capability of second generation fuel pins to survive a wide range of duty cycle and extended overpower events. 15 refs., 9 figs., 4 tabs

The use of the average plutonium-content for criticality evaluation of boiling water reactor mixed oxide-fuel transport and storage packages

International Nuclear Information System (INIS)

Mattera, C.

2003-01-01

Currently in France, criticality studies in transport configurations for Boiling Water Reactor Mixed Oxide fuel assemblies are based on conservative hypothesis assuming that all rods (Mixed Oxide (Uranium and Plutonium), Uranium Oxide, Uranium and (Gadolinium Oxide rods) are Mixed Oxide rods with the same Plutonium-content, corresponding to the maximum value. In that way, the real heterogeneous mapping of the assembly is masked and covered by an homogenous Plutonium-content assembly, enriched at the maximum value. As this calculation hypothesis is extremely conservative, Cogema Logistics (formerly Transnucleaire) has studied a new calculation method based on the use of the average Plutonium-content in the criticality studies. The use of the average Plutonium-content instead of the real Plutonium-content profiles provides a highest reactivity value that makes it globally conservative. This method can be applied for all Boiling Water Reactor Mixed Oxide complete fuel assemblies of type 8 x 8, 9 x 9 and 10 x 10 which Plutonium-content in mass weight does not exceed 15%; it provides advantages which are discussed in the paper. (author)

Mixing Characteristics during Fuel Coolant Interaction under Reactor Submerged Conditions

International Nuclear Information System (INIS)

Hong, S. W.; Na, Y. S.; Hong, S. H.; Song, J. H.

2014-01-01

A molten material is injected into an interaction chamber by free gravitation fall. This type of fuel coolant interaction could happen to operating plants. However, the flooding of a reactor cavity is considered as SAM measures for new PWRs such as APR-1400 and AP1000 to assure the IVR of a core melt. In this case, a molten corium in a reactor is directly injected into water surrounding the reactor vessel without a free fall. KAERI has carried out fuel coolant interaction tests without a free fall using ZrO 2 and corium to simulate the reactor submerged conditions. There are four phases in a steam explosion. The first phase is a premixing phase. The premixing is described in the literature as follows: during penetration of melt into water, hydrodynamic instabilities, generated by the velocities and density differences as well as vapor production, induce fragmentation of the melt into particles; the particles fragment in turn into smaller particles until they reach a critical size such that the cohesive forces (surface tension) balance exactly the disruptive forces (inertial); and the molten core material temperature (>2500 K) is such that the mixing always occurs in the film boiling regime of the water: It is very important to qualify and quantify this phase because it gives the initial conditions for a steam explosion This paper mainly focuses on the observation of the premixing phase between a case with 1 m free fall and a case without a free fall to simulate submerged reactor condition. The premixing behavior between a 1m free fall case and reactor case submerged without a free fall is observed experimentally. The average velocity of the melt front passing through 1m water pool; - Case without a free fall: The average velocity of corium, 2.7m/s, is faster than ZrO 2 , 2.3m/s, in water. - Cases of with a 1 m free fall and without a free fall : The case without a free fall is about two times faster than a case with a 1 m free fall. Bubble characteristics; - Case

Chemical states of fission products in irradiated uranium-plutonium mixed oxide fuel

International Nuclear Information System (INIS)

Kurosaki, Ken; Uno, Masayoshi; Yamanaka, Shinsuke

1999-01-01

The chemical states of fission products (FPs) in irradiated uranium-plutonium mixed oxide (MOX) fuel for the light water reactor (LWR) were estimated by thermodynamic equilibrium calculations on system of fuel and FPs by using ChemSage program. A stoichiometric MOX containing 6.1 wt. percent PuO 2 was taken as a loading fuel. The variation of chemical states of FPs was calculated as a function of oxygen potential. Some pieces of information obtained by the calculation were compared with the results of the post-irradiation examination (PIE) of UO 2 fuel. It was confirmed that the multicomponent and multiphase thermodynamic equilibrium calculation between fuel and FPs system was an effective tool for understanding the behavior of FPs in fuel. (author)

Rich-burn, flame-assisted fuel cell, quick-mix, lean-burn (RFQL) combustor and power generation

Science.gov (United States)

Milcarek, Ryan J.; Ahn, Jeongmin

2018-03-01

Micro-tubular flame-assisted fuel cells (mT-FFC) were recently proposed as a modified version of the direct flame fuel cell (DFFC) operating in a dual chamber configuration. In this work, a rich-burn, quick-mix, lean-burn (RQL) combustor is combined with a micro-tubular solid oxide fuel cell (mT-SOFC) stack to create a rich-burn, flame-assisted fuel cell, quick-mix, lean-burn (RFQL) combustor and power generation system. The system is tested for rapid startup and achieves peak power densities after only 35 min of testing. The mT-FFC power density and voltage are affected by changes in the fuel-lean and fuel-rich combustion equivalence ratio. Optimal mT-FFC performance favors high fuel-rich equivalence ratios and a fuel-lean combustion equivalence ratio around 0.80. The electrical efficiency increases by 150% by using an intermediate temperature cathode material and improving the insulation. The RFQL combustor and power generation system achieves rapid startup, a simplified balance of plant and may have applications for reduced NOx formation and combined heat and power.

Enrichment measurement in TRIGA type fuels

International Nuclear Information System (INIS)

Aguilar H, F.; Mazon R, R.

2001-05-01

The Department of Energy of the United States of North America, through the program 'Idaho Operations Nuclear Spent Fuel Program' of the Idaho National Engineering and Environmental Laboratory (INEEL), in Idaho Falls; Idaho USA, hires to Global Technologies Inc. (GTI) to develop a prototype device of detection enrichment uranium (DEU Detection of Enrichment of Uranium) to determine quantitatively the enrichment in remainder U-235 in a TRIGA fuel element at the end of it useful life. The characteristics of the prototype developed by GTI are the following ones: It allows to carry out no-destructive measurements of TRIGA type fuel. Easily transportable due to that reduced of it size. The determination of the enrichment (in grams of U-235) it is obtained with a precision of 5%. The National Institute of Nuclear Research (ININ), in its facilities of the Nuclear Center of Mexico, it has TRIGA type fuel of high and low enrichment (standard and FLIP) fresh and with burnt, it also has the infrastructure (hot cells, armor-plating of transport, etc) and qualified personnel to carry out the necessary maneuvers to prove the operation of the DEU prototype. For this its would be used standard type fuel elements and FLIP, so much fresh as with certain burnt one. In the case of the fresh fuels the measurement doesn't represent any risk, the fuels before and after the measurement its don't contain a quantity of fission products that its represent a radiological risk in its manipulation; but in the case of the fuels with burnt the handling of the same ones represents an important radiological risk reason why for its manipulation it was used the transport armor-plating and the hot cells. (Author)

Some alternatives to the mixed oxide fuel cycle

International Nuclear Information System (INIS)

Deonigi, D.E.; Eschbach, E.A.; Goldsmith, S.; Pankaskie, P.J.; Rohrmann, C.A.; Widrig, R.D.

1977-02-01

While on initial examination each of the six fuel cycle concepts (tandem cycle, extended burnup, fuel rejuvenation, coprocessing, partial reprocessing, and thorium) described in the report may have some potential for improving safeguards, none of the six appears to have any other major or compelling advantages over the mixed oxide (MOX) fuel cycle. Compared to the MOX cycle, all but coprocessing appear to have major disadvantages, including severe cost penalties. Three of the concepts-tandem, extended burnup, and rejuvenation--share the basic problems of the throwaway cycle (GESMO Alternative 6): without reprocessing, high-level waste volumes and costs are substantially increased, and overall uranium utilization decreases for three reasons. First, the parasitic fission products left in the fuel absorb neutrons in later irradiation steps reducing the overall neutronic efficiencies of these cycles. Second, discarded fuel still has sufficient fissile values to warrant recycle. Third, perhaps most important, the plutonium needed for breeder start-up will not be available; without the breeder, uranium utilization would drop by about a factor of sixty. Two of the concepts--coprocessing and partial reprocessing--involve variations of the basic MOX fuel cycle's chemical reprocessing step to make plutonium diversion potentially more difficult. These concepts could be used with the MOX fuel cycle or in conjunction with the tandem, extended burnup and rejuvenation concepts to eliminate some of the problems with those cycles. But in so doing, the basic impetus for those cycles--elimination of reprocessing for safeguards purposes--no longer exists. Of all the concepts considered, only coprocessing--and particularly the ''master blend'' version--appears to have sufficient promise to warrant a more detailed study. The master blend concept could possibly make plutonium diversion more difficult with minimal impact on the reprocessing and MOX fuel fabrication operations

Fuel assembly for FBR type reactor

International Nuclear Information System (INIS)

Hayashi, Hideyuki.

1995-01-01

Ordinary sodium bond-type fuel pins using nitride fuels, carbide fuels or metal fuels and pins incorporated with hydride moderators are loaded in a wrapper tube at a ratio of from 2 to 10% based on the total number of fuel pins. The hydride moderators are sealed in the hydride moderator incorporated pins at the position only for a range from the upper end to a reactor core upper position of substantially 1/4 of the height of the reactor core from the upper end of the reactor core as a center. Then, even upon occurrence of ULOF (loss of flow rate scram failure phenomenon), it gives characteristic of reducing the power only by a doppler coefficient and not causing boiling of coolant sodium but providing stable cooling to the reactor core. Therefore, a way of thinking on the assurance of passive safety is simplified to make a verification including on the reactor structure unnecessary. In an LMFBR type reactor using the fuel assembly, a critical experiment for confirming accuracy of nuclear design is sufficient for the item required for study and development, which provides a great economical effect. (N.H.)

Behavior of mixed-oxide fuel elements during the TOPI-1E transient overpower test

International Nuclear Information System (INIS)

Tsai, H.; Neimark, L.A.; Yamamoto, K.; Hirai, K.; Shikakura, S.

1993-12-01

A slow-ramp, extended overpower transient test was conducted on a group of nineteen preirradiated mixed-oxide fuel elements in EBR-II. During the transient two of the test elements with high-density fuel and tempered martensitic cladding (PNC-FMS) breached at an overpower of ∼75%. Fuel elements with austenitic claddings (D9, PNC316, and PNC150), many with aggressive design features and high burnups, survived the overpower transient and incurred little or no cladding strain. Fuel elements with annual fuel or heterogeneous fuel columns also behaved well

Fuel-cladding chemical interaction in mixed-oxide fuels

International Nuclear Information System (INIS)

Lawrence, L.A.; Weber, J.W.; Devary, J.L.

1978-10-01

The character and extent of fuel-cladding chemical interaction (FCCI) was established for UO 2 -25 wt% PuO 2 clad with 20% cold worked Type 316 stainless steel irradiated at high cladding temperatures to peak burnups greater than 8 atom %. The data base consists of 153 data sets from fuel pins irradiated in EBR-II with peak burnups to 9.5 atom %, local cladding inner surface temperatures to 725 0 C, and exposure times to 415 equivalent full power days. As-fabricated oxygen-to-metal ratios (O/M) ranged from 1.938 to 1.984 with the bulk of the data in the range 1.96 to 1.98. HEDL P-15 pins provided data at low heat rates, approx. 200 W/cm, and P-23 series pins provided data at higher heat rates, approx. 400 W/cm. A design practice for breeder reactors is to consider an initial reduction of 50 microns in cladding thickness to compensate for possible FCCI. This approach was considered to be a conservative approximation in the absence of a comprehensive design correlation for extent of interaction. This work provides to the designer a statistically based correlation for depth of FCCI which reflects the influences of the major fuel and operating parameters on FCCI

Advanced Light-Duty SI Engine Fuels Research: Multiple Optical Diagnostics of Well-mixed and Stratified Operation.

Energy Technology Data Exchange (ETDEWEB)

Sjoberg, Carl Magnus Goran [Sandia National Lab. (SNL-CA), Livermore, CA (United States); Vuilleumier, David [Sandia National Lab. (SNL-CA), Livermore, CA (United States)

2018-02-01

Ever tighter fuel economy standards and concerns about energy security motivate efforts to improve engine efficiency and to develop alternative fuels. This project contributes to the science base needed by industry to develop highly efficient direct injection spark ignition (DISI) engines that also beneficially exploit the different properties of alternative fuels. Here, the emphasis is on lean operation, which can provide higher efficiencies than traditional non-dilute stoichiometric operation. Since lean operation can lead to issues with ignition stability, slow flame propagation and low combustion efficiency, the focus is on techniques that can overcome these challenges. Specifically, fuel stratification is used to ensure ignition and completeness of combustion but this technique has soot and NOx emissions challenges. For ultra-lean well-mixed operation, turbulent deflagration can be combined with controlled end-gas autoignition to render mixed-mode combustion for sufficiently fast heat release. However, such mixed-mode combustion requires very stable inflammation, motivating studies on the effects of near-spark flow and turbulence, and the use of small amounts of fuel stratification near the spark plug.

Nuclear power plant types and the management of plutonium and minor actinides - in search of fuel cycle flexibility

International Nuclear Information System (INIS)

Thomas, J.B.

2002-01-01

Transuranics management concerns all NPP types, because of the specifications for sustainable development. Multiple recycling is mandatory. Neutronic abundance can be obtained in fast spectrum, or by adding external neutrons or (temporarily) with additional 235 U. The LWRs can control the plutonium inventory and significantly reduce the amount of transuranics transferred to the geological repository, thanks to the use of innovative nuclear fuel in a limited part of the NPP fleet. HTR adapted to transuranics burning can help. In the future, in addition to the liquid metal FBR, a strategy based on a gas cooled technological line and advanced fuel opens a second path towards fast spectra. Strategies for defining the optimal mix of reactor types in the nuclear fleet at a given time and demonstrating the fuel cycle flexibility are under study. (author)

Fuel exchanger in FBR type reactor

International Nuclear Information System (INIS)

Shinden, Kazuhiko; Tanaka, Osamu.

1990-01-01

The present invention concerns a fuel exchanger for exchanging fuels in an LMFBR type reactor using liquid metals as coolants. An outer gripper cylinder rotating device for rotating an outer gripper cylinder that holds a gripper is driven, to lower the gripper driving portion and the outer gripper cylinder, fuels are caught by the finger at the top end of the outer gripper cylinder and elevated to extract the fuels from the reactor core. Then, the gripper driving portion casing and the outer gripper cylinder are rotated to rotate the fuels caught by the gripper. Subsequently, the gripper driving portion and the outer gripper cylinder are lowered to charge the fuels in the reactor core. This can directly shuffle the fuels in the reactor core without once transferring the fuels into a reactor storing pot and replacing with other fuels, thereby shortening the shuffling time. (I.N.)

Paired replacement fuel assemblies for BWR-type reactor

International Nuclear Information System (INIS)

Oguchi, Kazushige.

1997-01-01

There are disposed a large-diameter water rod constituting a non-boiling region at a central portion and paired replacement fuel assemblies for two streams having the same average enrichment degree and different amount of burnable poisons. The paired replacement fuel assemblies comprise a first fuel assembly having a less amount of burnable poisons and a second fuel assembly having a larger amount of burnable poisons. A number of burnable poison-containing fuel rods in adjacent with the large diameter water rod is increased in the second fuel assembly than the first fuel assembly. Then, the poison of the paired replacement fuel assemblies for the BWR type reactor can be annihilated simultaneously at the final stage of the cycle. Accordingly, fuels for a BWR type reactor excellent in economical property and safety and facilitating the design of the replacement reactor core can be obtained. (N.H.)

Safety criteria related to microheterogeneities in LWR mixed oxide fuels

International Nuclear Information System (INIS)

Renard, A.; Mostin, N.

1978-01-01

The main safety aspets of PuO 2 microheterogeneities in the pellets of LWR mixed oxide fuels are reviewed. Points of interest are studied, especially the transient behaviour in accidental conditions and criteria are deduced for use in the specification and quality control of the fabricated product. (author)

Measurement of PuO2 spot in the mixed-oxide fuels

International Nuclear Information System (INIS)

Yokosuka, Yoshifumi; Kaneda, Kenichiro; Ishikawa, Eiji; Nakajima, Katsuaki; Kashima, Sadamitsu

1974-01-01

The homogeneity of mixed-oxide fuel is evaluated with the maximum diameter of PuO 2 spots. PuO 2 spot distribution is influenced by fuel production process and raw materials used. Analytical measurement of the PuO 2 spot distribution was carried out. α-autoradiograph was taken with nitrocellulose film as a detector. The film was exposed for 15 and 30 minutes to the sample pellets, and etched in 10% NaOH for 2 minutes at 60 0 C. Measurement of PuO 2 spots was performed manually with a projector. PuO 2 spot distribution was measured. There is no difference of spot distribution owing to the location in a pellet. However some discrepancy exists in spot number and maximum diameter between surface and inner part. This phenomenon is explained by considering that the diffusion on surface is more effective than in inner part in sintering process. Three pellets were taken out from a lot at random, and α-autoradiograph was taken longitudinally with a pellet and transversely with the others. PuO 2 spot number counted from three autoradiographs was about 150/cm 2 , and the standard deviation was about 1.5. No difference was observed among them. Concerning the influence of mixing method on PuO 2 spot formation, three methods were examined; ball mill mixing (with and without lining) and Kneader mixing. Kneader mixing gave the worst results in maximum spot diameter and spot distribution range. (Tai, I.)

Present status of uranium-plutonium mixed carbide fuel development for LMFBRs

International Nuclear Information System (INIS)

Handa, Muneo; Suzuki, Yasufumi

1984-01-01

The feature of carbide fuel is that it has the doubling time as short as about 13 years, that is, close to one half as compared with oxide fuel. The development of the carbide fuel in the past 10 years has been started in amazement. Especially in the program of new fuel development in USA started in 1974, He and Na bond fuel attained the burnup of 16 a/o without causing the breaking of cladding tubes. In 1984, the irradiation of the assembly composed of 91 fuel pins in the FFTF is expected. On the other hand in Japan, the fuel research laboratory was constructed in 1974 in the Oarai Laboratory, Japan Atomic Energy Research Institute, to carry out the studies on carbide fuel. In the autumn of 1982, two carbide fuel pins with different chemical composition have been successfully made. Accordingly, the recent status of the development is explained. The uranium-plutonium mixed carbide fuel is suitable to liquid metal-cooled fast breeder reactors because of large heat conductivity and the high density of nuclear fission substances. The thermal and nuclear characteristics of carbide fuel, the features of the reactor core using carbide fuel, the chemical and mechanical interaction of fuel and cladding tubes, the selection of bond materials, the manufacturing techniques for the fuel, the development of the analysis code for fuel behavior, and the research and development of carbide fuel in Japan are described. (Kako, I.)

Fuel assemblies for PWR type reactors: fuel rods, fuel plates. CEA work presentation

International Nuclear Information System (INIS)

Delafosse, Jacques.

1976-01-01

French work on PWR type reactors is reported: basic knowledge on Zr and its alloys and on uranium oxide; experience gained on other programs (fast neutron and heavy water reactors); zircaloy-2 or zircaloy-4 clad UO 2 fuel rods; fuel plates consisting of zircaloy-2 clad UO 2 squares of thickness varying between 2 and 4mm [fr

Studies on the dissolution of mixed oxide spent fuel from FBR

International Nuclear Information System (INIS)

Nemoto, Shin-ichi; Shibata, Atsuhiro; Shioura, Takao; Okamoto, Fumitoshi; Tanaka, Yasumasa

1995-01-01

At the Chemical Processing Facility(CPF) in the Tokai Works of the Power Reactor and Nuclear Fuel Development Corporation(PNC), since 1982 Laboratory scale hot experiments have been carried out on the development of reprocessing technology for FBR mixed oxide fuel. The spent fuel pins which have been used in out experiments were irradiated in Experimental Fast Reactor 'Joyo' Phenix (France) and DFR(UK). Burn-up of the fuel pins were 4,400-100,000 MWd/t. This paper Summarizes a dissolution study that have been performed to define the Key parameters affecting dissolution rate such as concentration of nitric acid, burn-up, and temperature. And this paper also discusses about the character of releasing 85 Kr in chopping and dissolution process, and about the amount of insoluble residue. (author)

Investigation of coolant thermal mixing within 28-element CANDU fuel bundles using the ASSERT-PV thermal hydraulics code

International Nuclear Information System (INIS)

Lightston, M.F.; Rock, R.

1996-01-01

This paper presents the results of a study of the thermal mixing of single-phase coolant in 28-element CANDU fuel bundles under steady-state conditions. The study, which is based on simulations performed using the ASSERT-PV thermal hydraulic code, consists of two main parts. In the first part the various physical mechanisms that contribute to coolant mixing are identified and their impact is isolated via ASSERT-PV simulations. The second part is concerned with development of a preliminary model suitable for use in the fuel and fuel channel code FACTAR to predict the thermal mixing that occurs between flow annuli. (author)

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

International Nuclear Information System (INIS)

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

2016-01-01

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

Cesium relocation in mixed-oxide fuel pins resulting from increased temperature reirradiation

International Nuclear Information System (INIS)

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

1976-06-01

Mixed-oxide fuel pins from EBR-II test subassemblies PNL-3 and PNL-4 were reirradiated in the GETR to study effects of increased fuel and cladding temperatures on chemical and thermomechanical behavior. Radial and axial distributions of cesium were obtained using postirradiation nondestructive precision gamma-scanning techniques. Data presented relate to the dependence of cesium distribution and transport processes on temperature gradients which were altered after substantial steady-state operation

Safety problems related to microheterogeneities in physically mixed oxide fuels

International Nuclear Information System (INIS)

Renard, A.; Evrard, G.; Vanhellemont, G.

1976-01-01

The safety aspects of microheterogeneities in LMFBR mixed oxide fuel are reviewed from the point of view of the pin behaviour dynamic study, the fabrication and the quality control. The paper emphasizes some significant parameters in transient conditions, the prevention means in the fabrication process and the analysis methods for the control

Experimental evaluation of different mixing promoter for nuclear fuel element by means of a new thermal tracing technique

International Nuclear Information System (INIS)

Silin, Nicolas; Juanico, Luis; Delmastro, Dario

2004-01-01

In this work a new experimental method is used to experimentally evaluate the performance of different appendages promoting the turbulent mixing between the coupled subchannels of nuclear fuel elements.The method used will be introduced in another presentation and consists in the generation and measurement of small thermal traces in the refrigerating water flow between the fuel rods.Because it is suitable for heterogeneous and compact subchannels (as Argentinean fuels) with high water flows in simple and affordable tests at atmospheric pressure, this new method is specially well suited for the design of fuel elements, while it offers advantages over other methods of mixing measurement.The experiments carried out on a small test section proved that the buttons brazed to the fuel rods (similar to the 'turbulence promoters' of the Canflex fuel) had an excellent thermohydraulic performance as compared to different mixing vane designs studied.The thermal traces method developed has shown its potential as a thermohydraulic design tool for the development of advanced nuclear fuels, that eventually incorporate mixing promoter elements. In the case of CARA, and as it includes spacer grids, it could be possible to use them to incorporate these elements without the need of brazing them to the rods (as is the case in Canflex), and therefore without penalizing its integrity [es

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.)

Characterization of aerosols from industrial fabrication of mixed-oxide nuclear reactor fuels

International Nuclear Information System (INIS)

Hoover, M.D.; Newton, G.J.

1997-01-01

Recycling plutonium into mixed-oxide (MOX) fuel for nuclear reactors is being given serious consideration as a safe and environmentally sound method of managing plutonium from weapons programs. Planning for the proper design and safe operation of the MOX fuel fabrication facilities can take advantage of studies done in the 1970s, when recycling of plutonium from nuclear fuel was under serious consideration. At that time, it was recognized that the recycle of plutonium and uranium in irradiated fuel could provide a significant energy source and that the use of 239 Pu in light water reactor fuel would reduce the requirements for enriched 235 U as a reactor fuel. It was also recognized that the fabrication of uranium and plutonium reactor fuels would not be risk-free. Despite engineered safety precautions such as the handling of uranium and plutonium in glove-box enclosures, accidental releases of radioactive aerosols from normal containment might occur. Workers might then be exposed to the released materials by inhalation

CANDU type fuel behavior evaluation - a probabilistic approach

International Nuclear Information System (INIS)

Moscalu, D.R.; Horhoianu, G.; Popescu, I.A.; Olteanu, G.

1995-01-01

In order to realistically assess the behavior of the fuel elements during in-reactor operation, probabilistic methods have recently been introduced in the analysis of fuel performance. The present paper summarizes the achievements in this field at the Institute for Nuclear Research (INR), pointing out some advantages of the utilized method in the evaluation of CANDU type fuel behavior in steady state conditions. The Response Surface Method (RSM) has been selected for the investigation of the effects of the variability in fuel element computer code inputs on the code outputs (fuel element performance parameters). A new developed version of the probabilistic code APMESRA based on RSM is briefly presented. The examples of application include the analysis of the results of an in-reactor fuel element experiment and the investigation of the calculated performance parameter distribution for a new CANDU type extended burnup fuel element design. (author)

Markets and economics of mixed waste paper as a boiler fuel

International Nuclear Information System (INIS)

Lyons, J.K.; Kerstetter, J.D.

1991-01-01

Mixed waste paper (MWP) is the second largest component of the municipal solid waste steam disposal of in Washington State. Recent state legislation has mandated source separation of recycled material including MWP. The quantity collected will soon saturate both domestic and foreign markets. An alternative market could be as a fuel in existing combustors. The use of MWP as a fuel requires environmental and economic acceptance by potential users. MWP was analyzed for heavy metal concentrations and elemental composition and found to be similar to existing solid and fossil fuels burned in existing boilers. Existing regulations, however, may classify MWP as a municipal solid waste, thus increasing the capital and administrative costs of using this fuel. The cost of processing MWP into a fluff and a pellet was determined. Three existing facilities were studied to determine the capital and operating costs for them to use MWP fuel. In all cases, the cost of processing and transporting the fuel was greater than the break-even price that could be paid by the potential users

Mixed waste paper to ethanol fuel. A technology, market, and economic assessment for Washington

Energy Technology Data Exchange (ETDEWEB)

1991-01-01

The objectives of this study were to evaluate the use of mixed waste paper for the production of ethanol fuels and to review the available conversion technologies, and assess developmental status, current and future cost of production and economics, and the market potential. This report is based on the results of literature reviews, telephone conversations, and interviews. Mixed waste paper samples from residential and commercial recycling programs and pulp mill sludge provided by Weyerhauser were analyzed to determine the potential ethanol yields. The markets for ethanol fuel and the economics of converting paper into ethanol were investigated.

Evaluation of utilizing spent fuel and plutonium by optimization model for nuclear fuel cycle

International Nuclear Information System (INIS)

Yoshida, Naoto; Fujii, Yasumasa; Komiyama, Ryoichi

2016-01-01

The nuclear power generation has played an important role in power generation mix as a base load power supply. On the other hand, increasing spent fuel and separated plutonium is a long-standing problem. It is expected that advanced fast reactor and high temperature gas reactor could reduce nuclear waste and effectively consume it as valuable resources. Specific scenarios about spent fuel and the gross weight of plutonium are assumed in this study, and the installable potential of fuel cycle and the most suitable reactor mix are analyzed. The model is formulated as liner programing. The model identifies the best strategy of mix of nuclear reactor types to minimize the present value of total cost in a forecast period. As a result, Fast Breeder Reactor and High Temperature Gas Reactor reduce stored spent fuel and increase the consumptions of plutonium. (author)

Spallation as a dominant source of pusher-fuel and hot-spot mix in inertial confinement fusion capsules

Science.gov (United States)

Orth, Charles D.

2016-02-01

We suggest that a potentially dominant but previously neglected source of pusher-fuel and hot-spot "mix" may have been the main degradation mechanism for fusion energy yields of modern inertial confinement fusion (ICF) capsules designed and fielded to achieve high yields—not hydrodynamic instabilities. This potentially dominant mix source is the spallation of small chunks or "grains" of pusher material into the fuel regions whenever (1) the solid material adjacent to the fuel changes its phase by nucleation and (2) this solid material spalls under shock loading and sudden decompression. We describe this mix mechanism, support it with simulations and experimental evidence, and explain how to eliminate it and thereby allow higher yields for ICF capsules and possibly ignition at the National Ignition Facility.

Effect of aviation fuel type and fuel injection conditions on the spray characteristics of pressure swirl and hybrid air blast fuel injectors

Science.gov (United States)

Feddema, Rick

Feddema, Rick T. M.S.M.E., Purdue University, December 2013. Effect of Aviation Fuel Type and Fuel Injection Conditions on the Spray Characteristics of Pressure Swirl and Hybrid Air Blast Fuel Injectors. Major Professor: Dr. Paul E. Sojka, School of Mechanical Engineering Spray performance of pressure swirl and hybrid air blast fuel injectors are central to combustion stability, combustor heat management, and pollutant formation in aviation gas turbine engines. Next generation aviation gas turbine engines will optimize spray atomization characteristics of the fuel injector in order to achieve engine efficiency and emissions requirements. Fuel injector spray atomization performance is affected by the type of fuel injector, fuel liquid properties, fuel injection pressure, fuel injection temperature, and ambient pressure. Performance of pressure swirl atomizer and hybrid air blast nozzle type fuel injectors are compared in this study. Aviation jet fuels, JP-8, Jet A, JP-5, and JP-10 and their effect on fuel injector performance is investigated. Fuel injector set conditions involving fuel injector pressure, fuel temperature and ambient pressure are varied in order to compare each fuel type. One objective of this thesis is to contribute spray patternation measurements to the body of existing drop size data in the literature. Fuel droplet size tends to increase with decreasing fuel injection pressure, decreasing fuel injection temperature and increasing ambient injection pressure. The differences between fuel types at particular set conditions occur due to differences in liquid properties between fuels. Liquid viscosity and surface tension are identified to be fuel-specific properties that affect the drop size of the fuel. An open aspect of current research that this paper addresses is how much the type of aviation jet fuel affects spray atomization characteristics. Conventional aviation fuel specifications are becoming more important with new interest in alternative

Metabolic fuels: regulating fluxes to select mix.

Science.gov (United States)

Weber, Jean-Michel

2011-01-15

Animals must regulate the fluxes of multiple fuels to support changing metabolic rates that result from variation in physiological circumstances. The aim of fuel selection strategies is to exploit the advantages of individual substrates while minimizing the impact of disadvantages. All exercising mammals share a general pattern of fuel selection: at the same %V(O(2,max)) they oxidize the same ratio of lipids to carbohydrates. However, highly aerobic species rely more on intramuscular fuels because energy supply from the circulation is constrained by trans-sarcolemmal transfer. Fuel selection is performed by recruiting different muscles, different fibers within the same muscles or different pathways within the same fibers. Electromyographic analyses show that shivering humans can modulate carbohydrate oxidation either through the selective recruitment of type II fibers within the same muscles or by regulating pathway recruitment within type I fibers. The selection patterns of shivering and exercise are different: at the same %V(O(2,max)), a muscle producing only heat (shivering) or significant movement (exercise) strikes a different balance between lipid and carbohydrate oxidation. Long-distance migrants provide an excellent model to characterize how to increase maximal substrate fluxes. High lipid fluxes are achieved through the coordinated upregulation of mobilization, transport and oxidation by activating enzymes, lipid-solubilizing proteins and membrane transporters. These endurance athletes support record lipolytic rates in adipocytes, use lipoprotein shuttles to accelerate transport and show increased capacity for lipid oxidation in muscle mitochondria. Some migrant birds use dietary omega-3 fatty acids as performance-enhancing agents to boost their ability to process lipids. These dietary fatty acids become incorporated in membrane phospholipids and bind to peroxisome proliferator-activated receptors to activate membrane proteins and modify gene expression.

Investigations on burning efficiency and exhaust emission of in-line type emulsified fuel system

Energy Technology Data Exchange (ETDEWEB)

Tseng, Y.K. [National Chinyi University of Technology (Taiwan). Dept. of Mechanical Engineering; Cheng, H.C. [Point Environmental Protection Technology Company Limited (Taiwan)

2011-07-28

In this research, the burning efficiency as well as exhaust emission of a new water-in-oil emulsified fuel system was studied. This emulsified system contains two core processes, the first one is to mix 97% water with 3% emulsifier by volume, and get the milk-like emulsified liquid, while the second one is to compound the milk-like emulsified liquid with heavy oil then obtain the emulsified fuel. In order to overcome the used demulsification problem during in reserve or in transport, this system was designed as a made and use in-line type. From the results of a series of burning tests, the fuel saving can be 8--15%. Also, from the comparison of decline for the heat value and total energy output of emulsified fuel, one can find that the water as the dispersed phase in the combustion process will lead to a micro-explosion as well as the water gas effect, both can raise the combustion temperature and burning efficiency. By comparing the waste gas emission of different types of emulsified fuel, one can know that, the CO2 emission reduces approximately 14%, and NOx emission reduces above 46%, meaning the reduction of the exhaust gas is truly effective. From the exhaust temperature of tail pipe, the waste heat discharge also may reduce 27%, it is quite advantageous to the global warming as well as earth environmental protection.

Energy use and environmental impact of new alternative fuel mix in electricity generation in Malaysia

International Nuclear Information System (INIS)

Al-Amin, A.Q.; Siwar, C.; Jaafar, A.H.

2009-01-01

The Government of Malaysia introduced a five-fuel diversification strategy in 1999 to ensure security of energy supply. This strategy will continue until 2020 to reduce Malaysia's dependence on fossil fuels for generating electricity. This paper empirically explored the economic impact of electricity generation and scenario analysis that separately identifies impact on the environment of coal, fuel and hydro generating electricity technologies. It also evaluated emissions of carbon dioxide, sulphur dioxide and nitrogen oxide for the year 1991 and 2000 based on business as usual techniques and projection of those emissions based on business as usual and fuel mix strategy as specified in the fuel diversification strategy. The strategy in the electricity sector aims for a gradual change in fuel use from 74.9 per cent natural gas, 9.7 per cent coal, 10.4 per cent hydro, and 5 per cent petroleum in the year 2000 to 40 per cent natural gas, 30 per cent hydro, 29 per cent coal, and only 1 per cent petroleum by the year 2020. This paper presented the underlying model which is based on input-output techniques. The pollution emission levels from the fossil fuels were estimated. The study revealed that the fuel mix envisioned by the Fuel Diversification Strategy, designed to reduce Malaysia's dependence on fuel oil and increase its energy security would result in an increase in undesired emissions. 16 refs., 5 tabs., 3 figs

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

International Nuclear Information System (INIS)

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

1992-01-01

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

Effects of Fuel Type and Fuel Delivery System on Pollutant Emissions of Pride and Samand Vehicles

Directory of Open Access Journals (Sweden)

Akbar Sarhadi

2017-04-01

Full Text Available This research was aimed to study the effect of the type of fuel delivery system (petrol, dedicated or bifuel, the type of consumed fuel (petrol or gas, the portion of consumed fuel and also the duration of dual-fuelling in producing carbon monoxide, carbon dioxide and unburned hydrocarbons from Pride and Samand. According to research objectives, data gathering from 2000 vehicles has been done by visiting Hafiz Vehicle Inspection Center every day for 2 months. The results of this survey indicated that although there is no significant difference between various fuel delivery systems in terms of producing the carbon monoxide, carbon dioxide and unburned hydrocarbons by Samand, considering the emission amount of carbon dioxide, the engine performance of Pride in bifuel and dedicated state in GTXI and 132 types is more unsatisfactory than that of petrol state by 0.3 and 0.4%, respectively. On the other hand, consuming natural gas increases the amount of carbon monoxide emission in dual- fuel Pride by 0.18% and decreases that in dual-fuel Samand by 1.2%, which signifies the better design of Samand in terms of fuel pumps, used kit type and other engine parts to use this alternative fuel compared to Pride. Since the portion of consumed fuel and also duration of dual-fuelling does not have a significant effect on the amount of output pollutants from the studied vehicles, it can be claimed that the output substances from the vehicle exhaust are more related to the vehicle’s condition than the fuel type.

Effect of 17 x 17 fuel assembly geometry on interchannel thermal mixing

International Nuclear Information System (INIS)

Motley, F.E.; Wenzell, A.H.; Cadek, F.F.

1975-01-01

A test to determine the value of the thermal diffusion coefficient (TDC) in the 17 x 17 fuel assembly geometry was conducted. The test section was a 5 x 5 rod bundle with a radial power difference of 4.5 to 1. The rod OD and pitch are identical to the 17 x 17 fuel assembly, as is the mixing vane grid design. The value of thermal diffusion coefficient (TDC) was determined by matching the experimental exit enthalpy distribution to that predicted by the THINC computer code. The mean value of TDC for the 17 x 17 fuel assembly geometry is TDC = .059. 6 references

Neutronic feasibility of PWR core with mixed oxide fuels in the Republic of Korea

International Nuclear Information System (INIS)

Kim, Y.J.; Joo, H.K.; Jung, H.G.; Sohn, D.S.

1997-01-01

Neutronic feasibility of a PWR core with mixed oxide (MOX) fuels has been investigated as part of the feasibility study for recycling spent fuels in Korea. A typical 3-loop PWR with 900 MWe capacity is selected as reference plant to develop equilibrium core designs with low-leakage fuel management scheme, while incorporating various MOX loading. The fuel management analyses and limited safety analyses show that, safely stated, MOX recycling with 1/3 reload fraction can be accommodated for both annual and 18 month fuel cycle schemes in Korean PWRs, without major design modifications on the reactor systems. (author). 12 refs, 4 figs, 3 tabs

Enhanced air/fuel mixing for automotive stirling engine turbulator-type combustors

Science.gov (United States)

Riecke, George T.; Stotts, Robert E.

1992-01-01

The invention relates to the improved combustion of fuel in a combustion chamber of a stirling engine and the like by dividing combustion into primary and secondary combustion zones through the use of a diverter plate.

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

International Nuclear Information System (INIS)

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

1992-01-01

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

Oxygen-to-metal ratio control during fabrication of mixed oxide fast breeder reactor fuel pellets

International Nuclear Information System (INIS)

Rasmussen, D.E.; Benecke, M.W.; Jentzen, W.R.; McCord, R.B.

1979-05-01

Oxygen-to-metal ratio (O/M) of mixed oxide fuel pellets can be controlled during fabrication by proper selection of binder (type and content) and sintering conditions. Sintering condition adjustments involved the passing of Ar--8% H 2 sintering gas across a cryostat ice bath controlled to temperatures ranging from -5 to -60 0 C to control as-sintered pellet O/M ratio. As-sintered fuel pellet O/M decreased with increasing Sterotex binder and PuO 2 concentrations, increasing sintering temperature, and decreasing sintering gas dew point. Approximate relationships between Sterotex binder level and O/M were established for PuO 2 --UO 2 and PuO 2 --ThO 2 fuels. O/M was relatively insensitive to Carbowax binder concentration. Several methods of increasing O/M using post-sintering pellet heat treatments were demonstrated, with the most reliable being a two-step process of first raising the O/M to 2.00 (stoichiometric) at 650 0 C in Ar--8% H 2 bubbled through H 2 O, followed by hydrogen reduction to specification O/M in oxygen-gettered Ar-8% H 2 at temperatures ranging from 1200 to 1690 0 C

Fuel element burnup determination in HEU-LEU mixed TRIGA research reactor core

International Nuclear Information System (INIS)

Zagar, Tomaz; Ravnik, Matjaz

2000-01-01

This paper presents the results of a burnup calculations and burnup measurements for TRIGA FLIP HEU fuel elements and standard TRIGA LEU fuel elements used simultaneously in small TRIGA Mark II research reactor in Ljubljana, Slovenija. The fuel element burnup for approximately 15 years of operation was calculated with two different in house computer codes TRIGAP and TRIGLAV (both codes are available at OECD NEA Data Bank). The calculation is performed in one-dimensional radial geometry in TRIGAP and in two-dimensional (r,φ) geometry in TRIGLAV. Inter-comparison of results shows important influence of in-core water gaps, irradiation channels and mixed rings on burnup calculation accuracy. Burnup of 5 HEU and 27 LEU fuel elements was also measured with reactivity method. Measured and calculated burnup values are inter-compared for these elements (author)

Expansion of lymph node metastasis in mixed-type submucosal invasive gastric cancer.

Science.gov (United States)

Mikami, Koji; Hirano, Yukiko; Futami, Kitaro; Maekawa, Takafumi

2017-07-18

Mixed-type early gastric cancer (differentiated and undifferentiated components) incurs a higher risk of lymph node metastasis than pure-type early gastric cancer (only differentiated or only undifferentiated components). Therefore, we investigated the expansion of lymph node metastasis in mixed-type submucosal invasive gastric cancer in order to establish the most appropriate treatment for mixed-type cancer. We retrospectively analyzed 279 consecutive patients with submucosal invasive gastric cancer who underwent curative gastrectomy for gastric cancer between 1996 and 2015. We classified the patients into the mixed-type and pure-type groups according to histologic examination and evaluated the expansion of lymph node metastasis. The rate of lymph node metastasis was 23.7% (66/279) in the total patients, 36.4% (36/99) in the mixed-type group, and 16.6% (30/180) in the pure-type group. The significant independent risk factors for lymph node metastasis were tumor size ≥2.0 cm (P = 0.014), mixed-type gastric cancer (P mixed-type group. The rates of no. 7 lymph node metastasis in the total patients and mixed-type group were 2.9% (8/279) and 5.1% (5/99), respectively; the rates of no. 8a lymph node metastasis were 1.4% (4/279) and 4.0% (4/99), respectively. Mixed histological type is an independent risk factor for lymph node metastasis. Lymph node metastasis in mixed-type gastric cancer involves expansion to the no. 7 and no. 8a lymph nodes. Therefore, lymphadenectomy for mixed-type submucosal invasive gastric cancer requires D1+ or D2 dissection. Copyright © 2017. Published by Elsevier Taiwan.

Fueling method in LMFBR type reactors

International Nuclear Information System (INIS)

Kawashima, Katsuyuki; Inoue, Kotaro.

1985-01-01

Purpose: To extend the burning cycle and decrease the number of fuel exchange batches without increasing the excess reactivity at the initial stage of burning cycles upon fuel loading to an LMFBR type reactor. Method: Each of the burning cycles is divided into a plurality of burning sections. Fuels are charged at the first burning section in each of the cycles such that driver fuel assemblies and blanket assemblies or those assemblies containing neutron absorbers such as boron are distributed in mixture in the reactor core region. At the final stage of the first burning section, the blanket assemblies or neutron absorber-containing assemblies present in mixture are partially or entirely replaced with driver fuel assemblies depending on the number of burning sections such that all of them are replaced with the driver fuel assemblies till the start of the final burning section of the abovementioned cycle. The object of this invention can thus be attained. (Horiuchi, T.)

Investigations of fuel cladding chemical interaction in irradiated LMFBR type oxide fuel pins

International Nuclear Information System (INIS)

Roake, W.E.; Adamson, M.G.; Hilbert, R.F.; Langer, S.

1977-01-01

Understanding and controlling the chemical attack of fuel pin cladding by fuel and fission products are major objectives of the U.S. LMFBR Mixed Oxide Irradiation Testing Program. Fuel-cladding chemical interaction (FCCI) has been recognized as an important factor in the ability to achieve goal peak burnups of 8% (80.MWd/kg) in FFTF and in excess of 10% (100.MWd/kg) in the LMFBR demonstration reactors while maintaining coolant bulk outlet temperatures up to ∼60 deg. C (1100 deg. F). In this paper we review pertinent parts of the irradiation program and describe recent observation of FCCI in the fuel pins of this program. One goal of the FCCI investigations is to obtain a sufficiently quantitative understanding of FCCI such that correlations can be developed relating loss of effective cladding thickness to irradiation and fuel pin fabrication parameters. Wastage correlations being developed using different approaches are discussed. Much of the early data on FCCI obtained in the U.S. Mixed Oxide Fuel Program came from capsule tests irradiated in both fast and thermal flux facilities. The fast flux irradiated encapsulated fuel pins continue to provide valuable data and insight into FCCI. Currently, however, bare pins with prototypic fuels and cladding irradiated in the fast flux Experimental Breeder Reactor-II (EBR-II) as multiple pin assemblies under prototypic powers, temperatures and thermal gradients are providing growing quantities of data on FCCI characteristics and cladding thickness losses from FCCI. A few special encapsulated fuel pin tests are being conducted in the General Electric Test Reactor (GETR) and EBR-II, but these are aimed at providing specific information under irradiation conditions not achievable in the fast flux bare pin assemblies or because EBR-II Operation or Safety requirements dictate that the pins be encapsulated. The discussion in this paper is limited to fast flux irradiation test results from encapsulated pins and multiple pin

Investigations of fuel cladding chemical interaction in irradiated LMFBR type oxide fuel pins

Energy Technology Data Exchange (ETDEWEB)

Roake, W E [Westinghouse-Hanford Co., Richland, WA (United States); Adamson, M G [General Electric Company, Vallecitos Nuclear Center, Pleasanton, CA (United States); Hilbert, R F; Langer, S

1977-04-01

Understanding and controlling the chemical attack of fuel pin cladding by fuel and fission products are major objectives of the U.S. LMFBR Mixed Oxide Irradiation Testing Program. Fuel-cladding chemical interaction (FCCI) has been recognized as an important factor in the ability to achieve goal peak burnups of 8% (80.MWd/kg) in FFTF and in excess of 10% (100.MWd/kg) in the LMFBR demonstration reactors while maintaining coolant bulk outlet temperatures up to {approx}60 deg. C (1100 deg. F). In this paper we review pertinent parts of the irradiation program and describe recent observation of FCCI in the fuel pins of this program. One goal of the FCCI investigations is to obtain a sufficiently quantitative understanding of FCCI such that correlations can be developed relating loss of effective cladding thickness to irradiation and fuel pin fabrication parameters. Wastage correlations being developed using different approaches are discussed. Much of the early data on FCCI obtained in the U.S. Mixed Oxide Fuel Program came from capsule tests irradiated in both fast and thermal flux facilities. The fast flux irradiated encapsulated fuel pins continue to provide valuable data and insight into FCCI. Currently, however, bare pins with prototypic fuels and cladding irradiated in the fast flux Experimental Breeder Reactor-II (EBR-II) as multiple pin assemblies under prototypic powers, temperatures and thermal gradients are providing growing quantities of data on FCCI characteristics and cladding thickness losses from FCCI. A few special encapsulated fuel pin tests are being conducted in the General Electric Test Reactor (GETR) and EBR-II, but these are aimed at providing specific information under irradiation conditions not achievable in the fast flux bare pin assemblies or because EBR-II Operation or Safety requirements dictate that the pins be encapsulated. The discussion in this paper is limited to fast flux irradiation test results from encapsulated pins and multiple pin

Stepwise evolution of fuel assembly design toward a sustainable fuel cycle with hard neutron spectrum light water reactors

International Nuclear Information System (INIS)

Uchikawa, Sadao; Okubo, Tsutomu; Nakano, Yoshihiro

2011-01-01

An advanced LWR with hard neutron spectrum, FLWR, aims at efficient and flexible utilization of nuclear resources by evolving its fuel assembly design keeping the same core configuration. A proposed evolution process of the design toward a sustainable fuel cycle is composed of three stages, the first one based on the LWR fuel cycle infrastructures, the second one for transitioning from the LWR fuel cycle to the FR fuel cycle, and the third one based on the FR fuel cycle infrastructures. For the first stage, a fuel assembly design concept named FLWR/MIX has been developed in which enriched UO 2 fuel rods are arranged in the peripheral region of the assembly, surrounding the MOX fuel rods in the central region. The FLWR/MIX design realizes a breeder type operation under the framework of the LWR-MOX technologies and there experience. A modified FLWR/MIX design with low Pu inventory for the second stage has a potential of high Puf conversion ratio of 1.1 and can contribute to smooth and speedy transition from the LWR fuel cycle to the FR fuel cycle. For the third stage, the FLWR/MIX design is extended into a design with natural UO 2 fuel rods to realize multiple Pu recycling keeping a Puf conversion ratio of around 1.0. (author)

Fuel fragmentation model advances using TEXAS-V

Energy Technology Data Exchange (ETDEWEB)

Corradini, M.L.; El-Beshbeeshy, M.; Nilsuwankowsit, S.; Tang, J. [Wisconsin Univ., Madison, WI (United States). Dept. of Nuclear Engineering and Engineering Physics

1998-01-01

Because an energetic fuel-coolant interaction may be a safety hazard, experiments are being conducted to investigate the fuel-coolant mixing/quenching process (FARO) as well as the energetics of vapor explosion propagation for high temperature fuel melt simulants (KROTOS, WFCI, ZrEX). In both types of experiments, the dynamic breakup of the fuel is one of the key aspects that must be fundamentally understood to better estimate the magnitude of the mixing/quenching process or the explosion energetics. To aid our understanding the TEXAS fuel-coolant interaction computer model has been developed and is being used to analyze these experiments. Recently, the models for dynamic fuel fragmentation during the mixing and explosion phases of the FCI have been improved by further insights into these processes. The purpose of this paper is to describe these enhancements and to demonstrate their improvements by analysis of particular JRC FCI data. (author)

Technical report: fabrication of PWR type rodlet fuel

International Nuclear Information System (INIS)

Yanagisawa, Kazuaki; Uno, Hisao; Sasajima, Hideo

1990-06-01

With respect to the simulated reactivity initiated accident (RIA) experiments with pre-irradiated LWR type fuel rods at nuclear safety research reactor (NSRR), there were principally three technical difficulties which should be overcome: (1) Fabrication of the rodlet fuel; Fuel rods from the commercial power reactors had an active column length by 3.6m. To utilize this for NSRR pulse experiment, rodlet fuel having an active column length by 0.12m (reduced to one thirtieth) is requested to fabricate without changing the inside fuel conditions. (2) Development of in-core instrumentations: During pre-irradiation stages, a long-sized fuel rod had dimensional changes by waterside corrosion, bowing, creep down and so on. The fuel also had greater amount of radioactive fission products. This condition is significant to in-core instrumentations to be attached to the fuel rods. Well characterized data to be obtained from these, however, are quite necessary and important from research point of view. Remote handling techniques to attach the rod pressure sensor, the cladding extensometer, the fuel extensometer, and the cladding surface thermocouple to pre-irradiated fuel rods are, therefore, requested to develop. (3) Installation of PIE equipments for pulsed rodlet fuels: PIE on the pulsed rodlet fuels are necessary to better understanding the fuel performance detaily. Equipments which can easily detect the data related to PCMI type fuel failure are matter of concern. Since 1986, the technical difficulties have been tried to overcome by all staffs belonging to Reactivity Accident Laboratory, NSRR Operation Division, Department of Reactor Fuel Examination and Hot Laboratory. This report describes the technical achievements obtained through four years work. (author)

Fuel and core design study of the sodium-cooled fast reactors. Studies on metallic fuel cores in the JFY2002

International Nuclear Information System (INIS)

Sugino, Kazuteru; Mizuno, Tomoyasu

2003-06-01

Based on the results obtained in the former feasibility study, the metallic fueled core of ordinary-type, that is, 2-region homogeneous core, has been established aiming at the improvement in the core performance, and subsequent comparison has been performed with the mixed oxide fueled core. Further, the attractive concept of the metallic fueled core of high outlet temperature has been constructed which has good nuclear features as a metallic fueled core and has identical outlet temperature to mixed oxide fuelled core. Following items have been found as a result of the investigation on the ordinary-type core. The metallic fueled core whose maximum fast neutron fluence (En>0.1MeV) is set identical (5x10 23 n/cm 2 ) to the mixed oxide fueled cores with core discharge burnup 150GWd/t has sufficient core performances as a metallic fueled core, e.g. higher breeding ratio and longer operation period compared with mixed oxide fueled cores, but the core discharge burnup is limited up to 100GWd/t. However effective discharge burnup including the contribution of the blanket region is comparative to mixed oxide cores under the same breeding ratio condition. In order to enlarge the core discharge burnup to 150GWd/t keeping the core performance identical to above mentioned core's, the irradiation deformation of structural material should be reduced to that of mixed oxide fueled cores. Further the maximum fast neutron fluence reaches to 7-8x10 23 n/cm 2 (En>0.1MeV). The investigations on the core of high outlet temperature have clarified following items. Even in the change of core regions by pin-diameter form 3-region to 2-region and in the limited maximum fuel pin diameter 8.5 mm, realization of the identical outlet/inlet temperatures to the mixed oxide cores (550/395degC) is feasible under the criteria of the maximum temperature 650degC at the inner surface of the cladding. The constructed core accommodates the targets of breeding ratio from about 1.0 to 1.2 only by adjusting

The feasibility study on fuel types for the KALIMER

International Nuclear Information System (INIS)

Hwang, W.; Nam, C.; Yim, J. S.; Na, B. C.; Hahn, D. H.; Kim, Y. I.; Kim, Y. C.; Park, C. K.

1997-08-01

The economics of LMR is largely dependent on the construction cost of the power plant, and the fuel cycle options usually constitute 20 to 30 % of total electricity generation cost. The choice of fuel cycle technology and the fuel type is important in order to develop a LMR with better economics, performance and safety. The LMR fuel types, whose performances have been proven up to 15 at% burnup, are MOX and IFR metal fuel. The base alloy, binary (U-10% Zr) metal fuel with HT9 is used as structural materials of KALIMER. The design concept of KALIMER fuel has been established through the investigation of technical feasibilities on the fuel and recycle systems for MOX and IFR metal fuel. According to the results of comparative analysis for MOX and metal fuel, metal fuel is better than MOX in view of safety, in-reactor performance, nuclear characteristics, economics and non-proliferation, while MOX fuels have advantages in the developmental status and technical cooperation potential. The overall performance of binary (U-10% Zr) metal fuel with HT9 cladding, which is a potential start-up fuel for KALIMER, is not only superior to that of MOX fuel, but also has enough technical feasibility in its high-burnup performance, safety and economics. (author). 54 ref., 13 tabs., 20 figs

The underwater coincidence counter for plutonium measurements in mixed-oxide fuel assemblies manual

International Nuclear Information System (INIS)

Eccleston, G.W.; Menlove, H.O.; Abhold, M.; Baker, M.; Pecos, J.

1999-01-01

This manual describes the Underwater Coincidence Counter (UWCC) that has been designed for the measurement of plutonium in mixed-oxide (MOX) fuel assemblies prior to irradiation. The UWCC uses high-efficiency 3 He neutron detectors to measure the spontaneous-fission and induced-fission rates in the fuel assembly. Measurements can be made on MOX fuel assemblies in air or underwater. The neutron counting rate is analyzed for singles, doubles, and triples time correlations to determine the 240 Pu effective mass per unit length of the fuel assembly. The system can verify the plutonium loading per unit length to a precision of less than 1% in a measurement time of 2 to 3 minutes. System design, components, performance tests, and operational characteristics are described in this manual

Review of direct electrical heating experiments on irradiated mixed-oxide fuel

International Nuclear Information System (INIS)

Fenske, G.R.; Bandyopadhyay, G.

1982-01-01

Results of approximately 50 out-of-reactor experiments that simulated various stages of a loss-of-flow event with irradiated fuel are presented. The tests, which utilized the direct electrical heating technique to simulate nuclear heating, were performed either on fuel segments with their original cladding intact or on fuel segments that were extruded into quartz tubes. The test results demonstrated that the macro- and microscopic fuel behavior was dependent on a number of variables including fuel heating rate, thermal history prior to a transient, the number of heating cycles, type of cladding (quartz vs stainless steel), and fuel burnup

Mixed waste paper as a fuel

International Nuclear Information System (INIS)

Kersletter, J.D.; Lyons, J.K.

1991-01-01

A successful recycling program requires several components: education and promotion, convenient collection service, and most importantly, a market for collected materials. In Washington state, domestic markets currently have, or are building, the capacity to use most of the glass, newsprint, aluminum, tin cans, and corrugated materials that are collected. Unfortunately, markets for mixed waste paper (MWP), a major component of the state's solid waste stream, have been slow to develop and are unable to absorb the tremendous volumes of material generated. The American Paper Stock Institute classifies MWP as low grade paper such as magazines, books, scrap paper, non-corrugated cardboard (boxboard/chipboard), and construction paper. When viewed as part of a curbside collection program MWP consists primarily of catalogs, binder paper, magazines, brochures, junk mail, cereal boxes, and other household packaging items. A comprehensive analysis of Washington State's solid waste stream showed that during 1988, Washington citizens generated approximately 460,000 tons of mixed waste paper. No small amount, this is equivalent to more than 10% of the total solid waste generated in the state, and is expected to increase. Current projections of MWP generation rates indicated that Washington citizens could discard as much as 960,000 tons of MWP by the year 2010 making it one of the single largest components of the state's solid waste stream. This paper reports on the use of MWP as fuel source

Autoradiographic measurement of Pu distribution in mixed-oxide nuclear fuel

International Nuclear Information System (INIS)

Green, D.R.; Rasmussen, D.E.; Gray, W.H.

1976-09-01

The autoradiographic method described was developed for rapid, economical determination of the Pu distribution and microhomogeneity in mixed oxide fuel. High Pu concentration regions of any size down to 13 microns in diameter can be reproducibly resolved using this method. The new method uses computerized scanning and analysis, and includes automatic self-calibration to virtually elimate variations resulting from photographic film and processing. The speed of this new method allows analysis of enough data to ensure statistical reliability of occurrence frequencies, even for sparse populations of Pu-rich regions with diameters greater than 60 microns. Determination of these occurrence frequencies is an important factor in controlling fuel quality to ensure safe, efficient operation in a Liquid Metal Fast Breeder Reactor

Effect of engine parameters and type of gaseous fuel on the performance of dual-fuel gas diesel engines. A critical review

Energy Technology Data Exchange (ETDEWEB)

Sahoo, B.B. [Centre for Energy, Indian Institute of Technology, Guwahati 781039 (India); Sahoo, N.; Saha, U.K. [Department of Mechanical Engineering, Indian Institute of Technology, Guwahati 781039 (India)

2009-08-15

Petroleum resources are finite and, therefore, search for their alternative non-petroleum fuels for internal combustion engines is continuing all over the world. Moreover gases emitted by petroleum fuel driven vehicles have an adverse effect on the environment and human health. There is universal acceptance of the need to reduce such emissions. Towards this, scientists have proposed various solutions for diesel engines, one of which is the use of gaseous fuels as a supplement for liquid diesel fuel. These engines, which use conventional diesel fuel and gaseous fuel, are referred to as 'dual-fuel engines'. Natural gas and bio-derived gas appear more attractive alternative fuels for dual-fuel engines in view of their friendly environmental nature. In the gas-fumigated dual-fuel engine, the primary fuel is mixed outside the cylinder before it is inducted into the cylinder. A pilot quantity of liquid fuel is injected towards the end of the compression stroke to initiate combustion. When considering a gaseous fuel for use in existing diesel engines, a number of issues which include, the effects of engine operating and design parameters, and type of gaseous fuel, on the performance of the dual-fuel engines, are important. This paper reviews the research on above issues carried out by various scientists in different diesel engines. This paper touches upon performance, combustion and emission characteristics of dual-fuel engines which use natural gas, biogas, producer gas, methane, liquefied petroleum gas, propane, etc. as gaseous fuel. It reveals that 'dual-fuel concept' is a promising technique for controlling both NO{sub x} and soot emissions even on existing diesel engine. But, HC, CO emissions and 'bsfc' are higher for part load gas diesel engine operations. Thermal efficiency of dual-fuel engines improve either with increased engine speed, or with advanced injection timings, or with increased amount of pilot fuel. The ignition

Predicting surface fuel models and fuel metrics using lidar and CIR imagery in a dense mixed conifer forest

Science.gov (United States)

Marek K. Jakubowksi; Qinghua Guo; Brandon Collins; Scott Stephens; Maggi. Kelly

2013-01-01

We compared the ability of several classification and regression algorithms to predict forest stand structure metrics and standard surface fuel models. Our study area spans a dense, topographically complex Sierra Nevada mixed-conifer forest. We used clustering, regression trees, and support vector machine algorithms to analyze high density (average 9 pulses/m

Calculation of parameters for inspection planning and evaluation: mixed-oxide fuel fabrication facilities

International Nuclear Information System (INIS)

Reardon, P.T.; Mullen, M.F.

1982-08-01

As part of Task C.35 (Calculation of Parameters for Inspection Planning and Evaluation) of the US Program of Technical Assistance to IAEA Safeguards, Pacific Northwest Laboratory has performed some quantitative analyses of IAEA inspection activities for mixed-oxide fuel fabrication facilities. There were four distinct efforts involved in this task. These were as follows: show the effect on a material balance verification of using two variables measurement methods in some strata; perform additional calculations for the reference facility described in STR-89; modify the INSPECT computer programs to be used as an after-inspection analysis tool, as well as a preinspection planning tool; provide written comments and explantations of text and graphs of the first draft of STR-89, Safeguards Considerations for Mixed-Oxide Fuel Element Fabrication Facilities, by W. Bahm, T. Shea, and D. Tolchenkov, System Studies Section, IAEA

Fuel assemblies for use in nuclear reactors

International Nuclear Information System (INIS)

Mochida, Takaaki.

1987-01-01

Purpose: To increase the plutonium utilization amount and improve the uranium-saving effect in the fuel assemblies of PWR type reactor using mixed uranium-plutonium oxides. Constitution: MOX fuel rods comprising mixed plutonium-uranium oxides are disposed to the outer circumference of a fuel assembly and uranium fuel rods only composed of uranium oxides are disposed to the central portion thereof. In such a fuel assembly, since the uranium fuel rods are present at the periphery of the control rod, the control rod worth is the same as that of the uranium fuel assembly in the prior art. Further, since about 25 % of the entire fuel rods is composed of the MOX fuel rods, the plutonium utilization amount is increased. Further, since the MOX fuel rods at low enrichment degree are present at the outer circumferential portion, mismatching at the boundary to the adjacent MOX fuel assembly is reduced and the problem of local power peaking increase in the MOX fuel assembly is neither present. (Kamimura, M.)

Nuclear fuel

International Nuclear Information System (INIS)

Quinauk, J.P.

1990-01-01

Since 1985, Fragema has been marketing and selling the Advanced Fuel Assemby AFA whose main features are its zircaloy grids and removable top and bottom nozzles. It is this product, which exists for several different fuel assembly arrays and heights, that will be employed in the reactors at Daya Bay. Fragema employs gadolinium as the consumable poison to enable highperformance fuel management. More recently, the company has supplied fuel assemblies of the mixed-oxide(MOX) and enriched reprocessed uranium type. The reliability level of the fuel sold by Fragema is one of the highest in the world, thanks in particular to the excellence of the quality assurance and quality control programs that have been implemented at all stages of its design and manufacture

Experimental advances and preliminary mathematical modeling of the Swiss-roll mixed-reactant direct borohydride fuel cell

Science.gov (United States)

Aziznia, Amin; Oloman, Colin W.; Gyenge, Előd L.

2014-11-01

The Swiss-roll single-cell mixed reactant (SR-MRFC) borohydride - oxygen fuel cell equipped with Pt/carbon cloth 3D anode and either MnO2 or Ag gas-diffusion cathodes is investigated by a combination of experimental studies and preliminary mathematical modeling of the polarization curve. We investigate the effects of four variables: cathode side metallic mesh fluid distributor, separator type (Nafion 112® vs. Viledon®), cathode catalyst (MnO2 vs. Ag), and the hydrophilic pore volume fraction of the gas-diffusion cathode. Using a two-phase feed of alkaline borohydride solution (1 M NaBH4 - 2 M NaOH) and O2 gas in an SR-MRFC equipped with Pt/C 3D anode, MnO2 gas diffusion cathode, Viledon® porous diaphragm, expanded mesh cathode-side fluid distributor, the maximum superficial power density is 2230 W m-2 at 323 K and 105 kPa(abs). The latter superficial power density is almost 3.5 times higher than our previously reported superficial power density for the same catalyst combinations. Furthermore, with a Pt anode and Ag cathode catalyst combination, a superficial power density of 2500 W m-2 is achieved with superior performance durability compared to the MnO2 cathode. The fuel cell results are substantiated by impedance spectroscopy analysis and preliminary mathematical model predictions based on mixed potential theory.

Effect of spacer grid mixing vanes on coolant outlet temperature distribution

Energy Technology Data Exchange (ETDEWEB)

Raemae, Tommi; Lahtinen, Tuukka; Brandt, Tellervo; Toppila, Timo [Fortum Power and Heat, Fortum (Finland). Nuclear Competence Center

2012-08-15

In Loviisa VVER-440-type NPP the coolant outlet temperature of the hot subchannel is constantly monitored during the operation. According to the authority requirement the maximum subchannel outlet temperature must not exceed the saturation temperature. Coolant temperature distribution inside the fuel assembly is affected by the efficiency of the coolant mixing. In order to enhance the coolant mixing the fuel manufacturer is introducing the additional mixing vanes on the fuel bundle spacer grids. In the paper the effect of the different mixing vane modifications is studied with computational fluid dynamics (CFD) simulation. Goal of the modelling is to find vane modifications with which sufficient mixing is reached with acceptable increase in the spacer grid pressure loss. The results of the studies are discussed in the paper. (orig.)

Study of Advanced Reactor Mixed Oxide Fuel Production of (U,Th)O2

International Nuclear Information System (INIS)

Busron-Masduki; Damunir; Pristi-Hartati; R-Sukarsono; Bangun-Wasito

2000-01-01

The high price and starting scarcity of reserved of oil drive the people to drill the alternative nuclear energy. Accelerator-driven Transmutation Waste (ATW) is a prospective technology to solve the problem of used fuel waste, to reduce the anxiety of long term disposal waste, to increase the public acceptance of nuclear energy enter into the third millennium. The future of large nuclear energy appears in many-branched industry will depend on the capability to generate relatively low priced fuel on the basis of commercial nuclear energy. Utilization of uranium-233 -thorium cycle insures long-term fuel supply, makes the nuclear energy production more flexible and enables the self-provision regime to be realized in future. Flowsheet of mixed oxide fuel production for advanced reactor of (U,Th)O 2 is a combination of existing manufacturing equipment and quality assurance program from commercial LWR and HTR. The front-end of flowsheet using sol-gel process. The external sol-gel process is chosen due to simple equipment can anticipate refabrication of U-233 which always contains a few hundred ppm of U-232 and its gamma-emitting daughters, besides yielding smaller waste. The decision to choose external sol-gel process encourages to develop External Gelation Thorium (EGT). In order to get higher density and relatively low compaction pressures (i.e. for advanced LWR) adopted flowsheet EGT is developed to be Sol-Gel Microsphere Pelletization (SGMP). Using the optimal parameters, SGMP become established flowsheet for producing mixed oxide fuel of (U,Th)O 2 for advanced reactor. (author)

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

Thermal conductivity of beginning-of-life uranium-plutonium mixed oxide fuel for fast reactor (Interim report)

International Nuclear Information System (INIS)

Inoue, Masaki; Mizuno, Tomoyasu; Asaga, Takeo

1997-11-01

Thermal conductivity of uranium-plutonium mixed oxide fuel for fast reactor at beginning-of-life was correlated based on the recent results in order to apply to the fuel design and the fuel performance analysis. A number of experimental results of unirradiated fuel specimens were corrected from open literatures and PNC internal reports and examined for the database. In this work two porosity correction factors were needed for high density fuel and low density fuel (around the current Monju specification). The universal porosity correction factor was not determined in this work. In the next step, theoretical and analytical considerations should be taken into account. (J.P.N.)

14 CFR 26.37 - Pending type certification projects: Fuel tank flammability.

Science.gov (United States)

2010-01-01

... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Pending type certification projects: Fuel tank flammability. 26.37 Section 26.37 Aeronautics and Space FEDERAL AVIATION ADMINISTRATION... AIRPLANES Fuel Tank Flammability § 26.37 Pending type certification projects: Fuel tank flammability. (a...

Numerical exploration of mixing and combustion in ethylene fueled scramjet combustor

Science.gov (United States)

Dharavath, Malsur; Manna, P.; Chakraborty, Debasis

2015-12-01

Numerical simulations are performed for full scale scramjet combustor of a hypersonic airbreathing vehicle with ethylene fuel at ground test conditions corresponding to flight Mach number, altitude and stagnation enthalpy of 6.0, 30 km and 1.61 MJ/kg respectively. Three dimensional RANS equations are solved along with species transport equations and SST-kω turbulence model using Commercial CFD software CFX-11. Both nonreacting (with fuel injection) and reacting flow simulations [using a single step global reaction of ethylene-air with combined combustion model (CCM)] are carried out. The computational methodology is first validated against experimental results available in the literature and the performance parameters of full scale combustor in terms of thrust, combustion efficiency and total pressure loss are estimated from the simulation results. Parametric studies are conducted to study the effect of fuel equivalence ratio on the mixing and combustion behavior of the combustor.

Fuel/Air Mixing Characteristics of Strut Injections for Scramjet Combustor Applications (Postprint)

Science.gov (United States)

2008-08-01

regions, and drag will be increased, as suggested by Povinelli .26 Both the total pressure recovery and mixing efficiency for the forward-swept strut are...Experimental Study of Cavity-Strut Combustion in Supersonic Flow,” AIAA Paper 2007-5394, 2007. 26. Povinelli , L.A., “Aerodynamic Drag and Fuel Spreading

Plutonium spot of mixed oxide fuel, 2

International Nuclear Information System (INIS)

Suzuki, Yukio; Maruishi, Yoshihiro; Satoh, Masaichi; Aoki, Toshimasa; Muto, Tadashi

1974-01-01

In a fast reactor, the specification for the homogeneity of plutonium in plutonium-uranium mixed-oxide fuel is mainly dependent on the nuclear characteristics, whereas in a thermal reactor, on thermal characteristics. This homogeneity is measured by autoradiography as the plutonium spot size of the specimens which are arbitrarily chosen fuel pellets from a lot. Although this is a kind of random sampling, it is difficult to apply this method to conventional digital standards including JIS standards. So a special sampling inspection method was studied. First, it is assumed that the shape of plutonium spots is spherical, the size distribution is logarithmic normal, and the standard deviation is constant. Then, if standard deviation and mean spot size are given, the logarithmic normal distribution is decided unitarily, and further if the total weight of plutonium spots for a lot of pellets is known, the number of the spots (No) which does not conform to the specification can be obtained. Then, the fraction defective is defined as No devided by the number of pellets per lot. As to the lot with such fraction defective, the acceptance coefficient of the lot was obtained through calculation, in which the number of sampling, acceptable diameter limit observed and acceptable conditions were used as parameters. (Tai, I.)

Synthesis and characterization of brannerite wasteforms for the immobilization of mixed oxide fuel residues

Energy Technology Data Exchange (ETDEWEB)

Bailey, D.J.; Stennett, M.C.; Hyatt, N.C. [Immobilisation Science Laboratory, Department of Materials Science and Engineering, University of Sheffield, Sheffield, S1 3JD (United Kingdom)

2016-07-01

A possible method for the reduction of civil Pu stockpiles is the reuse of Pu in mixed oxide fuel (MOX). During MOX fuel production, residues unsuitable for further recycle will be produced. Due to their high actinide content MOX residues require immobilization within a robust host matrix. Although it is possible to immobilize actinides in vitreous wasteforms; ceramic phases, such as brannerite (UTi{sub 2}O{sub 6}), are attractive due to their high waste loading capacity and relative insolubility. A range of uranium brannerite, formulated Gd{sub x}U{sub 1-x}Ti{sub 2}O{sub 6}, were prepared using a mixed oxide route. Charge compensation of divalent and trivalent cations was expected to occur via the oxidation of U{sup 4+} to higher valence states (U{sup 5+} or U{sup 6+}). Gd{sup 3+} was added to act as a neutron absorber in the final Pu bearing wasteform. X-ray powder diffraction of synthesised specimens found that phase distribution was strongly affected by processing atmosphere (air or Ar). In all cases prototypical brannerite was formed accompanied by different secondary phases dependent on processing atmosphere. Microstructural analysis (SEM) of the sintered samples confirmed the results of the X-ray powder diffraction. The preliminary results presented here indicate that brannerite is a promising host matrix for mixed oxide fuel residues. (authors)

Analysis of injury types for mixed martial arts athletes.

Science.gov (United States)

Ji, MinJoon

2016-05-01

[Purpose] The purpose of the present study was to examine the types of injuries associated with mixed martial arts and their location in order to provide substantial information to help reduce the risk of these injuries during mixed martial arts. [Subjects and Methods] Data were collected from 455 mixed martial arts athletes who practiced mixed martial arts or who participated in mixed martial arts competitions in the Seoul Metropolitan City and Gyeongnam Province of Korea between June 3, 2015, and November 6, 2015. Questionnaires were used to collect the data. The convenience sampling method was used, based on the non-probability sampling extraction method. [Results] The arm, neck, and head were the most frequent locations of the injuries; and lacerations, concussions, and contusions were the most frequently diagnosed types of injuries in the mixed martial arts athletes in this study. [Conclusion] Reducing the risk of injury by establishing an alert system and preventing critical injuries by incorporating safety measures are important.

Irradiation of mixed UO2-PuO2 oxide samples for fast neutron reactor fuel elements

International Nuclear Information System (INIS)

Mikailoff, H.; Mustelier, J.; Bloch, J.; Conte, M.; Hayet, L.; Lauthier, J.C.; Leclere, J.

1968-01-01

Thermal flux irradiation testings of small mixed oxide pellets UPuO 2 fuel elements were performed in support of the fuel reference design for the Phenix fast reactor. The effects of different parameters (stoichiometry, pellet density, pellet clad gap). on the behaviour of the oxide (temperature distribution, microstructural changes, fission gas release) were investigated in various irradiation conditions. In particular, the effect of fuel density decrease and power rate increase on thermal performances were determined on short term irradiations of porous fuels. (authors) [fr

Evaluation of plate type fuel options for small power reactors

International Nuclear Information System (INIS)

Andrzejewski, Claudio de Sa

2005-01-01

Plate type fuels are generally used in research reactor. The utilization of this kind of configuration improves significantly the overall performance fuel. The conception of new fuels for small power reactors based in plate-type configuration needs a complete review of the safety criteria originally used to conduce power and research reactor projects. In this work, a group of safety criteria is established for the utilization of plate-type fuels in small power reactors taking into consideration the characteristics of power and research reactors. The performance characteristics of fuel elements are strongly supported by its materials properties and the adopted configuration for its fissile particles. The present work makes an orientated bibliographic investigation searching the best material properties (structural materials and fuel compounds) related to the performance fuel. Looking for good parafermionic characteristics and manufacturing exequibility associated to existing facilities in national research centres, this work proposes several alternatives of plate type fuels, considering its utilization in small power reactors: dispersions of UO 2 in stainless steel, of UO 2 in zircaloy, and of U-Mo alloy in zircaloy, and monolithic plates of U-Mo cladded with zircaloy. Given the strong dependency of radiation damage with temperature increase, the safety criteria related to heat transfer were verified for all the alternatives, namely the DNBR; coolant temperature lower than saturation temperature; peak meat temperature to avoid swelling; peak fuel temperature to avoid meat-matrix reaction. It was found that all alternatives meet the safety criteria including the 0.5 mm monolithic U-Mo plate cladded with zircaloy. (author)

Raw mix designing for coal as a fuel in cement kiln as a major fuel and its impact on clinker parameters

International Nuclear Information System (INIS)

Noor-ul-Amin; Ali, K.

2011-01-01

In this paper the use of coal, found at Jabba Taar and Jabba Khushk Khyber Pakhtoon Khwa, in cement manufacturing as a major fuel and its impacts on the raw mix and clinker parameters has been discussed. The maximum amount of coal for pulverizing with raw mix was found to be 10% of the raw mix as calculated from the calorific value and the heat of clinkerization of coal. The coal residue left after burning was utilized in the cement raw material, for which a new raw mix was designed. The new raw mix was converted in to clinker. The resulting clinker was studied for all parameters as per British specification. It was found that the clinker obtained from the newly designed raw mix with coal ash, was in accordance to the British standard specifications. (author)

Biomass conversion to hydrocarbon fuels using the MixAlco™ process at a pilot-plant scale

International Nuclear Information System (INIS)

Taco Vasquez, Sebastian; Dunkleman, John; Chaudhuri, Swades K.; Bond, Austin; Holtzapple, Mark T.

2014-01-01

Texas A and M University has built a MixAlco™ pilot plant that converts biomass to hydrocarbons (i.e., jet fuel, gasoline) using the following steps: fermentation, descumming, dewatering, thermal ketonization, distillation, hydrogenation, and oligomerization. This study describes the pilot plant and reports results from an 11-month production campaign. The focus was to produce sufficient jet fuel to be tested by the U.S. military. Because the scale was relatively small, energy-saving features were not included in the pilot plant. Further, the equipment was operated in a manner to maximize productivity even if yields were low. During the production campaign, a total of 6.015 Mg of shredded paper and 120 kg of chicken manure (dry basis) were fermented to produce 126.5 m 3 of fermentation broth with an average concentration of 12.5 kg m −3 . A total of 1582 kg of carboxylate salts were converted to 587 L of raw ketones, which were distilled and hydrogenated to 470 L of mixed alcohols ranging from C3 to C12. These alcohols, plus 300 L of alcohols made by an industrial partner (Terrabon, Inc.) were shipped to an independent contractor (General Electric) and transformed to jet fuel (∼100 L) and gasoline (∼100 L) byproduct. - Highlights: • We produce hydrocarbons from paper and chicken manure in a pilot-scale production using the MixAlco™ process. • About 100 L of jet fuel were produced for military testing. • High production rates and good product quality were preferred rather than high yields or energy efficiency. • The MixAlco™ process converted successfully lignocellulosic biomass to hydrocarbons and viable for commercial-scale production

Premixer assembly for mixing air and fuel for combustion

Science.gov (United States)

York, William David; Johnson, Thomas Edward; Keener, Christopher Paul

2016-12-13

A premixer assembly for mixing air and fuel for combustion includes a plurality of tubes disposed at a head end of a combustor assembly. Also included is a tube of the plurality of tubes, the tube including an inlet end and an outlet end. Further included is at least one non-circular portion of the tube extending along a length of the tube, the at least one non-circular portion having a non-circular cross-section, and the tube having a substantially constant cross-sectional area along its length

Gasification Characteristics of Coal/Biomass Mixed Fuels

Energy Technology Data Exchange (ETDEWEB)

Mitchell, Reginald [Stanford Univ., CA (United States). Mechanical Engineering Dept.

2014-09-01

A research project was undertaken that had the overall objective of developing the models needed to accurately predict conversion rates of coal/biomass mixtures to synthesis gas under conditions relevant to a commercially-available coal gasification system configured to co-produce electric power as well as chemicals and liquid fuels. In our efforts to accomplish this goal, experiments were performed in an entrained flow reactor in order to produce coal and biomass chars at high heating rates and temperatures, typical of the heating rates and temperatures fuel particles experience in real systems. Mixed chars derived from coal/biomass mixtures containing up to 50% biomass and the chars of the pure coal and biomass components were subjected to a matrix of reactivity tests in a pressurized thermogravimetric analyzer (TGA) in order to obtain data on mass loss rates as functions of gas temperature, pressure and composition as well as to obtain information on the variations in mass specific surface area during char conversion under kinetically-limited conditions. The experimental data were used as targets when determining the unknown parameters in the chemical reactivity and specific surface area models developed. These parameters included rate coefficients for the reactions in the reaction mechanism, enthalpies of formation and absolute entropies of adsorbed species formed on the carbonaceous surfaces, and pore structure coefficients in the model used to describe how the mass specific surface area of the char varies with conversion. So that the reactivity models can be used at high temperatures when mass transport processes impact char conversion rates, Thiele modulus – effectiveness factor relations were also derived for the reaction mechanisms developed. In addition, the reactivity model and a mode of conversion model were combined in a char-particle gasification model that includes the effects of chemical reaction and diffusion of reactive gases through particle

Performance enhancement of a spark ignition engine fed by different fuel types

International Nuclear Information System (INIS)

Hedfi, Hachem; Jbara, Abdessalem; Jedli, Hedi; Slimi, Khalifa; Stoppato, Anna

2016-01-01

Highlights: • Biogas mixed with hydrogen is checked for a spark ignition engine. • An engine fed by biogas, hydrogen, natural gas or liquid petroleum gas is studied. • Efficiency is optimized with respect to consumption and exhaust gas recirculation. • Combustion reaction progress is characterized in real time. - Abstract: A numerical model based on thermodynamic and kinetic analyses has been established in order to evaluate biogas, hydrogen, natural gas or liquid petroleum gas as fuels in a spark ignition engine. For each fuel type, consumption as well as efficiency have been compared to gasoline in order to generate the same engine work (in the range of 0.28–0.43 W h/cycle). It was found that the spark ignition engine can be fed by an equimolar mixture of biogas and hydrogen. Moreover, thermal efficiency has been enhanced with respect to fuel consumption and exhaust gas recirculation (EGR). It was shown that an equimolar mixture between biogas and hydrogen increases the ITE by around 2.2% and decreases the mass consumption by less than 0.01 g/cycle. In addition, the combustion reaction progresses as well as CO and CO_2 emissions have been characterized in real time.

Neutronic and Logistic Proposal for Transmutation of Plutonium from Spent Nuclear Fuel as Mixed-Oxide Fuel in Existing Light Water Reactors

International Nuclear Information System (INIS)

Trellue, Holly R.

2004-01-01

The use of light water reactors (LWRs) for the destruction of plutonium and other actinides [especially those in spent nuclear fuel (SNF)] is being examined worldwide. One possibility for transmutation of this material is the use of mixed-oxide (MOX) fuel, which is a combination of uranium and plutonium oxides. MOX fuel is used in nuclear reactors worldwide, so a large experience base for its use already exists. However, to limit implementation of SNF transmutation to only a fraction of the LWRs in the United States with a reasonable number of license extensions, full cores of MOX fuel probably are required. This paper addresses the logistics associated with using LWRs for this mission and the design issues required for full cores of MOX fuel. Given limited design modifications, this paper shows that neutronic safety conditions can be met for full cores of MOX fuel with up to 8.3 wt% of plutonium

Lateral Flow Field Behavior Downstream of Mixing Vanes In a Simulated Nuclear Fuel Rod Bundle

International Nuclear Information System (INIS)

Conner, Michael E.; Smith, L. David III; Holloway, Mary V.; Beasley, Donald E.

2004-01-01

To assess the fuel assembly performance of PWR nuclear fuel assemblies, average subchannel flow values are used in design analyses. However, for this highly complex flow, it is known that local conditions around fuel rods vary dependent upon the location of the fuel rod in the fuel assembly and upon the support grid design that maintains the fuel rod pitch. To investigate the local flow in a simulated nuclear fuel rod bundle, a testing technique has been employed to measure the lateral flow field in a 5 x 5 rod bundle. Particle Image Velocimetry was used to measure the lateral flow field downstream of a support grid with mixing vanes for four unique subchannels in the 5 x 5 bundle. The dominant lateral flow structures for each subchannel are compared in this paper including the decay of these flow structures. (authors)

R and D activities on CANDU-type fuel in Indonesia

International Nuclear Information System (INIS)

Suripto, A.; Badruzzaman, M.; Latief, A.

1997-01-01

The status of R and D activities in Indonesia with respect of CANDU-type fuel development is presented. The activities have been started since the first feasibility study to introduce nuclear power plants was carried out in 1970s. The early research comprised the in-situ pilot production of yellow-cake in Kalimantan (Borneo) experimental mining site, uranium purification and pellet preparation. This program continued to gain a full support from the Government which culminated in the realisation of the construction by BATAN of a large fuel development laboratory in Serpong, starting from 1984 in co-operation with NIRA Ansaldo of Italy. The laboratory, which is called the Power Reactor Experimental Fuel Element Installation (EFEI) was originally designed as an experimental facility to integrate the acquired domestic R and D results gained so far on the CANDU-type fuel technology and the additional know-how received from NIRA Ansaldo which at that time was engaged, in developing a CANDU-type fuel, called the CIRENE fuel design. In the present days the facility houses the power reactor fuel development activities carried out to build up the national capability on power reactor fuel fabrication technology in anticipation to embark upon the nuclear energy era in the near future. (author)

Safety analysis of MOX fuels by fuel performance code

Energy Technology Data Exchange (ETDEWEB)

Suzuki, Motoe [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2002-12-01

Performance of plutonium rick mixed oxide fuels specified for the Reduced-Moderation Water Reactor (RMWR) has been analysed by modified fuel performance code. Thermodynamic properties of these fuels up to 120 GWd/t burnup have not been measured and estimated using existing uranium fuel models. Fission product release, pressure rise inside fuel rods and mechanical loads of fuel cans due to internal pressure have been preliminarily assessed based on assumed axial power distribution history, which show the integrity of fuel performance. Detailed evaluation of fuel-cladding interactions due to thermal expansion or swelling of fuel pellets due to high burnup will be required for safety analysis of mixed oxide fuels. Thermal conductivity and swelling of plutonium rich mixed oxide fuels shall be taken into consideration. (T. Tanaka)

Evolution of Particle Bed Reactor Fuel

Science.gov (United States)

Jensen, Russell R.; Evans, Robert S.; Husser, Dewayne L.; Kerr, John M.

1994-07-01

To realize the potential performance advantages inherent in a particle bed reactor (PBR) for nuclear thermal propulsion (NTP) applications, high performance particle fuel is required. This fuel must operate safely and without failure at high temperature in high pressure, flowing hydrogen propellant. The mixed mean outlet temperature of the propellant is an important characteristic of PBR performance. This temperature is also a critical parameter for fuel particle design because it dictates the required maximum fuel operating temperature. In this paper, the evolution in PBR fuel form to achieve higher operating temperatures is discussed and the potential thermal performance of the different fuel types is evaluated. It is shown that the optimum fuel type for operation under the demanding conditions in a PBR is a coated, solid carbide particle.

Planning of fuel coal imports using a mixed integer programming method

Energy Technology Data Exchange (ETDEWEB)

Shih, L.H. [National Cheng Kung University, Tainan (Taiwan). Dept. of Mineral and Petroleum Engineering

1997-12-31

In the public utility and commercial fuel industries, commodities from multiple supply sources are sometimes blended before use to reduce costs and assure quality. A typical example of these commodities is the fuel coal used in coal fired power plants. The diversity of the supply sources for these plants makes the planning and scheduling of fuel coal logistics difficult, especially for a power company that has more than one power plant. This study proposes a mixed integer programming model that provides planning and scheduling of coal imports from multiple suppliers for the Taiwan Power Company. The objective is to minimize total inventory cost by minimizing procurement cost, transportation cost and holding cost. Constraints on the system include company procurement policy, power plant demand, harbor unloading capacity, inventory balance equations, blending requirements, and safety stock. An example problem is presented using the central coal logistics system of the Taiwan Power Company to demonstrate the validity of the proposed model.

Planning of fuel coal imports using a mixed integer programming method

International Nuclear Information System (INIS)

Shih, L.H.

1997-01-01

In the public utility and commercial fuel industries, commodities from multiple supply sources are sometimes blended before use to reduce costs and assure quality. A typical example of these commodities is the fuel coal used in coal fired power plants. The diversity of the supply sources for these plants makes the planning and scheduling of fuel coal logistics difficult, especially for a power company that has more than one power plant. This study proposes a mixed integer programming model that provides planning and scheduling of coal imports from multiple suppliers for the Taiwan Power Company. The objective is to minimize total inventory cost by minimizing procurement cost, transportation cost and holding cost. Constraints on the system include company procurement policy, power plant demand, harbor unloading capacity, inventory balance equations, blending requirements, and safety stock. An example problem is presented using the central coal logistics system of the Taiwan Power Company to demonstrate the validity of the proposed model

14 CFR 26.33 - Holders of type certificates: Fuel tank flammability.

Science.gov (United States)

2010-01-01

... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Holders of type certificates: Fuel tank... Tank Flammability § 26.33 Holders of type certificates: Fuel tank flammability. (a) Applicability. This... part 25 of this chapter. (2) Exception. This paragraph (b) does not apply to— (i) Fuel tanks for which...

Impact of Flight Enthalpy, Fuel Simulant, and Chemical Reactions on the Mixing Characteristics of Several Injectors at Hypervelocity Flow Conditions

Science.gov (United States)

Drozda, Tomasz G.; Baurle, Robert A.; Drummond, J. Philip

2016-01-01

The high total temperatures or total enthalpies required to duplicate the high-speed flight conditions in ground experiments often place stringent requirements on the material selection and cooling needs for the test articles and intrusive flow diagnostic equipment. Furthermore, for internal flows, these conditions often complicate the use of nonintrusive diagnostics that need optical access to the test section and interior portions of the flowpath. Because of the technical challenges and increased costs associated with experimentation at high values of total enthalpy, an attempt is often made to reduce it. This is the case for the Enhanced Injection and Mixing Project (EIMP) currently underway in the Arc-Heated Scramjet Test Facility at the NASA Langley Research Center. The EIMP aims to investigate supersonic combustion ramjet (scramjet) fuel injection and mixing physics, improve the understanding of underlying physical processes, and develop enhancement strategies and functional relationships between mixing performance and losses relevant to flight Mach numbers greater than 8. The experiments will consider a "direct-connect" approach and utilize a Mach 6 nozzle to simulate the combustor entrance flow of a scramjet engine. However, while the value of the Mach number is matched to that expected at the combustor entrance in flight, the maximum value of the total enthalpy for these experiments is limited by the thermal-structural limits of the uncooled experimental hardware. Furthermore, the fuel simulant is helium, not hydrogen. The use of "cold" flows and non-reacting mixtures of fuel simulants for mixing experiments is not new and has been extensively utilized as a screening technique for scramjet fuel injectors. In this study, Reynolds-averaged simulations are utilized (RAS) to systematically verify the implicit assumptions used by the EIMP. This is accomplished by first performing RAS of mixing for two injector configurations at planned nominal experimental

Criticality analysis for mixed thorium-uranium fuel in the Angra-2 PWR reactor using KENO-VI

Energy Technology Data Exchange (ETDEWEB)

Wichrowski, Caio C.; Gonçalves, Isadora C.; Oliveira, Claudio L.; Vellozo, Sergio O.; Baptista, Camila O., E-mail: wichrowski@ime.eb.br, E-mail: isadora.goncalves@ime.eb.br, E-mail: d7luiz@yahoo.com.br, E-mail: vellozo@ime.eb.br, E-mail: camila.oliv.baptista@gmail.com [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil). Seção de Engenharia Nuclear

2017-07-01

The increasing energy demand associated to the current sustainability challenges have given the thorium nuclear fuel cycle renewed interest in the scientific community. Studies have focused on energy production in different reactor designs through the fission of uranium 233, the product of thorium fertilization by neutrons. In order to make it possible for near future applications a strategy based on the adaptation of current nuclear reactors for the use of thorium fuels is being considered. In this work, bearing in mind these limitations, a code was used to evaluate the effect on criticality (k{sub inf}) of the mixing of thorium and uranium in different proportions in the fuel of a PWR, the German designed Angra-2 Brazilian reactor in order to scrutinise its behaviour and determine the feasibility of an adapted ThO{sub 2}-UO{sub 2} mixed fuel cycle using current PWR technology. The analysis is performed using the KENO-VI module in the SCALE 6.1 nuclear safety analysis simulation code and the information is taken from the Angra-2 FSAR (Final Security Analysis Report). (author)

EBSD and TEM Characterization of High Burn-up Mixed Oxide Fuel

International Nuclear Information System (INIS)

Teague, Melissa C; Gorman, Brian P.; Miller, Brandon D; King, Jeffrey

2014-01-01

Understanding and studying the irradiation behavior of high burn-up oxide fuel is critical to licensing of future fast breeder reactors. Advancements in experimental techniques and equipment are allowing for new insights into previously irradiated samples. In this work dual column focused ion beam (FIB)/scanning electron microscope (SEM) was utilized to prepared transmission electron microscope samples from mixed oxide fuel with a burn-up of 6.7% FIMA. Utilizing the FIB/SEM for preparation resulted in samples with a dose rate of <0.5 mRem/h compared to approximately 1.1 R/h for a traditionally prepared TEM sample. The TEM analysis showed that the sample taken from the cooler rim region of the fuel pellet had approximately 2.5x higher dislocation density than that of the sample taken from the mid-radius due to the lower irradiation temperature of the rim. The dual column FIB/SEM was additionally used to prepared and serially slice approximately 25 um cubes. High quality electron back scatter diffraction (EBSD) were collected from the face at each step, showing, for the first time, the ability to obtain EBSD data from high activity irradiated fuel

Fabrication and control of fuels made of mixed carbides (U, Pu)C

International Nuclear Information System (INIS)

Lorenzelli, R.; Delaroche, P.

1980-01-01

Fabrication of this type of advanced fuel is described. The fuel is prepared by reduction of oxides with carbon and natural sintering. Density, thermal stability and thermal conductibility are more particularly studied [fr

Measurement of the Velocity and Pressure Drop in a Tubular Type Fuel

International Nuclear Information System (INIS)

Jonghark Park; Heetaek Chae; Cheol Park; Heonil Kim

2006-01-01

We have developed a tubular type fuel assembly design as one of candidates for fuel to be used in the Advanced HANARO Reactor (AHR). The tubular type fuel has several merits over a rod type fuel with respect to the thermal-hydraulic and structural safety; the larger ratio of surface area to volume makes the surface temperature of a fuel element become lower, and curved plate is stronger against longitudinal bending and vibration. In the other side, a disadvantage is expected such that the flow velocity can be distributed unevenly channel by channel because the flow channels are isolated from each other in a tubular type fuel assembly. In addition to the design development, we also investigated the flow characteristics of the tubular fuel experimentally. To examine the flow velocity distribution and pressure drop, we made an experiment facility and a mockup of the tubular fuel assembly. The fuel assembly consists of 6 concentric fuel tubes so that 7 layers are made between fuel tubes. Since each layer is divided into three sections by stiffeners, 21 isolated flow channels are made in total. We employed pitot-tubes to measure the coolant velocity in each channel. The maximum velocity was measured as large as about 28% of the average velocity. It was observed in the innermost channel contrarily to the expectation from the hydraulic diameter. A change in the total flow rate did not affect the flow distribution. Meanwhile, the pressure drop was measured as about 70% of the drop in the rod type fuel assembly in use in HANARO. (authors)

Uncommon Mixed Type I and II Choledochal Cyst: An Indonesian Experience

Directory of Open Access Journals (Sweden)

Fransisca J. Siahaya

2013-01-01

Full Text Available Bile duct cyst is an uncommon disease worldwide; however, its incidence is remarkably high in Asian population, primarily in children. Nevertheless, the mixed type choledochal cysts are extremely rare especially in adults. A case report of a 20-year-old female with a history of upper abdominal pain that was diagnosed with cholecystitis with stone and who underwent laparoscopic cholecystectomy is discussed. Choledochal malformation was found intraoperatively. Magnetic resonance cholangiography (MRCP and USG after first surgery revealed extrahepatic fusiform dilatation of the CBD; therefore, provisional diagnosis of type I choledochal cyst was made. Complete resection of the cyst was performed, and a mixed type I and II choledochal cyst was found intraoperatively. Bile duct reconstruction was carried out with Roux-en-Y hepaticojejunostomy. The mixed type I and II choledochal cysts are rare in adults, and this is the third adult case that has been reported. The mixed type can be missed on radiology imaging, and diagnosing the anomaly is only possible after a combination of imaging and intraoperative findings. Mixed type choledochal cyst classification should not be added to the existing classification since it does not affect the current operative techniques.

The fuel to clad heat transfer coefficient in advanced MX-type fuel pins

International Nuclear Information System (INIS)

Caligara, F.; Campana, M.; Mandler, R.; Blank, H.

1979-01-01

Advanced fuels (mixed carbides, nitrides and carbonitrides) are characterised by a high thermal conductivity compared to that of oxide fuels (5 times greater) and their behaviour under irradiation (amount of swelling, fracture behaviour, restructuring) is far more sensitive to the design parameters and to the operating temperature than that of oxide fuels. The use of advanced fuels is therefore conditioned by the possibility of mastering the above phenomena, and the full exploitation of their favorable neutron characteristics depends upon a good understanding of the mutual relationships of the various parameters, which eventually affect the mechanical stability of the pin. By far the most important parameter is the radial temperature profile which controls the swelling of the fuel and the build-up of stress fields within the pin. Since the rate of fission gas swelling of these fuels is relatively large, a sufficient amount of free space has to be provided within the pin. This space originally appears as fabrication porosity and as fuel-to-clad clearance. Due to the large initial gap width and to the high fuel thermal conductivity, the range of the fuel operating temperatures is mainly determined by the fuel-to-clad heat transfer coefficient h, whose correct determination becomes one of the central points in modelling. During the many years of modelling activity in the field of oxide fuels, several theoretical models have been developed to calculate h, and a large amount of experimental data has been produced for the empirical adjustment of the parameters involved, so that the situation may be regarded as rather satisfactory. The analysis lead to the following conclusions. A quantitative comparison of experimental h-values with existing models for h requires rather sophisticated instrumented irradiation capsules, which permit the measurement of mechanical data (concerning fuel and clad) together with heat rating and temperatures. More and better well

Analyses for licensing of new fuel types at Paks NPP

International Nuclear Information System (INIS)

Kereszturi, A.; Bogatyr, S.; Miko, S.; Nemes, I.

2003-01-01

In the last years Paks NPP initiated several projects aiming at the introduction of new fuel types and resulting in more economic fuel cycles. The motivations, the reasons, and the economic consequences of the above modifications are detailed. The application of a new fuel type requires the renewal of the relevant chapters of the Safety Analysis Report. The fulfilment of fuel design basis requirements, to be summarised briefly also in the paper, must be investigated during normal and accidental conditions. The characteristics of the different codes, the data transfer between them are detailed. After, the cases of the Normal Operation, Anticipated Operation Occurrence, and the Postulated Accidents, judged as the most relevant ones in case of fuel modifications, are overviewed. In the last part, selected examples of the licensing calculations, performed by the above tools are presented. In conclusion, modifications of the WWER fuel, namely increased enrichment, application of burnable fuel pins, modified geometry make more economic fuel cycles (larger discharge burnup, power up-rate, reduced pressure vessel fluence) are possible. The further step (increased enrichment, burnable poison) of the fuel modernisation at NPP Paks is necessary for more economic fuel cycles and fuel consuming. A sound basis of licensing methodology, safety analysis, and necessary computer codes for the WWER fuel modernisation is available

Fabrication of uranium-plutonium mixed nitride fuel pins (88F-5A) for first irradiation test at JMTR

International Nuclear Information System (INIS)

Suzuki, Yasufumi; Iwai, Takashi; Arai, Yasuo; Sasayama, Tatsuo; Shiozawa, Ken-ichi; Ohmichi, Toshihiko; Handa, Muneo

1990-07-01

A couple of uranium-plutonium mixed nitride fuel pins was fabricated for the first irradiation tests at JMTR for the purpose of understanding the irradiation behavior and establishing the feasibility of nitride fuels as advanced FBR fuels. The one of the pins was fitted with thermocouples in order to observe the central fuel temperature. In this report, the fabrication procedure of the pins such as pin design, fuel pellet fabrication and characterizations, welding of fuel pins, and inspection of pins are described, together with the outline of the new TIG welder installed recently. (author)

Study on effect of mixing mechanism by the transverse gaseous injection flow in scramjet engine with variable parameters

Science.gov (United States)

Yadav, Siddhita; Pandey, K. M.

2018-04-01

In scramjet engine the mixing mechanism of fuel and atmospheric air is very complicated, because the fuel have time in milliseconds for mixing with atmospheric air in combustion chamber having supersonic speed. Mixing efficiency of fuel and atmospheric air depends on mainly these parameters: Aspect ratio of injector, vibration amplitude, shock type, number of injector, jet to transverse flow momentum flux ratio, injector geometry, injection angle, molecular weight, incoming air stream angle, jet to transverse flow pressure ratio, spacing variation, mass flow rate of fuel etc. here is a very brief study of these parameters from previously done research on these parameters for the improvement of mixing efficiency. The mixing process have the significant role for the working of engine, and mixing between the atmospheric air and the jet fuel is significant factor for improving the overall thrust of the engine. The results obtained by study of papers are obtained by the 3D-Reynolds Average-Nervier-Stokes(RANS) equations along with the 2-equation k-ω shear-stress-transport (SST) turbulence model. Engine having multi air jets have 60% more mixing efficiency than single air jet, thus if the jets are increased, the mixing efficiency of engine can also be increased up to 150% by changing jet from 1 to 16. When using delta shape of injector the mixing efficiency is inversely proportional to the pressure ratio. When the fuel is injected inside the combustor from the top and bottom walls of the engine efficiency of mixing in reacting zone is higher than the single wall injection and in comparison to parallel flow, the transverse type flow is better as the atmospheric air jet can penetrate smoothly in the fuel jets and mixes well in less time. Hence this study of parameters and their effects on mixing can enhance the efficiency of mixing in engine.

A comprehensive guide to fuel management practices for dry mixed conifer forests in the northwestern United States: Mechanical, chemical, and biological fuel treatment methods

Science.gov (United States)

Theresa B. Jain; Mike A. Battaglia; Han-Sup Han; Russell T. Graham; Christopher R. Keyes; Jeremy S. Fried; Jonathan E. Sandquist

2014-01-01

Several mechanical approaches to managing vegetation fuels hold promise when applied to the dry mixed conifer forests in the western United States. These are most useful to treat surface, ladder, and crown fuels. There are a variety of techniques to remove or alter all kinds of plant biomass (live, dead, or decomposed) that affect forest resilience. It is important for...

PWR-to-PWR fuel cycle model using dry process

International Nuclear Information System (INIS)

Iqbal, M.; Jeong, Chang Joon; Rho, Gyu Hong

2002-03-01

PWR-to-PWR fuel cycle model has been developed to recycle the spent fuel using the dry fabrication process. Two types of fuels were considered; first fuel was based on low initial enrichment with low discharge burnup and second one was based on more initial enrichment with high discharge burnup in PWR. For recycling calculations, the HELIOS code was used, in which all of the available fission products were considered. The decay of 10 years was applied for reuse of the spent fuel. Sensitivity analysis for the fresh feed material enrichment has also been carried out. If enrichment of the mixing material is increased the saving of uranium reserves would be decreased. The uranium saving of low burned fuel increased from 4.2% to 7.4% in fifth recycling step for 5 wt% to 19.00wt% mixing material enrichment. While for high burned fuel, there was no uranium saving, which implies that higher uranium enrichment required than 5 wt%. For mixing of 15 wt% enriched fuel, the required mixing is about 21.0% and 37.0% of total fuel volume for low and high burned fuel, respectively. With multiple recycling, reductions in waste for low and high burned fuel became 80% and 60%, for first recycling, respectively. In this way, waste can be reduced more and the cost of the waste disposal reduction can provide the economic balance

Development of an engineered safeguards system concept for a mixed-oxide fuel fabrication facility

International Nuclear Information System (INIS)

Chapman, L.D.; de Montmollin, J.M.; Deveney, J.E.; Fienning, W.C.; Hickman, J.W.; Watkins, L.D.; Winblad, A.E.

1976-08-01

An initial concept of an Engineered Safeguards System for a representative commercial mixed-oxide fuel fabrication facility is presented. Computer simulation techniques for evaluation and further development of the concept are described. An outline of future activity is included

Fuel self-sufficient and low proliferation risk multi-recycling of spent fuel

International Nuclear Information System (INIS)

Cho, N. Z.; Hong, S. G.; Kim, T. H.; Greenspan, E.; Kastenberg, W. E.

1998-01-01

A preliminary feasibility study has been performed in search of promising nuclear energy systems which could make efficient use of the spent fuel from LWRs and be proliferation resistant. The energy considered consist of a dry process and a fuel-self-sufficient reactor which are synergistic. D 2 O, H 2 O and Pb (or Pb-Bi) are considered for the coolant. The most promising identified consists of Pb-cooled reactors with either an AIROX or an IFR-like reprocessing. H 2 O- (possibly mixed with D 2 O) cooled reactors can be designed to be fuel-self-sufficient and multi-recycle LWR spent fuel, provided they are accelerator driven. Moderator-free, D 2 O-cooled critical reactors can multi-recycle Th- 233 U fuel using IFR-type reprocessing; they are significantly more attractive than their thermal counterparts. H 2 O- (possibly mixed with D 2 O) cooled, accelerator-driven reactors appear attractive for converting Th into denatured 233 U using LWR spent fuel and the IFR process. The CANDU reactor technology appears highly synergistic with accelerator-driven systems. (author). 25 refs., 3 tabs., 6 figs

Development of core technology for research reactors using plate type fuels

International Nuclear Information System (INIS)

Ha, Jae Joo; Lee, Doo Jeong; Park, Cheol

2009-12-01

Around 250 research reactors are under operation over the world. However, about 2/3 have been operated more than 30 years and demands for replacements are expected in the near future. The number of expected units is around 110, and around 55 units from 40 countries will be expected to be bid in the world market. In 2007, Netherlands started international bidding process to construct a new 80MW RR (named PALLAS) with the target of commercial operation in 2016, which will replace the existing HFR(45MW). KAERI consortium has been participated in that bid. Most of RRs use plate type fuels as a fuel assembly, Be and Graphite as a reflector. On the other hand, in Korea, the KAERI is operating the HANARO, which uses a rod type fuel assembly and heavy water as a reflector. Hence, core technologies for RRs using plate type fuels are in short. Therefore, core technologies should be secured for exporting a RR. In chapter 2, the conceptual design of PALLAS which use plate type fuels are described including core, cooling system and connected systems, layout of general components. Experimental verification tests for the plate type fuel and second shutdown system and the code verification for nuclear design are explained in Chapter 3 and 4, respectively

Optimal fuel-mix in CHP plants under a stochastic permit price. Risk-neutrality versus risk-aversion

International Nuclear Information System (INIS)

Lappi, Pauli; Ollikka, Kimmo; Ollikainen, Markku

2010-01-01

This paper studies the optimal fuel-mix of a CHP producer under emission permit price risk. The producer's multi-fuel plant uses two CO 2 -intensive fuels and one clean fuel. Using a mean-variance framework we develop three models. The models are divided into spot-models (risk neutral and risk averse cases) and a forward-model (risk averse case). We derive the effects of price risk on optimal fuel use. An increase in price risk can in fact increase the use of CO 2 -intensive fuel in the spot-model. In the forward-model, the production and financial decisions are separate. We also evaluate the risk-bearing behavior of seven Finnish CHP producers. We found that risk-neutrality describes behavior better than risk-aversion. (author)

Results of the post-irradiation examination of a highly-rated mixed oxide fuel rod from the Mol 7B experiment

International Nuclear Information System (INIS)

Coquerelle, M.; Walker, C.T.; Whitlow, W.H.

1980-01-01

The experiment MOL 7B was carried out in a epithermal flux in the Belgian reactor BR2. The pin examined contained fuel of initial composition (Usub(0.7)Pusub(0.3))Osub(1.98). It had been irradiated to a maximum burn-up of 13.2 at% at a maximum linear power of 568Wcm -1 . The fuel was clad with coldworked stainless steel. Electron microprobe analysis indicated that a Cr 2 O 3 type oxide was the main constituent of the grey phases in the gap. A metallic phase on the fuel surface had apparently resulted from the mechanical compaction of fragments of cladding that had been depleted in chromium by oxidation. Thus the main components of the phase were iron and nickel. Chromium loss from the inner cladding surface was significant only in the upper regions of the pin. In pin sections that were metallographically examined sigma phase and carbides of the type M 23 C 6 were present at the grain boundaries of the cladding. Cladding corrosion was not an Arrhenius function of the cladding temperature: the amount of metal lost from the inner cladding surface decreased with rise in cladding temperature above 910 K. A contributor to metal loss was the mechanical detachment of fragments of cladding which reformed as a metallic layer on the surface of the fuel. Chromium depletion and sigma phase formation at grain boundaries lowered the cohesive forces between grains which were then mechanically detached. Chromium loss from grain boundaries is mainly the results of oxidation of the cladding by the mixed oxide fuel. Data are presented to support the view that the local average O/M of the fuel determined the rate of oxidation and consequently the extent of chromium depletion. Fuel-cladding mechanical interactions were weak in the upper regions of the pin where metal loss was small

Ultra low injection angle fuel holes in a combustor fuel nozzle

Science.gov (United States)

York, William David

2012-10-23

A fuel nozzle for a combustor includes a mixing passage through which fluid is directed toward a combustion area and a plurality of swirler vanes disposed in the mixing passage. Each swirler vane of the plurality of swirler vanes includes at least one fuel hole through which fuel enters the mixing passage in an injection direction substantially parallel to an outer surface of the plurality of swirler vanes thereby decreasing a flameholding tendency of the fuel nozzle. A method of operating a fuel nozzle for a combustor includes flowing a fluid through a mixing passage past a plurality of swirler vanes and injecting a fuel into the mixing passage in an injection direction substantially parallel to an outer surface of the plurality of swirler vanes.

On the actual cathode mixed potential in direct methanol fuel cells

Science.gov (United States)

Zago, M.; Bisello, A.; Baricci, A.; Rabissi, C.; Brightman, E.; Hinds, G.; Casalegno, A.

2016-09-01

Methanol crossover is one of the most critical issues hindering commercialization of direct methanol fuel cells since it leads to waste of fuel and significantly affects cathode potential, forming a so-called mixed potential. Unfortunately, due to the sluggish anode kinetics, it is not possible to obtain a reliable estimation of cathode potential by simply measuring the cell voltage. In this work we address this limitation, quantifying the mixed potential by means of innovative open circuit voltage (OCV) tests with a methanol-hydrogen mixture fed to the anode. Over a wide range of operating conditions, the resulting cathode overpotential is between 250 and 430 mV and is strongly influenced by methanol crossover. We show using combined experimental and modelling analysis of cathode impedance that the methanol oxidation at the cathode mainly follows an electrochemical pathway. Finally, reference electrode measurements at both cathode inlet and outlet provide a local measurement of cathode potential, confirming the reliability of the innovative OCV tests and permitting the evaluation of cathode potential up to typical operating current. At 0.25 A cm-2 the operating cathode potential is around 0.85 V and the Ohmic drop through the catalyst layer is almost 50 mV, which is comparable to that in the membrane.

Fuel assemblies for BWR type reactors

International Nuclear Information System (INIS)

Ishizuka, Takao.

1981-01-01

Purpose: To enable effective failed fuel detection by the provision of water rod formed with a connecting section connected to a warmed water feed pipe of a sipping device at the lower portion and with a warmed water jetting port in the lower portion in a fuel assembly of a BWR type reactor to thereby carry out rapid sipping. Constitution: Fuel rods and water rods are contained in the channel box of a fuel assembly, and the water rod is provided at its upper portion with a connecting section connected to the warmed water feed pipe of the sipping device and formed at its lower portion with a warmed water jetting port for jetting warmed water fed from the warmed water feed pipe. Upon detection of failed fuels, the reactor operation is shut down and the reactor core is immersed in water. The cover for the reactor container is removed and the cap of the sipping device is inserted to connect the warmed water feed pipe to the connecting section of the water rod. Then, warmed water is fed to the water rod and jetted out from the warmed water jetting port to cause convection and unify the water of the channel box in a short time. Thereafter, specimen is sampled and analyzed for the detection of failed fuels. (Moriyama, K.)

Concept for Multi-cycle Nuclear Fuel Optimization Based On Parallel Simulated Annealing With Mixing of States

International Nuclear Information System (INIS)

Kropaczek, David J.

2008-01-01

A new concept for performing nuclear fuel optimization over a multi-cycle planning horizon is presented. The method provides for an implicit coupling between traditionally separate in-core and out-of-core fuel management decisions including determination of: fresh fuel batch size, enrichment and bundle design; exposed fuel reuse; and core loading pattern. The algorithm uses simulated annealing optimization, modified with a technique called mixing of states that allows for deployment in a scalable parallel environment. Analysis of algorithm performance for a transition cycle design (i.e. a PWR 6 month cycle length extension) demonstrates the feasibility of the approach as a production tool for fuel procurement and multi-cycle core design. (authors)

Development of ORIGEN libraries for mixed oxide (MOX) fuel assembly designs

International Nuclear Information System (INIS)

Mertyurek, Ugur; Gauld, Ian C.

2016-01-01

Highlights: • ORIGEN MOX library generation process is described. • SCALE burnup calculations are validated against measured MOX fuel samples from the MALIBU program. • ORIGEN MOX libraries are verified using the OECD Phase IV-B benchmark. • There is good agreement for calculated-to-measured isotopic distributions. - Abstract: ORIGEN cross section libraries for reactor-grade mixed oxide (MOX) fuel assembly designs have been developed to provide fast and accurate depletion calculations to predict nuclide inventories, radiation sources and thermal decay heat information needed in safety evaluations and safeguards verification measurements of spent nuclear fuel. These ORIGEN libraries are generated using two-dimensional lattice physics assembly models that include enrichment zoning and cross section data based on ENDF/B-VII.0 evaluations. Using the SCALE depletion sequence, burnup-dependent cross sections are created for selected commercial reactor assembly designs and a representative range of reactor operating conditions, fuel enrichments, and fuel burnup. The burnup dependent cross sections are then interpolated to provide problem-dependent cross sections for ORIGEN, avoiding the need for time-consuming lattice physics calculations. The ORIGEN libraries for MOX assembly designs are validated against destructive radiochemical assay measurements of MOX fuel from the MALIBU international experimental program. This program included measurements of MOX fuel from a 15 × 15 pressurized water reactor assembly and a 9 × 9 boiling water reactor assembly. The ORIGEN MOX libraries are also compared against detailed assembly calculations from the Phase IV-B numerical MOX fuel burnup credit benchmark coordinated by the Nuclear Energy Agency within the Organization for Economic Cooperation and Development. The nuclide compositions calculated by ORIGEN using the MOX libraries are shown to be in good agreement with other physics codes and with experimental data.

CFD modeling of turbulent mixing through vertical pressure tube type boiling water reactor fuel rod bundles with spacer-grids

Science.gov (United States)

Verma, Shashi Kant; Sinha, S. L.; Chandraker, D. K.

2018-05-01

Numerical simulation has been carried out for the study of natural mixing of a Tracer (Passive scalar) to describe the development of turbulent diffusion in an injected sub-channel and, afterwards on, cross-mixing between adjacent sub-channels. In this investigation, post benchmark evaluation of the inter-subchannel mixing was initiated to test the ability of state-of-the-art Computational Fluid Dynamics (CFD) codes to numerically predict the important turbulence parameters downstream of a ring type spacer grid in a rod-bundle. A three-dimensional Computational Fluid Dynamics (CFD) tool (STAR-CCM+) was used to model the single phase flow through a 30° segment or 1/12th of the cross segment of a 54-rod bundle with a ring shaped spacer grid. Polyhedrons were used to discretize the computational domain, along with prismatic cells near the walls, with an overall mesh count of 5.2 M cell volumes. The Reynolds Stress Models (RSM) was tested because of RSM accounts for the turbulence anisotropy, to assess their capability in predicting the velocities as well as mass fraction of potassium nitrate measured in the experiment. In this way, the line probes are located in the different position of subchannels which could be used to characterize the progress of the mixing along the flow direction, and the degree of cross-mixing assessed using the quantity of tracer arriving in the neighbouring sub-channels. The predicted dimensionless mixing scalar along the length, however, was in good agreement with the measurements downstream of spacers.

Survey of Worldwide Light Water Reactor Experience with Mixed Uranium-Plutonium Oxide Fuel

Energy Technology Data Exchange (ETDEWEB)

Cowell, B.S.; Fisher, S.E.

1999-02-01

The US and the Former Soviet Union (FSU) have recently declared quantities of weapons materials, including weapons-grade (WG) plutonium, excess to strategic requirements. One of the leading candidates for the disposition of excess WG plutonium is irradiation in light water reactors (LWRs) as mixed uranium-plutonium oxide (MOX) fuel. A description of the MOX fuel fabrication techniques in worldwide use is presented. A comprehensive examination of the domestic MOX experience in US reactors obtained during the 1960s, 1970s, and early 1980s is also presented. This experience is described by manufacturer and is also categorized by the reactor facility that irradiated the MOX fuel. A limited summary of the international experience with MOX fuels is also presented. A review of MOX fuel and its performance is conducted in view of the special considerations associated with the disposition of WG plutonium. Based on the available information, it appears that adoption of foreign commercial MOX technology from one of the successful MOX fuel vendors will minimize the technical risks to the overall mission. The conclusion is made that the existing MOX fuel experience base suggests that disposition of excess weapons plutonium through irradiation in LWRs is a technically attractive option.

Survey of Worldwide Light Water Reactor Experience with Mixed Uranium-Plutonium Oxide Fuel

International Nuclear Information System (INIS)

Cowell, B.S.; Fisher, S.E.

1999-01-01

The US and the Former Soviet Union (FSU) have recently declared quantities of weapons materials, including weapons-grade (WG) plutonium, excess to strategic requirements. One of the leading candidates for the disposition of excess WG plutonium is irradiation in light water reactors (LWRs) as mixed uranium-plutonium oxide (MOX) fuel. A description of the MOX fuel fabrication techniques in worldwide use is presented. A comprehensive examination of the domestic MOX experience in US reactors obtained during the 1960s, 1970s, and early 1980s is also presented. This experience is described by manufacturer and is also categorized by the reactor facility that irradiated the MOX fuel. A limited summary of the international experience with MOX fuels is also presented. A review of MOX fuel and its performance is conducted in view of the special considerations associated with the disposition of WG plutonium. Based on the available information, it appears that adoption of foreign commercial MOX technology from one of the successful MOX fuel vendors will minimize the technical risks to the overall mission. The conclusion is made that the existing MOX fuel experience base suggests that disposition of excess weapons plutonium through irradiation in LWRs is a technically attractive option

Fuel assembly for nuclear reactor

International Nuclear Information System (INIS)

Yamanaka, Akihiro; Haikawa, Katsumasa; Haraguchi, Yuko; Nakamura, Mitsuya; Aoyama, Motoo; Koyama, Jun-ichi.

1996-01-01

In a BWR type fuel assembly comprising first fuel rods filled with nuclear fission products and second fuel rods filled with burnable poisons and nuclear fission products, the concentration of the burnable poisons mixed to a portion of the second fuel rods is controlled so that it is reduced at the upper portion and increased at the lower portion in the axial direction. In addition, a product of the difference of an average concentration of burnable poisons between the upper portion and the lower portion and the number of fuel rods is determined to higher than a first set value determined corresponding to the limit value of a maximum linear power density. The sum of the difference of the average concentration of the burnable poisons between the upper portion and the lower portion of the second fuel rod and the number of the second fuel rods is determined to lower than a second set value determined corresponding to a required value of a surplus reactivity. If the number of the fuel rods mixed with the burnable poisons is increased, the infinite multiplication factor at an initial stage of the burning is lowered and, if the concentration of the mixed burnable poisons is increased, the time of exhaustion of the burnable poisons is delayed. As a result, the maximum value of the infinite multiplication factor is suppressed thereby enabling to control surplus reactivity. (N.H.)

THE EFFECT OF SKULDUGGERY IN FUEL OF DIESEL ENGINES ON THE PERFORMANCE OF I. C. ENGINE

Directory of Open Access Journals (Sweden)

Raed R. Jasem

2013-05-01

Full Text Available The current research aimed to study the effect of fraud in the diesel fuel on environmental pollution,  the study included two samples of diesel fuel., first sample is used currently in all diesel engines vehicles, and it produced in colander of oil  of Baiji, the second sample is producer manually from mixing of the Lubricating oils and kerosene with ratio(1/40, were prepared and tested in research laboratories and quality control of the North Refineries Company /BAIJI by using standard engine (CFR. comparison between two models of fuel in terms of the properties of the mixing fuel and the properties of diesel fuel standard. The results proved that the process of mixing these ,  leading to the minimization of Cetane number and flash point. While the viscosity increase in  mixing fuel, comparison with fuel producer in the refinery, and which identical to the minimum standard specifications of diesel fuel.The tests had been carried out using the engine of (TQ four stroke type (TD115 with a single-cylinder and compression ratio (21:1 a complement to the hydraulic type Dynamo meter (TD115.

Design characteristics of pantograph type in vessel fuel handling system in SFR

International Nuclear Information System (INIS)

Kim, S. H.; Koo, G. H.

2012-01-01

The pantograph type in vessel fuel handling system in a sodium cooled fast reactor (SFR), which requires installation space for the slot in the upper internal structure attached under the rotating plug, is composed of an in vessel transfer machine (IVTM), a single rotating plug, in vessel storage, and a fuel transfer port (FTP). The pantograph type IVTM can exchange fuel assemblies through a slot, the design requirement of which should be essentially considered in the design of the in vessel fuel handling system. In addition, the spent fuel assemblies temporarily stored in the in vessel storage of the reactor vessel are removed to the outside of the reactor vessel through the FTP. The fuel transfer basket is then provided in the FTP, and a fuel transfer is performed by using it. In this study, the design characteristics for a pantograph type in vessel fuel handling system are reviewed, and the preconceptual designs are studied

Design characteristics of pantograph type in vessel fuel handling system in SFR

Energy Technology Data Exchange (ETDEWEB)

Kim, S. H.; Koo, G. H. [KAERI, Daejeon (Korea, Republic of)

2012-10-15

The pantograph type in vessel fuel handling system in a sodium cooled fast reactor (SFR), which requires installation space for the slot in the upper internal structure attached under the rotating plug, is composed of an in vessel transfer machine (IVTM), a single rotating plug, in vessel storage, and a fuel transfer port (FTP). The pantograph type IVTM can exchange fuel assemblies through a slot, the design requirement of which should be essentially considered in the design of the in vessel fuel handling system. In addition, the spent fuel assemblies temporarily stored in the in vessel storage of the reactor vessel are removed to the outside of the reactor vessel through the FTP. The fuel transfer basket is then provided in the FTP, and a fuel transfer is performed by using it. In this study, the design characteristics for a pantograph type in vessel fuel handling system are reviewed, and the preconceptual designs are studied.

Effect of a time varying power level in EBR-II on mixed-oxide fuel burnup

International Nuclear Information System (INIS)

Stone, I.Z.; Jost, J.W.; Baker, R.B.

1979-01-01

A refined prediction of burnup of mixed-oxide fuel in EBR-2 is compared with measured data. The calculation utilizes a time-varying power factor and results in a general improvement to previous calculations

Optimal fuel-mix in CHP plants under a stochastic permit price. Risk-neutrality versus risk-aversion

Energy Technology Data Exchange (ETDEWEB)

Lappi, Pauli; Ollikka, Kimmo; Ollikainen, Markku [Department of Economics and Management, P.O. Box 27, University of Helsinki, FIN-00014 Helsinki (Finland)

2010-02-15

This paper studies the optimal fuel-mix of a CHP producer under emission permit price risk. The producer's multi-fuel plant uses two CO{sub 2}-intensive fuels and one clean fuel. Using a mean-variance framework we develop three models. The models are divided into spot-models (risk neutral and risk averse cases) and a forward-model (risk averse case). We derive the effects of price risk on optimal fuel use. An increase in price risk can in fact increase the use of CO{sub 2}-intensive fuel in the spot-model. In the forward-model, the production and financial decisions are separate. We also evaluate the risk-bearing behavior of seven Finnish CHP producers. We found that risk-neutrality describes behavior better than risk-aversion. (author)

Optimal fuel-mix in CHP plants under a stochastic permit price: Risk-neutrality versus risk-aversion

Energy Technology Data Exchange (ETDEWEB)

Lappi, Pauli, E-mail: pauli.lappi@helsinki.f [Department of Economics and Management, P.O. Box 27, University of Helsinki, FIN-00014 Helsinki (Finland); Ollikka, Kimmo, E-mail: kimmo.ollikka@helsinki.f [Department of Economics and Management, P.O. Box 27, University of Helsinki, FIN-00014 Helsinki (Finland); Ollikainen, Markku, E-mail: markku.ollikainen@helsinki.f [Department of Economics and Management, P.O. Box 27, University of Helsinki, FIN-00014 Helsinki (Finland)

2010-02-15

This paper studies the optimal fuel-mix of a CHP producer under emission permit price risk. The producer's multi-fuel plant uses two CO{sub 2}-intensive fuels and one clean fuel. Using a mean-variance framework we develop three models. The models are divided into spot-models (risk neutral and risk averse cases) and a forward-model (risk averse case). We derive the effects of price risk on optimal fuel use. An increase in price risk can in fact increase the use of CO{sub 2}-intensive fuel in the spot-model. In the forward-model, the production and financial decisions are separate. We also evaluate the risk-bearing behavior of seven Finnish CHP producers. We found that risk-neutrality describes behavior better than risk-aversion.

Pilot and pilot-commercial plants for reprocessing spent fuels of FBR type reactors

International Nuclear Information System (INIS)

Shaldaev, V.S.; Sokolova, I.D.

1988-01-01

A review of modern state of investigations on the FBR mixed oxide uranium-plutonium fuel reprocessing abroad is given. Great Britain and France occupy the leading place in this field, operating pilot plants of 5 tons a year capacity. Technology of spent fuel reprocessing and specific features of certain stages of the technological process are considered. Projects of pilot and pilot-commercial plants of Great Britain, France, Japan, USA are described. Economic problems of the FBR fuel reprocessing are touched upon

Sooting limit in counterflow diffusion flames of ethylene/propane fuels and implication to threshold soot index

KAUST Repository

Joo, Peter H.

2013-01-01

Sooting limits in counterflow diffusion flames of propane/ethylene fuels have been studied experimentally using a light scattering technique, including the effects of dilution, fuel mixing, and strain rate. The results are discussed in view of the threshold soot index (TSI). In soot-formation (SF) flames, where the flame is located on the oxidizer side of the stagnation plane, the sooting limit depends critically on fuel type and subsequently on flame temperature. The sooting limit has a non-linear dependence on the fuel-mixing ratio, which is similar to the non-linear mixing rule for TSI observed experimentally in rich premixed flames, where soot oxidation is absent for both SF and rich premixed flames. In soot-formation-oxidation (SFO) flames, where the flame is located on the fuel side, the sooting limit depends critically on flame temperature, while it is relatively independent on fuel type. This result suggests a linear mixing rule for sooting limits in SFO flames, which is similar to the TSI behavior for coflow diffusion flames. Soot oxidation takes place for both types of flames. The aerodynamic strain effect on the sooting limits has also been studied and an appreciable influence has been observed. Under sooting conditions, soot volume fraction was measured using a light extinction technique. The soot loadings in SF flames of the mixture fuels demonstrated a synergistic effect, i.e., soot production increased for certain mixture fuels as compared to the respective singlecomponent fuels. © 2012 The Combustion Institute.

Design of an engineered safeguards system for a mixed-oxide fuel fabrication facility

International Nuclear Information System (INIS)

Winblad, A.E.; McKnight, R.P.; Fienning, W.C.; Fenchel, B.R.

1977-06-01

Several Engineered Safeguards System concepts and designs are described that provide increased protection against a wide spectrum of adversary threats. An adversary sequence diagram that outlines all possible adversary paths through the safeguards elements in a mixed-oxide fuel fabrication facility is shown. An example of a critical adversary path is given

Uranium and plutonium distribution in unirradiated mixed oxide fuel from industrial fabrication

International Nuclear Information System (INIS)

Hanus, D.; Kleykamp, H.

1982-01-01

Different process variants developed in the last few years by the firm ALKEM to manufacture FBR and LWR mixed oxide fuel are given. The uranium and plutonium distribution is determined on the pellets manufactured with the help of the electron beam microprobe. The stepwise improvement of the uranium-plutonium homogeneity in the short-term developed granulate variants and in the long-term developed new processes are illustrated starting with early standard processes for FBR fuel. An almost uniform uranium-plutonium distribution could be achieved for the long-term developed new processes (OKOM, AuPuC). The uranium-plutonium homogeneity are quantified in the pellets manufactured according to the considered process variants with a newly defined quality number. (orig.)

Stressed and strained state for cermetic-rod-type fuel element

International Nuclear Information System (INIS)

Kulikov, I.S.

1987-01-01

Calculation technique for designing the stress-strained state of a cermetic rod-type fuel element has been proposed. The technique is based on the time-dependent step-by-step method and the solution of the deformation equilibrium equation for continuous and thick-wall long cylinders at every temporal step by the finite difference method. Additional strains, caused by thermal expansion and radiation swelling, have been taken into account. The transion from the non-contact model to the stiff-contact model has been provided in the case of cladding-fuel gap dissappearing in one or a number of cross-sections along the fuel element height. The method is supplemented by the formula for fuel cans stability estimation in the case of high coolant external pressure. The example of estimation of the cermetic-rod-type fuel elements are considered as an example

Fuel formulation and mixing strategy for rate of heat release control with PCCI combustion

NARCIS (Netherlands)

Zegers, R.P.C.; Yu, M.; Luijten, C.C.M.; Dam, N.J.; Baert, R.S.G.; Goey, de L.P.H.

2009-01-01

Premixed charge compression ignition (or PCCI) is a new combustion concept that promises very low emissions of nitrogen oxides and of particulate matter by internal combustion engines. In the PCCIcombustion mode fuel, products from previous combustion events and air are mixed and compresseduntil the

The Jump Problem for Mixed-Type Equations with Defects on the Type Change Line

Directory of Open Access Journals (Sweden)

Ahmed Maher

2010-01-01

Full Text Available The jump problem and problems with defects on the type change line for model mixed-type equations in the mixed domains are investigated. The explicit solutions of the jump problem are obtained by the method of integral equations and by the Fourier transformation method. The problems with defects are reduced to singular integral equations. Some results for the solution of the equation under consideration are discussed concerning the existence and uniqueness for the solution of the suggested problem.

Analysis Influence of Mixing Gd2O3 in the Silicide Fuel Element to Core Excess Reactivity of RSG-GAS

International Nuclear Information System (INIS)

Susilo, Jati

2004-01-01

Gadolinium (Gd 2 O 3 ) is a burnable poison material mixed in the pin fuel element of the LWR core used to decrease core excess reactivity. In this research, analysis influence of mixing Gd 2 O 3 in the silicide fuel element to excess reactivity of the RSG-GAS core had been done. Equivalent cell of the equilibrium core developed by L.E.Strawbridge from Westing House Co. burn-up calculation has been done using SRAC-PIJ computer code achieve infinite multiplication factor (k x ). Value of Gd 2 O 3 concentration in the fuel element (pcm) showed by mass ratio of Gd 2 O 3 (gram) to that U 3 Si 2 (gram) times 10 5 , that is 0 pcm ∼ 100 pcm. From the calculation results analysis showed that Gd 2 O 3 concentration added should be considered. because a large number of Gd 2 O 3 will result in not achieving criticality at the Beginning Of Cycle. The maximum concentration of Gd 2 O 3 for RSG-GAS equilibrium fueled silicide 2.96 grU/cc is 80 pcm or 52.02 mgram/fuel plate. Maximum reduction of core excess reactivity due to mixing of Gd 2 O 3 in the RSG-GAS silicide fuels was around 1.502 %Δk/k, and hence not achieving the standard nominal excess reactivity for RSG-GAS core using high density of U 3 Si 2 -Al fuel. (author)

Fuel performance of rod-type research reactor fuel using a centrifugally atomized U-Mo powder

International Nuclear Information System (INIS)

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

2009-01-01

A low enriched uranium nuclear fuel for research reactors has been developed in order to replace a highly enriched uranium fuel according to the non-proliferation policy under the reduced enrichment for research and test reactors (RERTR) program. In KAERI, a rod-type U 3 Si dispersion fuel has been developed for a localization of the HANARO fuel and a U 3 Si/Al dispersion fuel of 3.15 gU/cc has been used at HANARO as a driver fuel since 2005. Although uranium silicide dispersion fuels such as U 3 Si 2 /Al and U 3 Si/Al are being used widely, high uranium density dispersion fuels (8-9 g/cm 3 ) are required for some high performance research reactors. U-Mo alloys have been considered as one of the most promising uranium alloys for a dispersion fuel due to their good irradiation performance. An international qualification program on U-Mo fuel to replace a uranium silicide dispersion fuel with a U-Mo dispersion fuel has been carried out

The prospects of use of alternative types of fuel in road transport ...

African Journals Online (AJOL)

The article is devoted to the analysis of possibilities of using alternative types of fuel in transport. Gas engine fuels are considered as potential energy carriers for diesel engines. Since the constructions of vehicles, using gas and traditional types of fuel, have some differences, the most important are the issues of ensuring ...

FBR type reactor

International Nuclear Information System (INIS)

Hayase, Tamotsu.

1991-01-01

The present invention concerns an FBR type reactor in which transuranium elements are eliminated by nuclear conversion. There are loaded reactor core fuels being charged with mixed oxides of plutonium and uranium, and blanket fuels mainly comprising depleted uranium. Further, liquid sodium is used as coolants. As transuranium elements, isotope elements of neptunium, americium and curium contained in wastes taken out from light water reactors or the composition thereof are used. The reactor core comprises a region with a greater mixing ratio and a region with a less mixing ratio of the transuranium elements. The mixing ratio of the transuranium elements is made greater for the fuels in the reactor core region at the boundary with the blanket of great neutron leakage. With such a constitution, since the positive reactivity value at the reactor core central portion is small in the Na void reactivity distribution in the reactor core, the positive reactivity is small upon Na boiling in the reactor core central region upon occurrence of imaginable accident, to attain reactor safety. (I.N.)

Soot and NOx simultaneous reduction by use of CO2 mixed fuel; Ekika CO2 yokai nenryo ni yoru diesel kikan no susu, NOx no doji teigen

Energy Technology Data Exchange (ETDEWEB)

Senda, J; Yokoyama, T; Ikeda, M; Fujimoto, H [Doshisha University, Kyoto (Japan); Ifuku, Y [Kubota Corp., Osaka (Japan)

1997-10-01

We propose the new fuel injection system by use of diesel fuel dissolved with CO2 to reduce both soot and NOx simultaneously. In this paper spray combustion characteristics of CO2 mixed fuel is reported. It is revealed that flame temperature and KL factor at the CO2 mixed fuel combustion are lower than at the only n-tridecane combustion due to separation or partly flashing of CO2component. And the result of exhaust gas measurement shows the capability that CO2 mixed fuel is able to reduce both soot and NOx simultaneously. 12 refs., 7 figs., 1 tab.

In-use vs. type-approval fuel consumption of current passenger cars in Europe

International Nuclear Information System (INIS)

Ntziachristos, L.; Mellios, G.; Tsokolis, D.; Keller, M.; Hausberger, S.; Ligterink, N.E.; Dilara, P.

2014-01-01

In-use fuel consumption data of 924 passenger cars (611 petrol, 313 diesel) were collected from various European sources and were evaluated in comparison to their corresponding type-approval values. The analysis indicated that the average in-use fuel consumption was higher than the type-approval one by 11% for petrol cars and 16% for diesel cars. Comparison of this dataset with the Travelcard database in the Netherlands showed that the deviation increased for late model years and in particular for cars with low type-approval values. The deviation was higher than 60% for vehicles registered in 2012 within the 90–100 gCO 2 /km bin. Unrealistic vehicle resistances used in type-approval were identified as one of the prime reasons of the difference. A simplified linear model developed in the study may be used to predict in-use fuel consumption based on data publicly available. The model utilizes the fuel consumption measured in type-approval, the mass, and the engine capacity to provide in-use fuel consumption. This may be either used to correct fuel consumption factors currently utilized by emission models (e.g. COPERT, HBEFA, VERSIT+, and others) or could be used independently to make projections on how fuel consumption may develop on the basis of changing future passenger cars characteristics. - Highlights: • In-use fuel consumption of petrol and diesel passenger cars is 11% and 16% higher than type-approval, respectively. • The relative difference between in-use and type-approval increases for late model and vehicles with low consumption. • Unrealistically low vehicle resistances are identified as a prime reason of low type-approval fuel consumption. • A model developed predicts in-use consumption on the basis of type-approval consumption, vehicle mass, and engine capacity

Laminar simulation of intersubchannel mixing in a triangular nuclear fuel bundle geometry

International Nuclear Information System (INIS)

Zaretsky, A.; Lightstone, M.F.; Tullis, S.

2015-01-01

Highlights: • Quasi-periodic flow was observed through rod-to-wall gaps. • Triangular subchannel flows were fundamentally irregular. • Cross-gap flow was influenced both by local and adjacent cross-gap intensity. • Phase-linking between gaps induced cross-plane peripheral circulation through rod–wall gaps. • Cross-gap flow structure was dependent on subchannel geometry. - Abstract: Predicting temperature distributions in fuel rod bundles is an important component of nuclear reactor safety analysis. Intersubchannel mixing acts to homogenize coolant temperatures thus reducing the likelihood of localized regions of high fuel temperature. Previous research has shown that intersubchannel mixing in nuclear fuel rod bundles is enhanced by a large-scale quasi-periodic energetic fluid motion, which transports fluid on the cross-plane between the narrow gaps connecting subchannels. This phenomenon has also been observed in laminar flows. Unsteady laminar flow simulations were performed in a simplified bundle of three rods with a pipe. Three similar geometries of varying gap width were examined, and a thermal trace was implemented on the first geometry. Thermal mixing was driven by the advection of energy between subchannels by the cross-plane flow. Flow through the rod-to-wall gaps in the wall subchannels alternated with a dominant frequency, particularly when rod-to-wall gaps were smaller than rod-to-rod gaps. Significant phase-linking between rod-to-wall gaps was also observed such that a peripheral circulation occurred through each gap simultaneously. Cross-plane flow through the rod-to-rod gaps in the triangular subchannel was irregular in each case. This was due to the fundamental irregularity of the triangular subchannel geometry. Vortices were continually broken up by cross-plane flow from other gaps due to the odd number of fluid pathways within the central subchannel. Cross-plane flow in subchannel geometries is highly interconnected between gaps. The

Control in fabrication of PWR and BWR type reactor fuel elements

International Nuclear Information System (INIS)

Gorskij, V.V.

1981-01-01

Both destructive and non-destructive testing methods now in use in fabrication of BWR and PWR type reactor fuel elements at foreign plants are reviewed. Technological procedures applied in fabrication of fuel elements and fuel assemblies are described. Major attention is paid to radiographic, ultrasonic, metallographic, visual and autoclavic testings. A correspondence of the methods applied to the ASTM standards is discussed. The most part of the countries are concluded the apply similar testing methods enabling one to reliably evaluate the quality of primary materials and fabricated fuel elements and thus meeting the demands to contemporary PWR and BWR type reactor fuel elements. Practically all fuel element and pipe fabrication plants in Western Europe, Asia and America use the ASTM standards as the basis for the quality contr [ru

Development of DIPRES feed for the fabrication of mixed-oxide fuels for fast breeder reactors

International Nuclear Information System (INIS)

Griffin, C.W.; Rasmussen, D.E.; Lloyd, M.H.

1983-01-01

The DIrect PREss Spheroidized feed process combines the conversion of uranium-plutonium solutions into spheres by internal gelation with conventional pellet fabrication techniques. In this manner, gel spheres could replace conventional powders as the feed material for pellet fabrication of nuclear fuels. Objective of the DIPRES feed program is to develop and qualify a process to produce mixed-oxide fuel pellets from gel spheres for fast breeder reactors. This process development includes both conversion and fabrication activities

A CAREM type fuel element dynamic analysis

International Nuclear Information System (INIS)

Magoia, J.E.

1990-01-01

A first analysis on the dynamic behaviour of a fuel element designed for the CAREM nuclear reactor (Central Argentina de Elementos Modulares) was performed. The model used to represent this dynamic behaviour was satisfactorily evaluated. Using primary estimations for some of its numerical parameters, a first approximation to its natural vibrational modes was obtained. Results obtained from fuel elements frequently used in nuclear power plants of the PWR (Pressurized Water Reactors) type, are compared with values resulting from similar analysis. (Author) [es

Proton exchange membrane fuel cells modeling

CERN Document Server

Gao, Fengge; Miraoui, Abdellatif

2013-01-01

The fuel cell is a potential candidate for energy storage and conversion in our future energy mix. It is able to directly convert the chemical energy stored in fuel (e.g. hydrogen) into electricity, without undergoing different intermediary conversion steps. In the field of mobile and stationary applications, it is considered to be one of the future energy solutions.Among the different fuel cell types, the proton exchange membrane (PEM) fuel cell has shown great potential in mobile applications, due to its low operating temperature, solid-state electrolyte and compactness.This book pre

Mixed PWR core loadings with inert matrix Pu-fuel assemblies

International Nuclear Information System (INIS)

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

1999-01-01

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

Electromagnetic Acoustic Test of the Artificial Defects for a Plate-type Nuclear Fuel

International Nuclear Information System (INIS)

Jung, Hyun Kyu; Kim, Dong Min; Lee, Yoon Sang; Cheong, Yong Moo

2011-01-01

Most research and test reactors use the nuclear fuel plates which are consisted of a fuel meat in aluminum alloy. Last year, KAERI signed a deal with the Jordan Atomic Energy Commission to build the research reactor and have to supply the plate-type nuclear fuels. For the demands of world market, KAERI started the research and development of the plate-type fuel elements and endeavored to achieve a localization of the plate-type fuel fabrication. For the inspection of plate-type fuel elements to be used in Research Reactors, an immersion pulse-echo ultrasonic technique was applied. This inspection was done under immersion condition, so a nuclear fuel was immersed to be prone to corrosion and needed to have time and cost due to an additional process. The sample that will be examined is a non-ferromagnetic material such as aluminum with a good acousto-elastic property, which requires an effective inspection of a bond quality for a nuclear fuel under a manufacturing environment. The purpose of this study is to investigate the feasibility of an Electromagnetic Acoustic Transducer (EMAT) technology for an automated inspection of a nuclear fuel without water

Mixed problem with integral boundary condition for a high order mixed type partial differential equation

OpenAIRE

M. Denche; A. L. Marhoune

2003-01-01

In this paper, we study a mixed problem with integral boundary conditions for a high order partial differential equation of mixed type. We prove the existence and uniqueness of the solution. The proof is based on energy inequality, and on the density of the range of the operator generated by the considered problem.

Status of plutonium recycle from mixed oxide fuel fabrication wastes (U,Pu)O2 facility activities

International Nuclear Information System (INIS)

Quesada, Calixto A.; Adelfang, Pablo; Greiner, G.; Orlando, Oscar S.; Mathot, Sergio R.

1999-01-01

Within the specific subject of mixed oxides corresponding to the Fuel Cycle activities performed at CNEA, the recovery of plutonium from wastes originated during tests and pre-fabrication stages is performed. (author)

Fast reactor fuel pin behavior analyses in a LOF type transient event

International Nuclear Information System (INIS)

Mizuno, Tomoyasu; Koyama, Shin-ichi; Kaito, Takeji; Uwaba, Tomoyuki; Tanaka, Kenya

2013-06-01

In order to evaluate integrity limiting parameters of fuel pins during fast reactor core transient events, such as fuel center line temperature and cladding maximum temperature, fuel pin behavior calculations were made using the fast reactor fuel pin performance code CEDAR. The temperature histories of fuel pins during a loss of flow (LOF) type transient events was calculated based on Ross and Stoute type gap conductance model and constant gap conductance model, which is used in a core transient calculation code like HIPRAC. The calculated maximum temperatures of cladding and adjacent coolant channel were lower in the case with Ross and Stoute type model than in the case of constant gap conductance model due to the dynamic change of gap conductance of former case. It is indicated that core transient calculations with constant gap conductance give conservative cladding and coolant temperatures than that with Ross and Stoute type gap conductance model which is thought to be realistic. (author)

An optimized BWR fuel lattice for improved fuel utilization

International Nuclear Information System (INIS)

Bernander, O.; Helmersson, S.; Schoen, C.G.

1984-01-01

Optimization of the BWR fuel lattice has evolved into the water cross concept, termed ''SVEA'', whereby the improved moderation within bundles augments reactivity and thus improves fuel cycle economy. The novel design introduces into the assembly a cruciform and double-walled partition containing nonboiling water, thus forming four subchannels, each of which holds a 4x4 fuel rod bundle. In Scandinavian BWRs - for which commercial SVEA reloads are now scheduled - the reactivity gain is well exploited without adverse impact in other respects. In effect, the water cross design improves both mechanical and thermal-hydraulic performance. Increased average burnup is also promoted through achieving flatter local power distributions. The fuel utilization savings are in the order of 10%, depending on the basis of comparison, e.g. choice of discharge burnup and lattice type. This paper reviews the design considerations and the fuel utilization benefits of the water cross fuel for non-Scandinavian BWRs which have somewhat different core design parameters relative to ASEA-ATOM reactors. For one design proposal, comparisons are made with current standard 8x8 fuel rod bundles as well as with 9x9 type fuel in reactors with symmetric or asymmetric inter-assembly water gaps. The effect on reactivity coefficients and shutdown margin are estimated and an assessment is made of thermal-hydraulic properties. Consideration is also given to a novel and advantageous way of including mixed-oxide fuel in BWR reloads. (author)

Low grade bioethanol for fuel mixing on gasoline engine using distillation process

Science.gov (United States)

Abikusna, Setia; Sugiarto, Bambang; Suntoro, Dedi; Azami

2017-03-01

Utilization of renewable energy in Indonesia is still low, compared to 34% oil, 20% coal and 20% gas, utilization of energy sources for water 3%, geothermal 1%, 2% biofuels, and biomass 20%. Whereas renewable energy sources dwindling due to the increasing consumption of gasoline as a fuel. It makes us have to look for alternative renewable energy, one of which is bio ethanol. Several studies on the use of ethanol was done to the researchers. Our studies using low grade bio ethanol which begins with the disitillation independently utilize flue gas heat at compact distillator, produces high grade bio ethanol and ready to be mixed with gasoline. Stages of our study is the compact distillator design of the motor dynamic continued with good performance and emission testing and ethanol distilled. Some improvement is made is through the flue gas heat control mechanism in compact distillator using gate valve, at low, medium, and high speed engine. Compact distillator used is kind of a batch distillation column. Column design process using the shortcut method, then carried the tray design to determine the overall geometry. The distillation is done by comparing the separator with a tray of different distances. As well as by varying the volume of the feed and ethanol levels that will feed distilled. In this study, we analyzed the mixing of ethanol through variation between main jet and pilot jet in the carburetor separately interchangeably with gasoline. And finally mixing mechanism bio ethanol with gasoline improved with fuel mixer for performance.

Coordinated safeguards for materials management in a mixed-oxide fuel facility

International Nuclear Information System (INIS)

Shipley, J.P.; Cobb, D.D.; Dietz, R.J.; Evans, M.L.; Schelonka, E.P.; Smith, D.B.; Walton, R.B.

1977-02-01

A coordinated safeguards system is described for safeguarding strategic quantities of special nuclear materials in mixed-oxide recycle fuel fabrication facilities. The safeguards system is compatible with industrial process requirements and combines maximum effectiveness consistent with modest cost and minimal process interference. It is based on unit process accounting using a combination of conventional and state-of-the-art NDA measurement techniques. The effectiveness of the system against single and multiple thefts is evaluated using computer modeling and simulation techniques

Coordinated safeguards for materials management in a mixed-oxide fuel facility

Energy Technology Data Exchange (ETDEWEB)

Shipley, J.P.; Cobb, D.D.; Dietz, R.J.; Evans, M.L.; Schelonka, E.P.; Smith, D.B.; Walton, R.B.

1977-02-01

A coordinated safeguards system is described for safeguarding strategic quantities of special nuclear materials in mixed-oxide recycle fuel fabrication facilities. The safeguards system is compatible with industrial process requirements and combines maximum effectiveness consistent with modest cost and minimal process interference. It is based on unit process accounting using a combination of conventional and state-of-the-art NDA measurement techniques. The effectiveness of the system against single and multiple thefts is evaluated using computer modeling and simulation techniques.

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

International Nuclear Information System (INIS)

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

1993-01-01

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

Fuel/propellant mixing in an open-cycle gas core nuclear rocket engine

International Nuclear Information System (INIS)

Guo, X.; Wehrmeyer, J.A.

1997-01-01

A numerical investigation of the mixing of gaseous uranium and hydrogen inside an open-cycle gas core nuclear rocket engine (spherical geometry) is presented. The gaseous uranium fuel is injected near the centerline of the spherical engine cavity at a constant mass flow rate, and the hydrogen propellant is injected around the periphery of the engine at a five degree angle to the wall, at a constant mass flow rate. The main objective is to seek ways to minimize the mixing of uranium and hydrogen by choosing a suitable injector geometry for the mixing of light and heavy gas streams. Three different uranium inlet areas are presented, and also three different turbulent models (k-var-epsilon model, RNG k-var-epsilon model, and RSM model) are investigated. The commercial CFD code, FLUENT, is used to model the flow field. Uranium mole fraction, axial mass flux, and radial mass flux contours are obtained. copyright 1997 American Institute of Physics

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

International Nuclear Information System (INIS)

Hellwig, Ch.; Kasemeyer, U.

2001-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Hellwig, Ch.; Kasemeyer, U

2001-03-01

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

Fabrication of mixed oxide fuel using plutonium from dismantled weapons

International Nuclear Information System (INIS)

Blair, H.T.; Chidester, K.; Ramsey, K.B.

1996-01-01

A very brief summary is presented of experimental studies performed to support the use of plutonium from dismantled weapons in fabricating mixed oxide (MOX) fuel for commercial power reactors. Thermal treatment tests were performed on plutonium dioxide powder to determine if an effective dry gallium removal process could be devised. Fabrication tests were performed to determine the effects of various processing parameters on pellet quality. Thermal tests results showed that the final gallium content is highly dependent on the treatment temperature. Fabrication tests showed that the milling process, sintering parameters, and uranium feed did effect pellet properties. 1 ref., 1 tab

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

Dissolution behavior of irradiated mixed oxide fuel with short stroke shearing for fast reactor reprocessing

International Nuclear Information System (INIS)

Ikeuchi, Hirotomo; Sano, Yuichi; Shibata, Atsuhiro; Koizumi, Tsutomu; Washiya, Tadahiro

2013-01-01

An efficient dissolution process was established for future reprocessing in which mixed-oxide (MOX) fuels with high plutonium contents and dissolver solution with high heavy-metal (HM) concentrations (more than 500 g dm -3 ) will be treated. This dissolution process involves short stroke shearing of fuels (∼10 mm in length). The dissolution kinetics of irradiated MOX fuels and the effects of the Pu content, HM concentration, and fuel form on the dissolution rate were investigated. Irradiated fuel was found to dissolve as 10 2 -10 3 times fast as non-irradiated fuel, but the rate decreased with increasing Pu content. Kinetic analysis based on the fragmentation model, which considers the penetration and diffusion of nitric acid through fuel matrices prior to chemical reaction, indicated that the dissolution rate of irradiated fuel was affected not only by the volume ratio of liquid to solid (L/S ratio) but also by the exposed surface area per unit mole of nitric acid (A/m ratio). The penetration rate of nitric acid is expected to be decreased at high HM concentrations by a reduction in the L/S ratio, but enhanced by shearing the fuel pieces with short strokes and thus enlarging the A/m ratio. (author)

Quantitative Imaging of Turbulent Mixing Dynamics in High-Pressure Fuel Injection to Enable Predictive Simulations of Engine Combustion

Energy Technology Data Exchange (ETDEWEB)

Frank, Jonathan H. [Sandia National Lab. (SNL-CA), Livermore, CA (United States). Reacting Flows Dept.; Pickett, Lyle M. [Sandia National Lab. (SNL-CA), Livermore, CA (United States). Engine Combustion Dept.; Bisson, Scott E. [Sandia National Lab. (SNL-CA), Livermore, CA (United States). Remote Sensing and Energetic Materials Dept.; Patterson, Brian D. [Sandia National Lab. (SNL-CA), Livermore, CA (United States). combustion Chemistry Dept.; Ruggles, Adam J. [Sandia National Lab. (SNL-CA), Livermore, CA (United States). Reacting Flows Dept.; Skeen, Scott A. [Sandia National Lab. (SNL-CA), Livermore, CA (United States). Engine Combustion Dept.; Manin, Julien Luc [Sandia National Lab. (SNL-CA), Livermore, CA (United States). Engine Combustion Dept.; Huang, Erxiong [Sandia National Lab. (SNL-CA), Livermore, CA (United States). Reacting Flows Dept.; Cicone, Dave J. [Sandia National Lab. (SNL-CA), Livermore, CA (United States). Engine Combustion Dept.; Sphicas, Panos [Sandia National Lab. (SNL-CA), Livermore, CA (United States). Engine Combustion Dept.

2015-09-01

In this LDRD project, we developed a capability for quantitative high - speed imaging measurements of high - pressure fuel injection dynamics to advance understanding of turbulent mixing in transcritical flows, ignition, and flame stabilization mechanisms, and to provide e ssential validation data for developing predictive tools for engine combustion simulations. Advanced, fuel - efficient engine technologies rely on fuel injection into a high - pressure, high - temperature environment for mixture preparation and com bustion. Howe ver, the dynamics of fuel injection are not well understood and pose significant experimental and modeling challenges. To address the need for quantitative high - speed measurements, we developed a Nd:YAG laser that provides a 5ms burst of pulses at 100 kHz o n a robust mobile platform . Using this laser, we demonstrated s patially and temporally resolved Rayleigh scattering imaging and particle image velocimetry measurements of turbulent mixing in high - pressure gas - phase flows and vaporizing sprays . Quantitativ e interpretation of high - pressure measurements was advanced by reducing and correcting interferences and imaging artifacts.

Fuel assemblies for use in BWR type reactors

International Nuclear Information System (INIS)

Hirukawa, Koji.

1987-01-01

Purpose: To moderate the peak configuration of the burnup degree change curve for the infinite multiplication factor by applying an improvement to the arrangement of fuel rods. Constitution: In a fuel assembly for a BWR type reactor comprising a plurality of fuel rods and water rods arranged in a square lattice, fuel rods containing burnable poisons are arranged at four corners at the second and the third layers from the outside of the square lattice arrangement. Among them, the Cd poison effect in the burnable poison incorporated fuel rods disposed at the second layer is somewhat greater at the initial burning stage and then rapidly decreased along with burning. While on the other hand, the poison effect of the burnable poison-incorporated fuel rods at the third layer is smaller than that at the second layer at the initial burning stage and the reduction in the poison effect due to burning is somewhat more moderate. Since these fuel rods are in adjacent with each other, they interfere to each other and also provide an effect of moderating the burning of the burnable poisons. (Takahashi, M.)

Post-irradiation examinations of uranium-plutonium mixed nitride fuel irradiated in JMTR (89F-3A capsule)

International Nuclear Information System (INIS)

Iwai, Takashi; Nakajima, Kunihisa; Kikuchi, Hironobu; Arai, Yasuo; Kimura, Yasuhiko; Nagashima, Hisao; Sekita, Noriaki

2000-03-01

Two helium-bonded fuel pins filled with uranium-plutonium mixed nitride pellets were encapsulated in 89F-3A and irradiated in JMTR up to 5.5% FIMA at a maximum linear power of 73 kW/m. The capsule cooled for ∼5 months was transported to Reactor Fuel Examination Facility and subjected to non-destructive and destructive post irradiation examinations. Any failure was not observed in the irradiated fuel pins. Very low fission gas release rate of about 2 ∼ 3% was observed, while the diametric increase of fuel pin was limited to ∼0.4% at the position of maximum reading. The inner surface of cladding tube did not show any signs of chemical interaction with fuel pellet. (author)

Large-scale fuel cycle centres

International Nuclear Information System (INIS)

Smiley, S.H.; Black, K.M.

1977-01-01

The US Nuclear Regulatory Commission (NRC) has considered the nuclear energy centre concept for fuel cycle plants in the Nuclear Energy Centre Site Survey 1975 (NECSS-75) Rep. No. NUREG-0001, an important study mandated by the US Congress in the Energy Reorganization Act of 1974 which created the NRC. For this study, the NRC defined fuel cycle centres as consisting of fuel reprocessing and mixed-oxide fuel fabrication plants, and optional high-level waste and transuranic waste management facilities. A range of fuel cycle centre sizes corresponded to the fuel throughput of power plants with a total capacity of 50,000-300,000MW(e). The types of fuel cycle facilities located at the fuel cycle centre permit the assessment of the role of fuel cycle centres in enhancing the safeguard of strategic special nuclear materials - plutonium and mixed oxides. Siting fuel cycle centres presents a smaller problem than siting reactors. A single reprocessing plant of the scale projected for use in the USA (1500-2000t/a) can reprocess fuel from reactors producing 50,000-65,000MW(e). Only two or three fuel cycle centres of the upper limit size considered in the NECSS-75 would be required in the USA by the year 2000. The NECSS-75 fuel cycle centre evaluation showed that large-scale fuel cycle centres present no real technical siting difficulties from a radiological effluent and safety standpoint. Some construction economies may be achievable with fuel cycle centres, which offer opportunities to improve waste-management systems. Combined centres consisting of reactors and fuel reprocessing and mixed-oxide fuel fabrication plants were also studied in the NECSS. Such centres can eliminate shipment not only of Pu but also mixed-oxide fuel. Increased fuel cycle costs result from implementation of combined centres unless the fuel reprocessing plants are commercial-sized. Development of Pu-burning reactors could reduce any economic penalties of combined centres. The need for effective fissile

[Fire behavior of ground surface fuels in Pinus koraiensis and Quercus mongolica mixed forest under no wind and zero slope condition: a prediction with extended Rothermel model].

Science.gov (United States)

Zhang, Ji-Li; Liu, Bo-Fei; Chu, Teng-Fei; Di, Xue-Ying; Jin, Sen

2012-06-01

A laboratory burning experiment was conducted to measure the fire spread speed, residual time, reaction intensity, fireline intensity, and flame length of the ground surface fuels collected from a Korean pine (Pinus koraiensis) and Mongolian oak (Quercus mongolica) mixed stand in Maoer Mountains of Northeast China under the conditions of no wind, zero slope, and different moisture content, load, and mixture ratio of the fuels. The results measured were compared with those predicted by the extended Rothermel model to test the performance of the model, especially for the effects of two different weighting methods on the fire behavior modeling of the mixed fuels. With the prediction of the model, the mean absolute errors of the fire spread speed and reaction intensity of the fuels were 0.04 m X min(-1) and 77 kW X m(-2), their mean relative errors were 16% and 22%, while the mean absolute errors of residual time, fireline intensity and flame length were 15.5 s, 17.3 kW X m(-1), and 9.7 cm, and their mean relative errors were 55.5%, 48.7%, and 24%, respectively, indicating that the predicted values of residual time, fireline intensity, and flame length were lower than the observed ones. These errors could be regarded as the lower limits for the application of the extended Rothermel model in predicting the fire behavior of similar fuel types, and provide valuable information for using the model to predict the fire behavior under the similar field conditions. As a whole, the two different weighting methods did not show significant difference in predicting the fire behavior of the mixed fuels by extended Rothermel model. When the proportion of Korean pine fuels was lower, the predicted values of spread speed and reaction intensity obtained by surface area weighting method and those of fireline intensity and flame length obtained by load weighting method were higher; when the proportion of Korean pine needles was higher, the contrary results were obtained.

Continuous spin fields of mixed-symmetry type

Science.gov (United States)

Alkalaev, Konstantin; Grigoriev, Maxim

2018-03-01

We propose a description of continuous spin massless fields of mixed-symmetry type in Minkowski space at the level of equations of motion. It is based on the appropriately modified version of the constrained system originally used to describe massless bosonic fields of mixed-symmetry type. The description is shown to produce generalized versions of triplet, metric-like, and light-cone formulations. In particular, for scalar continuous spin fields we reproduce the Bekaert-Mourad formulation and the Schuster-Toro formulation. Because a continuous spin system inevitably involves infinite number of fields, specification of the allowed class of field configurations becomes a part of its definition. We show that the naive choice leads to an empty system and propose a suitable class resulting in the correct degrees of freedom. We also demonstrate that the gauge symmetries present in the formulation are all Stueckelberg-like so that the continuous spin system is not a genuine gauge theory.

Fuel elements based on mixed oxides UO{sub 2} - PuO{sub 2}; Gorivni elementi na bazi mesanih oksida UO{sub 2} - PuO{sub 2}

Energy Technology Data Exchange (ETDEWEB)

Katanic-Popovic, J; Stevanovic, M [Boris Kidric Institute of nuclear sciences, Vinca, Belgrade (Yugoslavia)

1978-07-01

Questions concerning utilization of plutonium as a fissionable material in fuel elements for nuclear power plants have been discussed. Characteristics and application of fuel elements with mixed UO{sub 2} - PuO{sub 2} fuel for thermal and fast breeder reactors have also been dealt with. In the presentation of technological processes for production of fuel elements based on mixed oxides specific characteristics are given with respect to the work with plutonium and relatively high production costs as compared to classical fuel elements based on sintered UO{sub 2}. (author)

Effect of engine parameters and gaseous fuel type on the cyclic variability of dual fuel engines

Energy Technology Data Exchange (ETDEWEB)

Mohamed Y.E. Selim [United Arab Emirates University, Al-Ain (United Arab Emirates). Mechanical Engineering Department, Faculty of Engineering

2005-05-01

This paper presents an analysis of the cycle-to-cycle combustion variation as reflected in the combustion pressure data of a single cylinder, naturally aspirated, four stroke, Ricardo E6 engine converted to run as dual fuel engine on diesel and gaseous fuel of LPG or methane. A measuring set-up consisting of a piezo-electric pressure transducer with charge amplifier and fast data acquisition card installed on an IBM microcomputer was used to gather the data of up to 1200 consecutive combustion cycles of the cylinder under various combination of engine operating and design parameters. These parameters included type of gaseous fuel, engine load, compression ratio, pilot fuel injection timing, pilot fuel mass, and engine speed. The data for each operating conditions were analyzed for the maximum pressure, the maximum rate of pressure rise representing the combustion noise, and indicated mean effective pressure. The cycle-to-cycle variation is expressed as the mean value, standard deviation, and coefficient of variation of these three parameters. It was found that the type of gaseous fuel and engine operating and design parameters affected the combustion noise and its cyclic variation and these effects have been presented. 21 refs., 6 figs., 1 tab.

Fluid mixing and the deep biosphere of a fossil Lost City-type hydrothermal system at the Iberia Margin.

Science.gov (United States)

Klein, Frieder; Humphris, Susan E; Guo, Weifu; Schubotz, Florence; Schwarzenbach, Esther M; Orsi, William D

2015-09-29

Subseafloor mixing of reduced hydrothermal fluids with seawater is believed to provide the energy and substrates needed to support deep chemolithoautotrophic life in the hydrated oceanic mantle (i.e., serpentinite). However, geosphere-biosphere interactions in serpentinite-hosted subseafloor mixing zones remain poorly constrained. Here we examine fossil microbial communities and fluid mixing processes in the subseafloor of a Cretaceous Lost City-type hydrothermal system at the magma-poor passive Iberia Margin (Ocean Drilling Program Leg 149, Hole 897D). Brucite-calcite mineral assemblages precipitated from mixed fluids ca. 65 m below the Cretaceous paleo-seafloor at temperatures of 31.7 ± 4.3 °C within steep chemical gradients between weathered, carbonate-rich serpentinite breccia and serpentinite. Mixing of oxidized seawater and strongly reducing hydrothermal fluid at moderate temperatures created conditions capable of supporting microbial activity. Dense microbial colonies are fossilized in brucite-calcite veins that are strongly enriched in organic carbon (up to 0.5 wt.% of the total carbon) but depleted in (13)C (δ(13)C(TOC) = -19.4‰). We detected a combination of bacterial diether lipid biomarkers, archaeol, and archaeal tetraethers analogous to those found in carbonate chimneys at the active Lost City hydrothermal field. The exposure of mantle rocks to seawater during the breakup of Pangaea fueled chemolithoautotrophic microbial communities at the Iberia Margin, possibly before the onset of seafloor spreading. Lost City-type serpentinization systems have been discovered at midocean ridges, in forearc settings of subduction zones, and at continental margins. It appears that, wherever they occur, they can support microbial life, even in deep subseafloor environments.

Mixed problem with nonlocal boundary conditions for a third-order partial differential equation of mixed type

OpenAIRE

Denche, M.; Marhoune, A. L.

2001-01-01

We study a mixed problem with integral boundary conditions for a third-order partial differential equation of mixed type. We prove the existence and uniqueness of the solution. The proof is based on two-sided a priori estimates and on the density of the range of the operator generated by the considered problem.

Power distribution gradients in WWER type cores and fuel failure root causes

Energy Technology Data Exchange (ETDEWEB)

Mikuš, Ján M., E-mail: JanMikus.nrc@hotmail.com

2014-02-15

Highlights: • Power (fission rate) distribution gradients can represent fuel failure root causes. • Positions with above gradients were investigated in WWER type cores on reactor LR-0. • Above gradients were evaluated near core heterogeneities and construction materials. • Results can be used for code validation and fuel failure occurrence investigation. - Abstract: Neutron flux non-uniformity and gradients of neutron current resulting in corresponding power (fission rate) distribution changes can represent root causes of the fuel failure. Such situation can be expected in vicinity of some core heterogeneities and construction materials. Since needed data cannot be obtained from nuclear power plant (NPP), results of some benchmark type experiments performed on light water, zero-power research reactor LR-0 were used for investigation of the above phenomenon. Attention was focused on determination of the spatial power distribution changes in fuel assemblies (FAs): Containing fuel rods (FRs) with Gd burnable absorber in WWER-440 and WWER-1000 type cores, Neighboring the core blanket and dummy steel assembly simulators on the periphery of the WWER-440 standard and low leakage type cores, resp., Neighboring baffle in WWER-1000 type cores, and Neighboring control rod (CR) in WWER-440 type cores, namely (a) power peak in axial power distribution in periphery FRs of the adjacent FAs near the area between CR fuel part and butt joint to the CR absorbing part and (b) decrease in radial power distribution in FRs near CR absorbing part. An overview of relevant experimental results from reactor LR-0 and some information concerning leaking FAs on NPP Temelín are presented. Obtained data can be used for code validation and subsequently for the fuel failure occurrence investigation.

Heat transfer coefficient testing in nuclear fuel rod bundles with mixing vane grids

International Nuclear Information System (INIS)

Conner, Michael E.; Smith, L. David III; Holloway, Mary V.; Beasley, Donald E.

2005-01-01

An air heat transfer test facility was developed to test the heat transfer downstream of support grids in simulated PWR nuclear fuel rod bundles. The goal of this testing is to study the single-phase heat transfer coefficients downstream of grids with mixing vanes in a square-pitch rod bundle. The technique developed utilizes fully-heated grid spans and a specially designed thermocouple holder that can be moved axially down the rod bundle and aximuthally within a test rod. From this testing, the axial and aximuthally varying heat transfer coefficient can be determined. Different grid designs are tested and compared to determine the heat transfer enhancement associated with key grid features such as mixing vanes. (author)

Licensing issues associated with the use of mixed-oxide fuel in U.S. commercial nuclear reactors

International Nuclear Information System (INIS)

Williams, D.L. Jr.

1997-04-01

On January 14, 1997, the Department of Energy, as part of its Record of Decision on the storage and disposition of surplus nuclear weapons materials, committed to pursue the use of excess weapons-usable plutonium in the fabrication of mixed-oxide (MOX) fuel for consumption in existing commercial nuclear power plants. Domestic use of MOX fuel has been deferred since the late 1970s, principally due to nuclear proliferation concerns. This report documents a review of past and present literature (i.e., correspondence, reports, etc.) on the domestic use of MOX fuel and provides discussion on the technical and regulatory issues that must be addressed by DOE (and the utility/consortia selected by DOE to effect the MOX fuel consumption strategy) in obtaining approval from the Nuclear Regulatory Commission to use MOX fuel in one or a group of existing commercial nuclear power plants

Fuel variability following wildfire in forests with mixed severity fire regimes, Cascade Range, USA

Science.gov (United States)

Jessica L. Hudec; David L. Peterson

2012-01-01

Fire severity influences post-burn structure and composition of a forest and the potential for a future fire to burn through the area. The effects of fire on forests with mixed severity fire regimes are difficult to predict and interpret because the quantity, structure, and composition of forest fuels vary considerably. This study examines the relationship between fire...

MOX fuel assembly and reactor core

International Nuclear Information System (INIS)

Shimada, Hidemitsu; Koyama, Jun-ichi; Aoyama, Motoo

1998-01-01

The MOX fuel assembly of the present invention is of a c-lattice type loaded to a BWR type reactor. 74 MOX fuel rods filled with mixed oxides of uranium and plutonium and two water rods disposed to a space equal to that for 7 MOX fuel rods are arranged in 9 x 9 matrix. MOX fuel rods having the lowest enrichment degree are disposed to four corners of the 9 x 9 matrix. The enrichment degree means a ratio of the weight of fission products based on the total weight of fuels. Two MOX fuel rods having the same enrichment degree are arranged in each direction so as to be continuous from the MOX fuel rods at four corners in the direction of the same row and different column and same column and the different row. In addition, among the outermost circumferential portion of the 9 x 9 matrix, MOX fuel rods having a lower enrichment degree next to the MOX fuel rods having the lowest enrichment degree are arranged, each by three to a portion where MOX fuel rods having the lowest enrichment degree are not disposed. (I.N.)

Implement of MOX fuel assemblies in the design of the fuel reload for a BWR; Implemento de ensambles de combustible MOX en el diseno de la recarga de combustible para un BWR

Energy Technology Data Exchange (ETDEWEB)

Enriquez C, P.; Ramirez S, J. R.; Alonso V, G.; Palacios H, J. C., E-mail: pastor.enriquez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2011-11-15

At the present time the use of mixed oxides as nuclear fuel is a technology that has been implemented in mixed reloads of fuel for light water reactors. Due to the plutonium production in power reactors, is necessary to realize a study that presents the plutonium use like nuclear fuel. In this work a study is presented that has been carried out on the design of a fuel assembly with MOX to be proposed in the supply of a fuel reload. The fissile relationship of uranium to plutonium is presented for the design of the MOX assembly starting from plutonium recovered in the reprocessing of spent fuel and the comparison of the behavior of the infinite multiplication factor is presented and of the local power peak factor, parameters of great importance in the fuel assemblies design. The study object is a fuel assembly 10 x 10 GNF2 type for a boiling water reactor. The design of the fuel reload pattern giving fuel assemblies with MOX, so the comparison of the behavior of the stop margin for a fuel reload with UO{sub 2} and a mixed reload, implementing 12 and 16 fuel assemblies with MOX are presented. The results show that the implement of fuel assemblies with MOX in a BWR is possible, but this type of fuels creates new problems that are necessary to study with more detail. In the development of this work the calculus tools were the codes: INTREPIN-3, CASMO-4, CMSLINK and SIMULATE-3. (Author)

Measurements and observations on microscopic swelling in MX-type fuels

International Nuclear Information System (INIS)

Ronchi, C.; Ray, I.L.F.; Thiele, H.; Laar, J. van de.

1978-01-01

Microscopic swelling has been investigated by electron microscopy in several MX-type fuels, irradiated in fast and thermal neutron flux. The results show that fission gas bubbles in these compounds grow to large sizes if the in-pile fuel temperature rises above a critical value (swelling critical temperature Tsub(C)). A comparison has been made of the swelling rates in fuels of different composition, showing that Tsub(C) increases from carbides to nitrides. In fuels subjected to in-pile restructuring (highly rated) He-bonded pins microscopic swelling is affected by pore and grain boundary migration. The influence of these phenomena on the fuel swelling performance has been discussed

Performance of IN-706 and PE-16 cladding in mixed-oxide fuel pins

International Nuclear Information System (INIS)

Makenas, B.J.; Lawrence, L.A.; Jensen, B.W.

1982-05-01

Iron-nickel base, precipitation-strengthened alloys, IN-706 and PE-16, advanced alloy cladding considered for breeder reactor applications, were irradiated in mixed-oxide fuel pins in the HEDL-P-60 subassembly in EBR-II. Initial selection of candidate advanced alloys was done using only nonfueled materials test results. However, to establish the performance characteristics of the candidate cladding alloys, i.e., dimensional stability and structural integrity under conditions of high neutron flux, elevated temperature, and applied stress, it was necessary to irradiate fuel pins under typical operating conditions. Fuel pins were clad with solution treated IN-706 and PE-16 and irradiated to peak fluences of 6.1 x 10 22 n/cm 2 (E > .1 MeV) and 8.8 x 10 22 n/cm 2 (E > .1 MeV) respectively. Fabrication and operating parameters for the fuel pins with the advanced cladding alloy candidates are summarized. Irradiation of HEDL-P-60 was interrupted with the breach of a pin with IN-706 cladding at 5.1 at % and the test was terminated with cladding breach in a pin with PE-16 cladding at 7.6 at %

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)

An Investigation of Fuel Mixing and Reaction in a CH4/Syngas/Air Premixed Impinging Flame with Varied H2/CO Proportion

Directory of Open Access Journals (Sweden)

Chih-Pin Chiu

2017-07-01

Full Text Available For industrial applications, we propose a concept of clean and efficient combustion through burning syngas on an impinging burner. We performed experimental measurements of particle image velocimetry, OH radical (OH* chemiluminescence, flame temperature, and CO emission to examine the fuel mixing and reaction of premixed impinging flames of CH4/syngas/air with H2/CO in varied proportions. The velocity distribution of the combustion flow field showed that a deceleration area in the main flow formed through the mutual impingement of two jet flows, which enhanced the mixing of fuel and air because of an increased momentum transfer. The deceleration area expanded with an increased CO proportion, which indicated that the mixing of fuel and air also increased with the increased CO proportion. Our examination of the OH* chemiluminescence demonstrated that its intensity increased with increased CO proportion, which showed that the reaction between fuel and air accordingly increased. CO provided in the syngas hence participated readily in the reaction of the CH4/syngas/air premixed impinging flames when the syngas contained CO in a large proportion. Although the volume flow rate of the provided CO quadrupled, the CO emission increased by only 12% to 15%. The results of this work are useful to improve the feasibility of fuel-injection systems using syngas as an alternative fuel.

Large and almost maximal neutrino mixing within the type II see-saw mechanism

International Nuclear Information System (INIS)

Lindner, Manfred; Rodejohann, Werner

2007-01-01

Within the type II see-saw mechanism the light neutrino mass matrix is given by a sum of a direct (or triplet) mass term and the conventional (type I) see-saw term. Both versions of the see-saw mechanism explain naturally small neutrino masses, but the type II scenario offers interesting additional possibilities to explain large or almost maximal or vanishing mixings which are discussed in this paper. We first introduce 'type II enhancement' of neutrino mixing, where moderate cancellations between the two terms can lead to large neutrino mixing even if all individual mass matrices and terms generate small mixing. However, nearly maximal or vanishing mixings are not naturally explained in this way, unless there is a certain initial structure (symmetry) which enforces certain elements of the matrices to be identical or related in a special way. We therefore assume that the leading structure of the neutrino mass matrix is the triplet term and corresponds to zero U e3 and maximal θ 23 . Small but necessary corrections are generated by the conventional see-saw term. Then we assume that one of the two terms corresponds to an extreme mixing scenario, such as bimaximal or tri-bimaximal mixing. Deviations from this scheme are introduced by the second term. One can mimic Quark-Lepton Complementarity in this way. Finally, we note that the neutrino mass matrix for tri-bimaximal mixing can be-depending on the mass hierarchy-written as a sum of two terms with simple structure. Their origin could be the two terms of type II see-saw

Fuel management of HTR-10

International Nuclear Information System (INIS)

Wu Zongxin; Jing Xingqing

2001-01-01

The 10 MW high temperature cooled reactor (HTR-10) built in Tsinghua University is a pebble bed type of HTGR. The continuous recharge and multiple-pass of spherical fuel elements are used for fuel management. The initiative stage of core is composed of the mix of spherical fuel elements and graphite elements. The equilibrium stage of core is composed of identical spherical fuel elements. The fuel management during the transition from the initiative stage to the equilibrium stage is a key issue for HTR-10 physical design. A fuel management strategy is proposed based on self-adjustment of core reactivity. The neutron physical code is used to simulate the process of fuel management. The results show that the graphite elements, the recharging fuel elements below the burn-up allowance, and the discharging fuel elements over the burn-up allowance could be identified by burn-up measurement. The maximum of burn-up fuel elements could be controlled below the burn-up limit

7 CFR 52.1858 - Grades of mixed types or varieties of raisins.

Science.gov (United States)

2010-01-01

... AGRICULTURAL MARKETING ACT OF 1946 PROCESSED FRUITS AND VEGETABLES, PROCESSED PRODUCTS THEREOF, AND CERTAIN OTHER PROCESSED FOOD PRODUCTS 1 United States Standards for Grades of Processed Raisins 1 Type Vi-Mixed... 7 Agriculture 2 2010-01-01 2010-01-01 false Grades of mixed types or varieties of raisins. 52.1858...

Re-qualification of MTR-type fuel plates fabrication process

International Nuclear Information System (INIS)

Elseaidy, I.M.; Ghoneim, M.M.

2010-01-01

The fabricability issues with increased uranium loading due to use low enrichment of uranium (LEU), i.e. less than 20 % of U 235 , increase the problems which occur during compact manufacturing, roll bonding of the fuel plates, potential difficulty in forming during rolling process, mechanical integrity of the core during fabrication, potential difficulty in meat homogeneity, and the ability to fabricate plates with thicker core as a means of increasing total uranium loading. To produce MTR- type fuel plates with high uranium loading (HUL) and keep the required quality of these plates, many of qualification process must be done in the commissioning step of fuel fabrication plant. After that any changing of the fabrication parameters, for example changing of any of the raw materials, devises, operators, and etc., a re- qualification process should be done in order to keep the quality of produced plates. Objective of the present work is the general description of the activities to be accomplished for re-qualification of manufacturing MTR- type nuclear fuel plates. For each process to be re-qualified, a detailed of re-qualification process were established. (author)

76 FR 65544 - Standard Format and Content of License Applications for Mixed Oxide Fuel Fabrication Facilities

Science.gov (United States)

2011-10-21

... NUCLEAR REGULATORY COMMISSION [NRC-2009-0323] Standard Format and Content of License Applications... revision to regulatory guide (RG) 3.39, ``Standard Format and Content of License Applications for Mixed Oxide Fuel Fabrication Facilities.'' This guide endorses the standard format and content for license...

Fuel switching in Harare : An almost ideal demand system approach

NARCIS (Netherlands)

Chambwera, Muyeye; Folmer, Henk

In urban areas several energy choices are available and the amount of (a given type of) fuel consumed is based on complex household decision processes. This paper analyzes urban fuel (particularly firewood) demand in an energy mix context by means of an Almost Ideal Demand System based on a survey

Fuel switching in Harare: An almost ideal demand system approach

NARCIS (Netherlands)

Chambwera, M.; Folmer, H.

2007-01-01

In urban areas several energy choices are available and the amount of (a given type of) fuel consumed is based on complex household decision processes. This paper analyzes urban fuel (particularly firewood) demand in an energy mix context by means of an Almost Ideal Demand System based on a survey

Municipal solid waste combustion: Fuel testing and characterization

Energy Technology Data Exchange (ETDEWEB)

Bushnell, D.J.; Canova, J.H.; Dadkhah-Nikoo, A.

1990-10-01

The objective of this study is to screen and characterize potential biomass fuels from waste streams. This will be accomplished by determining the types of pollutants produced while burning selected municipal waste, i.e., commercial mixed waste paper residential (curbside) mixed waste paper, and refuse derived fuel. These materials will be fired alone and in combination with wood, equal parts by weight. The data from these experiments could be utilized to size pollution control equipment required to meet emission standards. This document provides detailed descriptions of the testing methods and evaluation procedures used in the combustion testing and characterization project. The fuel samples will be examined thoroughly from the raw form to the exhaust emissions produced during the combustion test of a densified sample.

CFD thermal-hydraulic analysis of a CANDU fuel channel with SEU43 type fuel bundle

International Nuclear Information System (INIS)

Catana, A.; Prisecaru, Ilie; Dupleac, D.; Danila, Nicolae

2009-01-01

This paper presents the numerical investigation of a CANDU fuel channel using CFD (Computational Fluid Dynamics) methodology approach, when SEU43 fuel bundles are used. Comparisons with STD37 fuel bundles are done in order to evaluate the influence of geometrical differences of the fuel bundle types on fluid flow properties. We adopted a strategy to analyze only the significant segments of fuel channel, namely : - the fuel bundle junctions with adjacent segments; - the fuel bundle spacer planes with adjacent segments; - the fuel bundle segments with turbulence enhancement buttons; - and the regular segments of fuel bundles. The computer code used is an academic version of FLUENT code, available from UPB. The complex flow domain of fuel bundles contained in pressure tube and operating conditions determine a high turbulence flow and in some parts of fuel channel also a multi-phase flow. Numerical simulation of the flow in the fuel channel has been achieved by solving the equations for conservation of mass, momentum and energy. For turbulence model the standard k-model is employed although other turbulence models can be used. In this paper we do not consider heat generation and heat transfer capabilities of CFD methods. Boundary conditions for CFD analysis are provided by system and sub-channel analysis. In this paper the discussion is focused on some flow parameters behaviour at the bundle junction, spacer's plane configuration, etc. of a SEU43 fuel bundle in conditions of a typical CANDU 6 fuel channel starting from some experience gained in a previous work. (authors)

Utilization of thorium in PWR type reactors

International Nuclear Information System (INIS)

Correa, F.

1977-01-01

Uranium 235 consumption is comparatively evaluated with thorium cycle for a PWR type reactor. Modifications are only made in fuels components. U-235 consumption is pratically unchanged in both cycles. Some good results are promised to the mixed U-238/Th-232 fuel cycle in 1/1 proportion [pt

End plug welding of nuclear fuel elements-AFFF experience

International Nuclear Information System (INIS)

Bhatt, R.B.; Singh, S.; Aniruddha Kumar; Amit; Arun Kumar; Panakkal, J.P.; Kamath, H.S.

2004-01-01

Advanced Fuel Fabrication Facility is engaged in the fabrication of mixed oxide (U,Pu)O 2 fuel elements of various types of nuclear reactors. Fabrication of fuel elements involves pellet fabrication, stack making, stack loading and end plug welding. The requirement of helium bonding gas inside the fuel elements necessitates the top end plug welding to be carried out with helium as the shielding gas. The severity of the service conditions inside a nuclear reactor imposes strict quality control criteria, which demands for almost defect free welds. The top end plug welding being the last process step in fuel element fabrication, any rejection at this stage would lead to loss of effort prior to this step. Moreover, the job becomes all the more difficult with mixed oxide (MOX) as the entire fabrication work has to be carried out in glove box trains. In the case of weld rejection, accepted pellets are salvaged by cutting the clad tube. This is a difficult task and recovery of pellets is low (requiring scrap recovery operation) and also leads to active metallic waste generation. This paper discusses the experience gained at AFFF, in the past 12 years in the area of end plug welding for different types of MOX fuel elements

Large-scale fuel cycle centers

International Nuclear Information System (INIS)

Smiley, S.H.; Black, K.M.

1977-01-01

The United States Nuclear Regulatory Commission (NRC) has considered the nuclear energy center concept for fuel cycle plants in the Nuclear Energy Center Site Survey - 1975 (NECSS-75) -- an important study mandated by the U.S. Congress in the Energy Reorganization Act of 1974 which created the NRC. For the study, NRC defined fuel cycle centers to consist of fuel reprocessing and mixed oxide fuel fabrication plants, and optional high-level waste and transuranic waste management facilities. A range of fuel cycle center sizes corresponded to the fuel throughput of power plants with a total capacity of 50,000 - 300,000 MWe. The types of fuel cycle facilities located at the fuel cycle center permit the assessment of the role of fuel cycle centers in enhancing safeguarding of strategic special nuclear materials -- plutonium and mixed oxides. Siting of fuel cycle centers presents a considerably smaller problem than the siting of reactors. A single reprocessing plant of the scale projected for use in the United States (1500-2000 MT/yr) can reprocess the fuel from reactors producing 50,000-65,000 MWe. Only two or three fuel cycle centers of the upper limit size considered in the NECSS-75 would be required in the United States by the year 2000 . The NECSS-75 fuel cycle center evaluations showed that large scale fuel cycle centers present no real technical difficulties in siting from a radiological effluent and safety standpoint. Some construction economies may be attainable with fuel cycle centers; such centers offer opportunities for improved waste management systems. Combined centers consisting of reactors and fuel reprocessing and mixed oxide fuel fabrication plants were also studied in the NECSS. Such centers can eliminate not only shipment of plutonium, but also mixed oxide fuel. Increased fuel cycle costs result from implementation of combined centers unless the fuel reprocessing plants are commercial-sized. Development of plutonium-burning reactors could reduce any

Model development of UO_2-Zr dispersion plate-type fuel behavior at early phase of severe accident and molten fuel meat relocation

International Nuclear Information System (INIS)

Zhang Zhuohua; Yu Junchong; Peng Shinian

2014-01-01

According to former study on oxygen diffusion, Nb-Zr solid reaction and UO_2-Zr solid reaction, the models of oxidation, solid reaction in fuel meat and relocation of molten fuel meat are developed based on structure and material properties of UO_2-Zr dispersion plate-type fuel, The new models can supply theoretical elements for the safety analysis of the core assembled with dispersion plate-type fuel under severe accident. (authors)

Physical properties, evaporation and combustion characteristics of nanofluid-type fuels

OpenAIRE

Tanvir, Saad

2016-01-01

Nanofluids are liquids with stable suspension of nanoparticles. Limited studies in the past have shown that both energetic and catalytic nanoparticles once mixed with traditional liquid fuels can be advantageous in combustion applications, e.g., increased energy density and shortened ignition delay. Contradictions in existing literature, scarcity of experimental data and lack of understanding on how the added nanoparticles affect the physical properties as well as combustion characteristics o...

Formation of industrial mixed culture biofilm in chlorophenol cultivated medium of microbial fuel cell

Science.gov (United States)

Hassan, Huzairy; Jin, Bo; Dai, Sheng; Ngau, Cornelius

2016-11-01

The formation of microbial biofilm while maintaining the electricity output is a challenging topic in microbial fuel cell (MFC) studies. This MFC critical factor becomes more significant when handling with industrial wastewater which normally contains refractory and toxic compounds. This study explores the formation of industrial mixed culture biofilm in chlorophenol cultivated medium through observing and characterizing microscopically its establishment on MFC anode surface. The mixed culture was found to develop its biofilm on the anode surface in the chlorophenol environment and established its maturity and dispersal stages with concurrent electricity generation and phenolic degradation. The mixed culture biofilm engaged the electron transfer roles in MFC by generating current density of 1.4 mA/m2 and removing 53 % of 2,4-dichlorophenol. The results support further research especially on hazardous wastewater treatment using a benign and sustainable method.

Refining fuels of the heavy gas--oil type

Energy Technology Data Exchange (ETDEWEB)

Bruzac, J F.A.

1930-01-28

This invention has for its object the production of a new type of gas-oil fuel, obtained from crude petroleum, shale oil, and peat oil, according to the method of treatment mentioned, by means of which is obtained from gas oil, shale oil, lignite oil, and peat oil (deprived of asphaltic, and bituminous, resinous, and sulfur compounds), a fuel suitable for running Diesel, Junkers, and Clerget motors and all others of the same kind, by diminishing considerably the fouling and attack on the metal.

Automation of potentiometric titration for the determination of uranium in nuclear fuel materials

International Nuclear Information System (INIS)

Kelkar, Anoop; Pandey, Ashish; Kapoor, Y.S.; Kumar, Manish; Singh, Mamta; Fulzele, Ajeet; Prakash, Amrit; Afzal, Mohd; Panakkal, J.P.

2010-01-01

Advanced Fuel Fabrication Facility is fabricating various types of mixed oxide fuels, namely for PHWR, BWR, FBTR and PFBR. Precise determination of uranium in MOX fuel sample is important to get desired burn up in the reactor. The modified Davies and Gray method is routinely used for the potentiometric titration of uranium

Study of brushless fuel pump (improvement of pump and motor parts). 2nd Report. Blushless dendo fuel pump no kento. 2

Energy Technology Data Exchange (ETDEWEB)

Mine, K; Takada, S; Tatematsu, M; Takeuchi, H [Aisan Industry Co. Ltd., Aichi (Japan)

1992-10-01

A methanol use electrically driven fuel pump was developed as reported in the present report. Mixed fuel of gasoline with alcohol can be handled by a brushless fuel pump which was proposed and improved as reported. The flow rate performance was heightened to 25g/sec by heightening in output power of motor, while the high temperature performance was 17% heightened against the conventional ratio of lowering in flow rate by heightening in vapor jet capacity. Against the corrosiveness of methanol, an in-tank type was applied to the pump, and all its electrically conductive and other mechanical parts were made to be both anti-corrosive and anti-abrasive. It is structurally of a two-stage series turbine type of non-volume form. A sensor method was applied to the motor by confining the miniaturized control circuit of brushless motor in the motor so that the transistor is controlled against the heightening in temperature. The motor is a three-phase half-wave driving motor. Also developed was a fuel supply system which is useful for the mixed fuel covering a range of 100% methanol through 100% gasoline. The present pump is dimensionally interchangeable with the conventional gasoline use one. Its operational life is more than 10000 hours. 3 refs., 17 figs., 1 tab.

Partnership Selection Involving Mixed Types of Uncertain Preferences

Directory of Open Access Journals (Sweden)

Li-Ching Ma

2013-01-01

Full Text Available Partnership selection is an important issue in management science. This study proposes a general model based on mixed integer programming and goal-programming analytic hierarchy process (GP-AHP to solve partnership selection problems involving mixed types of uncertain or inconsistent preferences. The proposed approach is designed to deal with crisp, interval, step, fuzzy, or mixed comparison preferences, derive crisp priorities, and improve multiple solution problems. The degree of fulfillment of a decision maker’s preferences is also taken into account. The results show that the proposed approach keeps more solution ratios within the given preferred intervals and yields less deviation. In addition, the proposed approach can treat incomplete preference matrices with flexibility in reducing the number of pairwise comparisons required and can also be conveniently developed into a decision support system.

Eliciting mixed emotions: A meta-analysis comparing models, types and measures.

Directory of Open Access Journals (Sweden)

Raul eBerrios

2015-04-01

Full Text Available The idea that people can experience two oppositely valenced emotions has been controversial ever since early attempts to investigate the construct of mixed emotions. This meta-analysis examined the robustness with which mixed emotions have been elicited experimentally. A systematic literature search identified 63 experimental studies that instigated the experience of mixed emotions. Studies were distinguished according to the structure of the underlying affect model – dimensional or discrete – as well as according to the type of mixed emotions studied (e.g., happy-sad, fearful-happy, positive-negative. The meta-analysis using a random-effects model revealed a moderate to high effect size for the elicitation of mixed emotions (dIG+ = .77, which remained consistent regardless of the structure of the affect model, and across different types of mixed emotions. Several methodological and design moderators were tested. Studies using the minimum index (i.e., the minimum value between a pair of opposite valenced affects resulted in smaller effect sizes, whereas subjective measures of mixed emotions increased the effect sizes. The presence of more women in the samples was also associated with larger effect sizes. The current study indicates that mixed emotions are a robust, measurable and non-artifactual experience. The results are discussed in terms of the implications for an affect system that has greater versatility and flexibility than previously thought.

Eliciting mixed emotions: a meta-analysis comparing models, types, and measures

Science.gov (United States)

Berrios, Raul; Totterdell, Peter; Kellett, Stephen

2015-01-01

The idea that people can experience two oppositely valenced emotions has been controversial ever since early attempts to investigate the construct of mixed emotions. This meta-analysis examined the robustness with which mixed emotions have been elicited experimentally. A systematic literature search identified 63 experimental studies that instigated the experience of mixed emotions. Studies were distinguished according to the structure of the underlying affect model—dimensional or discrete—as well as according to the type of mixed emotions studied (e.g., happy-sad, fearful-happy, positive-negative). The meta-analysis using a random-effects model revealed a moderate to high effect size for the elicitation of mixed emotions (dIG+ = 0.77), which remained consistent regardless of the structure of the affect model, and across different types of mixed emotions. Several methodological and design moderators were tested. Studies using the minimum index (i.e., the minimum value between a pair of opposite valenced affects) resulted in smaller effect sizes, whereas subjective measures of mixed emotions increased the effect sizes. The presence of more women in the samples was also associated with larger effect sizes. The current study indicates that mixed emotions are a robust, measurable and non-artifactual experience. The results are discussed in terms of the implications for an affect system that has greater versatility and flexibility than previously thought. PMID:25926805

Fuel type characterization and potential fire behavior estimation in Sardinia and Corsica islands

Science.gov (United States)

Bacciu, V.; Pellizzaro, G.; Santoni, P.; Arca, B.; Ventura, A.; Salis, M.; Barboni, T.; Leroy, V.; Cancellieri, D.; Leoni, E.; Ferrat, L.; Perez, Y.; Duce, P.; Spano, D.

2012-04-01

Wildland fires represent a serious threat to forests and wooded areas of the Mediterranean Basin. As recorded by the European Commission (2009), during the last decade Southern Countries have experienced an annual average of about 50,000 forest fires and about 470,000 burned hectares. The factor that can be directly manipulated in order to minimize fire intensity and reduce other fire impacts, such as three mortality, smoke emission, and soil erosion, is wildland fuel. Fuel characteristics, such as vegetation cover, type, humidity status, and biomass and necromass loading are critical variables in affecting wildland fire occurrence, contributing to the spread, intensity, and severity of fires. Therefore, the availability of accurate fuel data at different spatial and temporal scales is needed for fire management applications, including fire behavior and danger prediction, fire fighting, fire effects simulation, and ecosystem simulation modeling. In this context, the main aims of our work are to describe the vegetation parameters involved in combustion processes and develop fire behavior fuel maps. The overall work plan is based firstly on the identification and description of the different fuel types mainly affected by fire occurrence in Sardinia (Italy) and Corsica (France) Islands, and secondly on the clusterization of the selected fuel types in relation to their potential fire behavior. In the first part of the work, the available time series of fire event perimeters and the land use map data were analyzed with the purpose of identifying the main land use types affected by fires. Thus, field sampling sites were randomly identified on the selected vegetation types and several fuel variables were collected (live and dead fuel load partitioned following Deeming et al., (1977), depth of fuel layer, plant cover, surface area-to-volume ratio, heat content). In the second part of the work, the potential fire behavior for every experimental site was simulated using

Study of UO2-10WT%Gd2O3 fuel pellets obtained by seeding method using AUC co-precipitation and mechanical mixing processes

International Nuclear Information System (INIS)

Lima, M.M.F.; Ferraz, W.B.A.; Santos, M.M. dos; Pinto, L.C.M.; Santos, A.

2008-01-01

The use of gadolinium and uranium mixed oxide as a nuclear fuel aims to obtain a fuel with a performance better than that of UO 2 fuel. In this work, seeding method was used to improve ionic diffusivity during sintering to produce high density pellets containing coarse grains by co-precipitation and mechanical mixing processes. Sintered UO 2 -10 wt% Gd 2 O 3 pellets were obtained using the reference processes with 2 wt% and 5 wt% UO 2 seeds with two granulometries, less than 20 μm and between 20 and 38 μm. Characterisation was carried out by chemical analysis, surface area, X-ray diffraction, SEM, WDS, image analysis, and densitometry. The seeding method using mechanical mixing process was more effective than the co-precipitation method. Furthermore, mechanical mixing process resulted in an increase in density of UO 2 -10wt% Gd 2 O 3 with seeds in relation to that of UO 2 -10wt% Gd 2 O 3 without seeds. (author)

Study on the Conversion of Fuel Nitrogen Into NOx

Directory of Open Access Journals (Sweden)

Raminta Plečkaitienė

2011-12-01

Full Text Available The aim of this work is to investigate NOx regularities combusting fuels having high concentration of nitrogen and to develop methods that will reduce the conversion of fuel nitrogen into NOx. There are three solutions to reducing NOx concentration: the combustion of fuel mixing it with other types of “clean” fuel containing small amounts of nitrogen, laundering fuel and the combustion of fuel using carbon additives. These solutions can help with reducing the amount of nitrogen in the wood waste of furniture by about 30% by washing fuel with water. Therefore, NOx value may decrease by about 35%.Article in Lithuanian

Estimates of relative areas for the disposal in bedded salt of LWR wastes from alternative fuel cycles

International Nuclear Information System (INIS)

Lincoln, R.C.; Larson, D.W.; Sisson, C.E.

1978-01-01

The relative mine-level areas (land use requirements) which would be required for the disposal of light-water reactor (LWR) radioactive wastes in a hypothetical bedded-salt formation have been estimated. Five waste types from alternative fuel cycles have been considered. The relative thermal response of each of five different site conditions to each waste type has been determined. The fuel cycles considered are the once-through (no recycle), the uranium-only recycle, and the uranium and plutonium recycle. The waste types which were considered include (1) unreprocessed spent reactor fuel, (2) solidified waste derived from reprocessing uranium oxide fuel, (3) plutonium recovered from reprocessing spent reactor fuel and doped with 1.5% of the accompanying waste from reprocessing uranium oxide fuel, (4) waste derived from reprocessing mixed uranium/plutonium oxide fuel in the third recycle, and (5) unreprocessed spent fuel after three recycles of mixed uranium/plutonium oxide fuels. The relative waste-disposal areas were determined from a calculated value of maximum thermal energy (MTE) content of the geologic formations. Results are presented for each geologic site condition in terms of area ratios. Disposal area requirements for each waste type are expressed as ratios relative to the smallest area requirement (for waste type No. 2 above). For the reference geologic site condition, the estimated mine-level disposal area ratios are 4.9 for waste type No. 1, 4.3 for No. 3, 2.6 for No. 4, and 11 for No. 5

Development of MHI PWR fuel assembly with high thermal performance

International Nuclear Information System (INIS)

Yasushi Makino; Masaya Hoshi; Masaji Mori; Hidetoshi Kido; Kazuo Ikeda

2005-01-01

Mitsubishi Heavy Industries, Ltd. (MHI) has been developing a PWR fuel assembly to meet the needs of Japanese fuel market with mainly improving its reliability such as a mechanical strength, a seismic strength and endurance. For burn-up extension of the fuel to 55 GWd/t, MHI has introduced a Zircaloy spacer grid with better neutron economics with retaining the reliability in an operating core. However, for a future power up-rating and a longer cycle operation, a higher thermal performance is required for PWR fuel assembly. To meet the needs of fuel market, MHI has developed an advanced type of Zircaloy spacer grid with a greater DNB performance while retaining the reliability of a fuel and a relatively low pressure drop. For the greater DNB performance, MHI optimized geometrical shape of mixing vane to promote a fluid mixing performance. In this report, higher DNB performance provided by the advanced Zircaloy spacer grid is presented. The results of 3D simulation for the flow behavior in 5 x 5 partial assembly, a mixing test and a water DNB test were compared between the current and the advanced spacer grids. Consequently, it was confirmed that a crossover vane enhanced a fluid mixing and the advanced spacer grid could significantly improve DNB performance compared with the current design of spacer grids. (authors)

Feasibility of Electromagnetic Acoustic Evaluation for Quality Test of a Plate-type Nuclear Fuel

International Nuclear Information System (INIS)

Jung, Hyun Kyu; Lee, Yoon Sang; Cheong, Yong Moo

2010-01-01

Most research and test reactors use the nuclear fuel plates which are consisted of a fuel core in aluminum alloy. Recently KAERI signed a deal with the Jordan Atomic Energy Commission to build the research reactor and have to supply the plate-type nuclear fuels. For the demands of world market, KAERI started the research and development of the plate-type fuel elements and endeavored to achieve a localization of fuel fabrication. For the inspection of plate-type fuel elements to be used in Research Reactors, an immersion pulse-echo ultrasonic technique was applied. This inspection was done with water, so a nuclear fuel was immersed to be prone to corrosion and needed to have time and cost due to an additional process. The sample that will be examined within this paper is a non-ferromagnetic material such as aluminum which has a good acousto-elastic property, for an effective inspection of a bond quality for a nuclear fuel under a manufacturing environment. The purpose of this study is to investigate the feasibility of an EMAT technology for an automated inspection of a nuclear fuel without water

Operation of Nuclear Fuel Based on Reprocessed Uranium for VVER-type Reactors in Competitive Nuclear Fuel Cycles

Energy Technology Data Exchange (ETDEWEB)

Troyanov, V.; Molchanov, V.; Tuzov, A. [TVEL Corporation, 49 Kashirskoe shosse, Moscow 115409 (Russian Federation); Semchenkov, Yu.; Lizorkin, M. [RRC ' Kurchatov Institute' (Russian Federation); Vasilchenko, I.; Lushin, V. [OKB ' Gidropress' (Russian Federation)

2009-06-15

Current nuclear fuel cycle of Russian nuclear power involves reprocessed low-enriched uranium in nuclear fuel production for some NPP units with VVER-type LWR. This paper discusses design and performance characteristics of commercial nuclear fuel based on natural and reprocessed uranium. It presents the review of results of commercial operation of nuclear fuel based on reprocessed uranium on Russian NPPs-unit No.2 of Kola NPP and unit No.2 of Kalinin NPP. The results of calculation and experimental validation of safe fuel operation including necessary isotope composition conformed to regulation requirements and results of pilot fuel operation are also considered. Meeting the customer requirements the possibility of high burn-up achieving was demonstrated. In addition the paper compares the characteristics of nuclear fuel cycles with maximum length based on reprocessed and natural uranium considering relevant 5% enrichment limitation and necessity of {sup 236}U compensation. The expedience of uranium-235 enrichment increasing over 5% is discussed with the aim to implement longer fuel cycles. (authors)

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.)

Ducted fuel injection

Energy Technology Data Exchange (ETDEWEB)

Mueller, Charles J.

2018-03-06

Various technologies presented herein relate to enhancing mixing inside a combustion chamber to form one or more locally premixed mixtures comprising fuel and charge-gas with low peak fuel to charge-gas ratios to enable minimal, or no, generation of soot and other undesired emissions during ignition and subsequent combustion of the locally premixed mixtures. To enable sufficient mixing of the fuel and charge-gas, a jet of fuel can be directed to pass through a bore of a duct causing charge-gas to be drawn into the bore creating turbulence to mix the fuel and the drawn charge-gas. The duct can be located proximate to an opening in a tip of a fuel injector. The duct can comprise of one or more holes along its length to enable charge-gas to be drawn into the bore, and further, the duct can cool the fuel and/or charge-gas prior to combustion.

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

International Nuclear Information System (INIS)

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

1993-01-01

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

Build-up and decay of fuel actinides in the fuel cycle of nuclear reactors

International Nuclear Information System (INIS)

Tasaka, Kanji; Kikuchi, Yasuyuki; Shindo, Ryuichi; Yoshida, Hiroyuki; Yasukawa, Shigeru

1976-05-01

For boiling water reactors, pressurized light-water reactors, pressure-tube-type heavy water reactors, high-temperature gas-cooled reactors, and sodium-cooled fast breeder reactors, uranium fueled and mixed-oxide fueled, each of 1000 MWe, the following have been studied: (1) quantities of plutonium and other fuel actinides built up in the reactor, (2) cooling behaviors of activities of plutonium and other fuel actinides in the spent fuels, and (3) activities of plutonium and other fuel actinides in the high-level reprocessing wastes as a function of storage time. The neutron cross section and decay data of respective actinide nuclides are presented, with their evaluations. For effective utilization of the uranium resources and easy reprocessing and high-level waste management, a thermal reactor must be fueled with uranium; the plutonium produced in a thermal reactor should be used in a fast reactor; and the plutonium produced in the blanket of a fast reactor is more appropriate for a fast reactor than that from a thermal reactor. (auth.)

Modern new nuclear fuel characteristics and radiation protection aspects.

Science.gov (United States)

Terry, Ian R

2005-01-01

The glut of fissile material from reprocessing plants and from the conclusion of the cold war has provided the opportunity to design new fuel types to beneficially dispose of such stocks by generating useful power. Thus, in addition to the normal reactor core complement of enriched uranium fuel assemblies, two other types are available on the world market. These are the ERU (enriched recycled uranium) and the MOX (mixed oxide) fuel assemblies. Framatome ANP produces ERU fuel assemblies by taking feed material from reprocessing facilities and blending this with highly enriched uranium from other sources. MOX fuel assemblies contain plutonium isotopes, thus exploiting the higher neutron yield of the plutonium fission process. This paper describes and evaluates the gamma, spontaneous and alpha reaction neutron source terms of these non-irradiated fuel assembly types by defining their nuclear characteristics. The dose rates which arise from these terms are provided along with an overview of radiation protection aspects for consideration in transporting and delivering such fuel assemblies to power generating utilities.

Evaluation of plate type fuel elements by eddy current test method

International Nuclear Information System (INIS)

Frade, Rangel Teixeira

2015-01-01

Plate type fuel elements are used in MTR research nuclear reactors. The fuel plates are manufactured by assembling a briquette containing the fissile material inserted in a frame, with metal plates in both sides of the set, to act as a cladding. This set is rolled under controlled conditions in order to obtain the fuel plate. In Brazil, this type of fuel is manufactured by IPEN and used in the IEA-R1 reactor. After fabrication of three batches of fuel plates, 24 plates, one of them is taken, in order to verify the thickness of the cladding. For this purpose, the plate is sectioned and the thickness measurements are carried out by using optical microscopy. This procedure implies in damage of the plate, with the consequent cost. Besides, the process of sample preparation for optical microscopy analysis is time consuming, it is necessary an infrastructure for handling radioactive materials and there is a generation of radioactive residues during the process. The objective of this study was verify the applicability of eddy current test method for nondestructive measurement of cladding thickness in plate type nuclear fuels, enabling the inspection of all manufactured fuel plates. For this purpose, reference standards, representative of the cladding of the fuel plates, were manufactured using thermomechanical processing conditions similar to those used for plates manufacturing. Due to no availability of fuel plates for performing the experiments, the presence of the plate’s core was simulated using materials with different electrical conductivities, fixed to the thickness reference standards. Probes of eddy current testing were designed and manufactured. They showed high sensitivity to thickness variations, being able to separate small thickness changes. The sensitivity was higher in tests performed on the reference standards and samples without the presence of the materials simulating the core. For examination of the cladding with influence of materials simulating the

Modeling of thermo-mechanical and irradiation behavior of mixed oxide fuel for sodium fast reactors

International Nuclear Information System (INIS)

Karahan, Aydin; Buongiorno, Jacopo

2010-01-01

An engineering code to model the irradiation behavior of UO 2 -PuO 2 mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named fuel engineering and structural analysis tool (FEAST-OXIDE). FEAST-OXIDE has several modules working in coupled form with an explicit numerical algorithm. These modules describe: (1) fission gas release and swelling, (2) fuel chemistry and restructuring, (3) temperature distribution, (4) fuel-clad chemical interaction and (5) fuel-clad mechanical analysis. Given the fuel pin geometry, composition and irradiation history, FEAST-OXIDE can analyze fuel and cladding thermo-mechanical behavior at both steady-state and design-basis transient scenarios. The code was written in FORTRAN-90 program language. The mechanical analysis module implements the LIFE algorithm. Fission gas release and swelling behavior is described by the OGRES and NEFIG models. However, the original OGRES model has been extended to include the effects of joint oxide gain (JOG) formation on fission gas release and swelling. A detailed fuel chemistry model has been included to describe the cesium radial migration and JOG formation, oxygen and plutonium radial distribution and the axial migration of cesium. The fuel restructuring model includes the effects of as-fabricated porosity migration, irradiation-induced fuel densification, grain growth, hot pressing and fuel cracking and relocation. Finally, a kinetics model is included to predict the clad wastage formation. FEAST-OXIDE predictions have been compared to the available FFTF, EBR-II and JOYO databases, as well as the LIFE-4 code predictions. The agreement was found to be satisfactory for steady-state and slow-ramp over-power accidents.

Modeling of thermo-mechanical and irradiation behavior of mixed oxide fuel for sodium fast reactors

Energy Technology Data Exchange (ETDEWEB)

Karahan, Aydin, E-mail: karahan@mit.ed [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering Department, Massachusetts Institute of Technology, MA (United States); Buongiorno, Jacopo [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering Department, Massachusetts Institute of Technology, MA (United States)

2010-01-31

An engineering code to model the irradiation behavior of UO{sub 2}-PuO{sub 2} mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named fuel engineering and structural analysis tool (FEAST-OXIDE). FEAST-OXIDE has several modules working in coupled form with an explicit numerical algorithm. These modules describe: (1) fission gas release and swelling, (2) fuel chemistry and restructuring, (3) temperature distribution, (4) fuel-clad chemical interaction and (5) fuel-clad mechanical analysis. Given the fuel pin geometry, composition and irradiation history, FEAST-OXIDE can analyze fuel and cladding thermo-mechanical behavior at both steady-state and design-basis transient scenarios. The code was written in FORTRAN-90 program language. The mechanical analysis module implements the LIFE algorithm. Fission gas release and swelling behavior is described by the OGRES and NEFIG models. However, the original OGRES model has been extended to include the effects of joint oxide gain (JOG) formation on fission gas release and swelling. A detailed fuel chemistry model has been included to describe the cesium radial migration and JOG formation, oxygen and plutonium radial distribution and the axial migration of cesium. The fuel restructuring model includes the effects of as-fabricated porosity migration, irradiation-induced fuel densification, grain growth, hot pressing and fuel cracking and relocation. Finally, a kinetics model is included to predict the clad wastage formation. FEAST-OXIDE predictions have been compared to the available FFTF, EBR-II and JOYO databases, as well as the LIFE-4 code predictions. The agreement was found to be satisfactory for steady-state and slow-ramp over-power accidents.

Photo guide for estimating fuel loading and fire behavior in mixed-oak forests of the Mid-Atlantic Region

Science.gov (United States)

Patrick H. Brose

2009-01-01

A field guide of 45 pairs of photographs depicting ericaceous shrub, leaf litter, and logging slash fuel types of eastern oak forests and observed fire behavior of these fuel types during prescribed burning. The guide contains instructions on how to use the photo guide to choose appropriate fuel models for prescribed fire planning.

A dynamic, dependent type system for nuclear fuel cycle code generation

Energy Technology Data Exchange (ETDEWEB)

Scopatz, A. [The University of Chicago 5754 S. Ellis Ave, Chicago, IL 60637 (United States)

2013-07-01

The nuclear fuel cycle may be interpreted as a network or graph, thus allowing methods from formal graph theory to be used. Nodes are often idealized as nuclear fuel cycle facilities (reactors, enrichment cascades, deep geologic repositories). With the advent of modern object-oriented programming languages - and fuel cycle simulators implemented in these languages - it is natural to define a class hierarchy of facility types. Bright is a quasi-static simulator, meaning that the number of material passes through a facility is tracked rather than natural time. Bright is implemented as a C++ library that models many canonical components such as reactors, storage facilities, and more. Cyclus is a discrete time simulator, meaning that natural time is tracked through out the simulation. Therefore a robust, dependent type system was developed to enable inter-operability between Bright and Cyclus. This system is capable of representing any fuel cycle facility. Types declared in this system can then be used to automatically generate code which binds a facility implementation to a simulator front end. Facility model wrappers may be used either internally to a fuel cycle simulator or as a mechanism for inter-operating multiple simulators. While such a tool has many potential use cases it has two main purposes: enabling easy performance of code-to-code comparisons and the verification and the validation of user input.

Pathfinder irradiation of advanced fuel (Th/U mixed oxide) in a power reactor

International Nuclear Information System (INIS)

Brant Pinheiro, R.

1993-01-01

Within the joint Brazilian-German cooperative R and D Program on Thorium Utilization in Pressurized Water Reactors carried out from 1979 to 1988 by Nuclebras/CDTN, KFA-Juelich, Siemens/KWU and NUKEM, a pathfinder irradiation of Th/U mixed oxide fuel in the Angra 1 nuclear power reactor was planned. The objectives of this irradiation testing, the irradiation strategy, the work performed and the status achieved at the end of the joint Program are presented. (author)

Measurement of Mixing Rate between Fuel Subchannels: Development of a new Experimental Technique

International Nuclear Information System (INIS)

Silin, Nicolas; Barbero, Jose; Bubach, Ernesto; Juanico, Luis

2000-01-01

A superficial heater of nickel applied over a ceramic substrate was designed and constructed, together with a system of high sensitivity to measure temperature differentials. The use of both techniques was evaluated and it might allow for the wider use of the method of differential thermal analysis to quantify the turbulent mixing between coupled hydraulic subchannels in fuel elements. Even more, the method presents important advantages as compared to the more complicated techniques known (laser Doppler anemometry)

Effect of the UO{sub 2} powder type and mixing method on microstructure of Mn-Al doped pellet

Energy Technology Data Exchange (ETDEWEB)

Na, Yeon Soo; Lim, Kwang Young; Choi, Min young; Jung, Tae Sik; Lee, Seung Jae; Yoo, Jong Sung [KEPCO, Daejeon (Korea, Republic of)

2016-05-15

Recently, the commercial LWRs are focused on the extending the burn-up and fuel cycle length in order to increase nuclear power plant economy as a maintenance and fuel cycle cost. Increasing the burn-up may lead to a faster and higher power variation such as a peak local linear power and normal operating transient (Load following operation). In such operating conditions, the risk of a fuel failure is considerably related to a pellet clad-interaction (PCI). So, recent development of advanced UO{sub 2} pellet for the LWRs is mainly focused on the large grain and soft pellet as they can reduce corrosive fission gas release and pellet-clad-interaction. In terms of the UO{sub 2} pellet, the prevention of PCI induced fuel failure can be achieved by enlarging the UO{sub 2} pellet grain size and enhancing the pellets deformation at an elevated temperature. In Korea, in order to increase the grain size and deformation of UO{sub 2} pellet on the high temperature, Mn-Al doped pellet with ADU (Ammonium Diuranate)-UO{sub 2} powder are developed in lab scale. But, the UO{sub 2} pellets for the commercial nuclear power plants in Korea are fabricated using the DC (Dry Conversion)-UO{sub 2} powder. So, it is necessary to understand the effect of microstructure on UO{sub 2} powder type for Mn-Al doped pellets. In this work, to investigate the effect of UO{sub 2} powder type and mixing method on the microstructure of the Mn-Al doped UO{sub 2} pellets, we fabricated the Mn-Al doped pellets using the DC-UO{sub 2} powder. The measurement of sintered density and mean grain size for fabricated pellets was performed, and then the results of test was evaluated in comparison with a Reference 2.

Fast breeder fuel element development

International Nuclear Information System (INIS)

Marth, W.; Muehling, G.

1983-08-01

This report is a compilation of the papers which have been presented during a seminar ''Fast Breeder Fuel Element Development'' held on November 15/16, 1982 at KfK. The papers give a survey of the status, of the obtained results and of the necessary work, which still has to be done in the frame of various development programmes for fast breeder fuel elements. In detail the following items were covered by the presentations: - the requirements and boundary conditions for the design of fuel pins and elements both for the reference concept of the SNR 300 core and for the large, commercial breeder type of the future (presentation 1,2 and 6); - the fabrication, properties and characterization of various mixed oxide fuel types (presentations 3,4 and 5); - the operational fuel pin behaviour, limits of different design concepts and possible mechanism for fuel pin failures (presentations (7 and 8); - the situation of cladding- and wrapper materials development especially with respect to the high burn-up values of commercial reactors (presentations 9 and 10); - the results of the irradiation experiments performed under steady-state and non-stationary operational conditions and with failed fuel pins (presentations 11, 12, 13 and 14). (orig./RW) [de

Evidos: optimisation of individual monitoring in mixed neutron/photon fields at workplaces of the nuclear fuel cycle

International Nuclear Information System (INIS)

Luszik-Bhadra, M.; Reginatto, M.; Schuhmacher, H.; Lacoste, V.; Muller, M.; Boschung, M.; Fiechtner, A.; Coeck, M.; Vanhavere, F.; Curzio, G.; D'errico, F.; Kyllonen, J.E.; Lindborg, L.; Molinos, C.; Tanner, R.; Derdau, D.; Lahaye, Th.

2005-01-01

Within its 5. Framework Programme, the EC is funding the project EVIDOS ('Evaluation of Individual Dosimetry in Mixed Neutron and Photon Radiation Fields'). The aim of this project is the optimisation of individual monitoring at workplaces of the nuclear fuel cycle with special regard to neutrons. Various dosemeters for mixed field application - passive and new electronic devices - are tested in selected workplace fields in nuclear installations in Europe. The fields are characterised using a series of spectrometers that provide the energy distribution of neutron fluence (Bonner spheres) and newly developed devices that provide the energy and directional distribution of the neutron fluence. Results from the first measurement campaign, carried out in simulated workplace fields (IRSN, Cadarache. France), and those of a second measurement campaign, carried out at workplaces at a boiling water reactor and at a storage cask with used fuel elements (Kernkraftwerk Kriimmel, Germany), are described. (authors)

Study of the neutronic performances of cores with mixed nitride fuel [(U,Pu)N] for fast neutron reactors

International Nuclear Information System (INIS)

Merzouk, Hamid

1992-01-01

This paper proposes a core design of fast reactor using mixed nitride fuel [(U,Pu)N], having small loss of reactivity and reaching a maximum thermal burn-up rate from 150 GWd/t, while being managed in single batch (renewal of the fuel in only one time for all the subassemblies of the core). This work was completed with aid of the studies of sensibilities of the fast reactors cores to principal parameters: general design of the core, volumetric percentages of the various mixture of materials composing the core, initial enrichments of the fuel. A detailed optimization study on the selected core was conducted complying with safety criteria taking into consideration of consequences of nitride material presence on fuel assembly design rules. (author) [fr

Cup-mixing type cup vending machines; Cup mixing shiki cup jido hanbaiki

Energy Technology Data Exchange (ETDEWEB)

Nishiwaki, S.; Nagasaki, T.; Hori, S. [Fuji Electric Co. Ltd., Tokyo (Japan)

1996-07-10

This paper introduces the newly developed cup-mixing type cup vending machines. In the vending machines, the powder material for each beverage is stored, and the powder material delivered to a cup is carried to the cooking block by the carrier mechanism and cooked and mixed using a propeller mixer after hot and cold water are put. These vending machines meet the needs to the diversified and quality taste. The ejection port of products was also improved, and the door was designed to be higher in grade. The main block of structure is described next. The cup carrier mechanism consists of X, Y, and Z axes, and a hand. The cup is positioned by the control of a stepping motor. The number of propeller rotations in the propeller mechanism can be set to nine levels. The mixing position can also be set freely. In the temporary material reservation mechanism, the stored material is temporarily reserved until a cup is carried. The ejection port door of a cup is opened or closed automatically. The control block that drives the cup carrier mechanism is connected with the host unit by serial communication. 9 figs., 1 tab.

Perspectives for practical application of the combined fuel kernels in VVER-type reactors

International Nuclear Information System (INIS)

Baranov, V.; Ternovykh, M.; Tikhomirov, G.; Khlunov, A.; Tenishev, A.; Kurina, I.

2011-01-01

The paper considers the main physical processes that take place in fuel kernels under real operation conditions of VVER-type reactors. Main attention is given to the effects induced by combinations of layers with different physical properties inside of fuel kernels on these physical processes. Basic neutron-physical characteristics were calculated for some combined fuel kernels in fuel rods of VVER-type reactors. There are many goals in development of the combined fuel kernels, and these goals define selecting the combinations and compositions of radial layers inside of the kernels. For example, the slower formation of the rim-layer on outer surface of the kernels made of enriched uranium dioxide can be achieved by introduction of inner layer made of natural or depleted uranium dioxide. Other potential goals (lower temperature in the kernel center, better conditions for burn-up of neutron poisons, better retention of toxic materials) could be reached by other combinations of fuel compositions in central and peripheral zones of the fuel kernels. Also, the paper presents the results obtained in experimental manufacturing of the combined fuel pellets. (authors)

Guidebook on quality control of mixed oxides and gadolinium bearing fuels for light water reactors

International Nuclear Information System (INIS)

1991-02-01

Under the coverage of an efficient quality assurance system, quality control in nuclear fuel fabrication is an essential element to assure the reliable performance of all its components in service. Incentives to increase fuel performance, by extending reactor cycles or achieving higher burnups and, in some countries to use recycled plutonium in light water reactors (LWRs) necessitated the development of new types of fuels. In the first case, due to higher uranium enrichments, a burnable neutron absorber was integrated to the fuel pellets. Gadolinia was found to form a solid solution with Uranium dioxide and, to present a burnup rate which matches fissile uranium depletion. (U,Gd)O 2 fuels which have been successfully used since the seventies, in boiling water reactors have more recently found an increased utilization, in pressurized water reactors. This amply justifies the publication of this TECDOC to encourage authorities, designers and manufacturers of these types of fuel to establish a more uniform, adapted and effective system of control, thus promoting improved materials reliability and good performance in advanced fuel for light water reactors. The Guidebook is subdivided into four chapters written by different authors. A separate abstract was prepared for each of these chapters. Refs, figs and tabs

Developments in MOX fuel pellet fabrication technology: Indian experience

International Nuclear Information System (INIS)

Kamath, H.S.; Majumdar, S.; Purusthotham, D.S.C.

1998-01-01

India is interested in mixed oxide (MOX) fuel technology for better utilisation of its nuclear fuel resources. In view of this, a programme involving MOX fuel design, fabrication and irradiation in research and power reactors has been taken up. A number of experimental irradiations in research reactors have been carried out and a few MOX assemblies of ''All Pu'' type have been loaded in our commercial BWRs at Tarapur. An island type of MOX fuel design is under study for use in PHWRs which can increase the burn-up of the fuel by more than 30% compared to natural UO 2 fuel. The MOX fuel pellet fabrication technology for the above purpose and R and D efforts in progress for achieving better fuel performance are described in the paper. The standard MOX fuel fabrication route involves mechanical mixing and milling of UO 2 and PuO 2 powders. After detailed investigations with several types of mixing and milling equipments, dry attritor milling has been found to be the most suitable for this operation. Neutron Coincident Counting (NCC) technique was found to be the most convenient and appropriate technique for quick analysis of Pu content in milled MOX powder and to know Pu mixing is homogenous or not. Both mechanical and hydraulic presses have been used for powder compaction for green pellet production although the latter has been preferred for better reproducibility. Low residue admixed lubricants have been used to facilitate easy compaction. The normal sintering temperature used in Nitrogen-Hydrogen atmosphere is between 1600 deg. C to 1700 deg. C. Low temperature sintering (LTS) using oxidative atmospheres such as carbon dioxide, Nitrogen and coarse vacuum have also been investigated on UO 2 and MOX on experimental scale and irradiation behaviour of such MOX pellets is under study. Ceramic fibre lined batch furnaces have been found to be the most suitable for MOX pellet production as they offer very good flexibility in sintering cycle, and ease of maintainability

Safety and Regulatory Issues of the Thorium Fuel Cycle

Energy Technology Data Exchange (ETDEWEB)

Ade, Brian [ORNL; Worrall, Andrew [ORNL; Powers, Jeffrey [ORNL; Bowman, Steve [ORNL; Flanagan, George [ORNL; Gehin, Jess [ORNL

2014-02-01

Thorium has been widely considered an alternative to uranium fuel because of its relatively large natural abundance and its ability to breed fissile fuel (233U) from natural thorium (232Th). Possible scenarios for using thorium in the nuclear fuel cycle include use in different nuclear reactor types (light water, high temperature gas cooled, fast spectrum sodium, molten salt, etc.), advanced accelerator-driven systems, or even fission-fusion hybrid systems. The most likely near-term application of thorium in the United States is in currently operating light water reactors (LWRs). This use is primarily based on concepts that mix thorium with uranium (UO2 + ThO2), add fertile thorium (ThO2) fuel pins to LWR fuel assemblies, or use mixed plutonium and thorium (PuO2 + ThO2) fuel assemblies. The addition of thorium to currently operating LWRs would result in a number of different phenomenological impacts on the nuclear fuel. Thorium and its irradiation products have nuclear characteristics that are different from those of uranium. In addition, ThO2, alone or mixed with UO2 fuel, leads to different chemical and physical properties of the fuel. These aspects are key to reactor safety-related issues. The primary objectives of this report are to summarize historical, current, and proposed uses of thorium in nuclear reactors; provide some important properties of thorium fuel; perform qualitative and quantitative evaluations of both in-reactor and out-of-reactor safety issues and requirements specific to a thorium-based fuel cycle for current LWR reactor designs; and identify key knowledge gaps and technical issues that need to be addressed for the licensing of thorium LWR fuel in the United States.

Fuel management of mixed reactor type power plant systems

International Nuclear Information System (INIS)

Csom, Gyula

1988-01-01

Breeding gain in symbiotic nuclear power plant system consisting of both thermal and fast breeder reactors depends on the characteristics and the ratio of thermal and fast reactors. The composition of the symbiotic power plant systems was determined for equilibrium and plutonium deficient systems. According to natural uranium utilization, symbiotic power plant systems are not less efficient than the systems containing only fast breeders. Depleted uranium can be applied in both types of systems. Reprocessing demands of the symbiotic power plant sytems were determined. (V.N.) 23 figs.; 1 tab

Methods for analysis of uranium-plutonium mixed fuel and transplutonium elements at RIAR (Preprint no. IT-25)

International Nuclear Information System (INIS)

Timofeev, G.A.

1991-02-01

Different methods for analysis of the uranium-plutonium mixed nuclear fuel and transplutonium elements are briefly discussed in this paper: coulometry, radiometric techniques, emission spectrography, mass-spectrometry, chromatography, spectrophotometry. The main analytical characteristics of the methods developed are given. (author). 30 refs., 2 tabs

Determining how much mixed waste will require disposal

International Nuclear Information System (INIS)

Kirner, N.P.

1990-01-01

Estimating needed mixed-waste disposal capacity to 1995 and beyond is an essential element in the safe management of low-level radioactive waste disposal capacity. Information on the types and quantities of mixed waste generated is needed by industry to allow development of treatment facilities and by states and others responsible for disposal and storage of this type of low-level radioactive waste. The design of a mixed waste disposal facility hinges on a detailed assessment of the types and quantities of mixed waste that will ultimately require land disposal. Although traditional liquid scintillation counting fluids using toluene and xylene are clearly recognized as mixed waste, characterization of other types of mixed waste has, however, been difficult. Liquid scintillation counting fluids comprise most of the mixed waste generated and this type of mixed waste is generally incinerated under the supplemental fuel provisions of the Resource Conservation and Recovery Act (RCRA) Because there are no Currently operating mixed waste land disposal facilities, it is impossible to make projections of waste requiring land disposal based on a continuation of current waste disposal practices. Evidence indicates the volume of mixed waste requiring land disposal is not large, since generators are apparently storing these wastes. Surveys conducted to date confirm that relatively small volumes of commercially generated mixed waste volume have relied heavily oil generators' knowledge of their wastes. Evidence exists that many generators are confused by the differences between the Atomic Energy Act and the Resource Conservation and Recovery Act (RCRA) on the issue of when a material becomes a waste. In spite of uncertainties, estimates of waste volumes requiring disposal can be made. This paper proposes an eight-step process for such estimates

Investigation of strut-ramp injector in a Scramjet combustor: Effect of strut geometry, fuel and jet diameter on mixing characteristics

Energy Technology Data Exchange (ETDEWEB)

Soni, Rahul Kumar; De, Ashoke De [Indian Institute of Technology Kanpur, Kanpur (India)

2017-03-15

The strut-based injector has been found to be one of the most promising injector designs for a supersonic combustor, offering enhanced mixing of fuel and air. The mixing and flow field characteristics of the straight (SS) and Tapered strut (TS), with fixed ramp angle and height at free stream Mach number 2 in conjunction with fuel injection at Mach 2.3 have been investigated numerically and reported. In the present investigation, hydrogen (H{sub 2}) and ethylene (C{sub 2}H{sub 4}) are injected in oncoming supersonic flow from the back of the strut, where jet to free stream momentum ratio is maintained at 0.79 and 0.69 for H2 and C{sub 2}H{sub 4}, respectively. The predicted wall static pressure and species mole fractions at various downstream locations are compared with the experimental data for TS case with 0.6 mm jet diameter and found to be in good agreement. Further, the effect of jet diameter and strut geometry on the near field mixing in strut ramp configuration is discussed for both the fuels. The numerical results are assessed based on various parameters for the performance evaluation of different strut ramp configurations. The SS configuration for both the injectant has been found to be an optimum candidate; also it is observed that for higher jet diameter larger combustor length is required to achieve satisfactory near field mixing.

Field experience of new nuclear fuel types on the Kola NPP

International Nuclear Information System (INIS)

Adeev, V.; Burlov, S.; Panov, A.; Saprykin, V.

2008-01-01

Specificity of the Kola nuclear power plant geographical position, conditions of region economics determine fuel management strategy. Isolation of Kola power supply system and, as a consequence, generating capacities redundancy cause operation of the nuclear power plant on reduced power level. At the same time there is a need to operate the power unit on the maximum power level in the case of not planned conditions. The basis of in-core fuel management is an achievement of the maximal burnup under providing of high installed capacity. At present there are not abilities to improve the fuel cycle based on traditional implementation fuel assemblies. Burnup maximum in these fuel cycles is achieved. At the core periphery installed highest possible quantity of the burned-up assemblies in the view of safety operation margins satisfaction. Works on application of the second generation fuel have been carried out on the Kola NPP since 2002. Fuel assemblies of this type are profiled. Burnable absorber, changed lattice spacing in relation to standard fuel, changed height of a fuel column, thickness of fuel pin clad are applied. In CR fuel followers modernized docking unit (with hafnium plates are intended for energy-release splash suppression) is used. At present 2-nd generation fuel is in experimental operation on unit 3 (18-21 fuel cycles, 2002-2007 years) and unit 4 (18-19 fuel cycles, 2005-2007 years). Safety margins did not exceeded. Coolant activity did not exceed the limiting value. There were not damaged fuel assemblies of second generation. Originally in the project of applications of new fuel it was supposed to refuel annually 78 fresh assemblies. At the moment annual refueling consists of 66 assemblies with effective enrichment 3.82 %. Cycle duration does not exceed 250-260 effective days. The part of assemblies is left on 5-th cycle of operation. In a similar fuel cycle in 2007 on the unit 1 operation with profiled fuel (enrichment of 3.82 %) of shakeproof type

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.

Study on new-type fuel-related assembly handling tools for PWR NPP

International Nuclear Information System (INIS)

Fan Xiumei

2013-01-01

This article describes the design and study on a set of new-type fuel-related assembly snatching tools used for PWR NPP. The purpose is mainly to enhance the tool safety, reliability and convenientness by improvement of the mechanism and structure of the tool for snatching preciseness and avoiding from falling and abrasion of fuel-related assemblies for any condition. The new-type fuel-related assembly handling tools are compared with similar equipment in worldwide in terms of function, main technical characteristic, and safety and protection, some of them are better than the similar equipment in that they have reliable loading and unloading and conveying capabilities. (author)

International safeguards for a modern MOX [mixed-oxide] fuel fabrication facility

International Nuclear Information System (INIS)

Pillay, K.K.S.; Stirpe, D.; Picard, R.R.

1987-03-01

Bulk-handling facilities that process plutonium for commercial fuel cycles offer considerable challenges to nuclear materials safeguards. Modern fuel fabrication facilities that handle mixed oxides of plutonium and uranium (MOX) often have large inventories of special nuclear materials in their process lines and in storage areas for feed and product materials. In addition, the remote automated processing prevalent at new MOX facilities, which is necessary to minimize radiation exposures to personnel, tends to limit access for measurements and inspections. The facility design considered in this study incorporates all these features as well as state-of-the-art measurement technologies for materials accounting. Key elements of International Atomic Energy Agency (IAEA) safeguards for such a fuel-cycle facility have been identified in this report, and several issues of primary importance to materials accountancy and IAEA verifications have been examined. We have calculated detection sensitivities for abrupt and protracted diversions of plutonium assuming a single materials balance area for all processing areas. To help achieve optimal use of limited IAEA inspection resources, we have calculated sampling plans for attributes/variables verification. In addition, we have demonstrated the usefulness of calculating σ/sub (MUF-D)/ and detection probabilities corresponding to specified material-loss scenarios and resource allocations. The data developed and the analyses performed during this study can assist both the facility operator and the IAEA in formulating necessary safeguards approaches and verification procedures to implement international safeguards for special nuclear materials

International safeguards for a modern MOX (mixed-oxide) fuel fabrication facility

Energy Technology Data Exchange (ETDEWEB)

Pillay, K.K.S.; Stirpe, D.; Picard, R.R.

1987-03-01

Bulk-handling facilities that process plutonium for commercial fuel cycles offer considerable challenges to nuclear materials safeguards. Modern fuel fabrication facilities that handle mixed oxides of plutonium and uranium (MOX) often have large inventories of special nuclear materials in their process lines and in storage areas for feed and product materials. In addition, the remote automated processing prevalent at new MOX facilities, which is necessary to minimize radiation exposures to personnel, tends to limit access for measurements and inspections. The facility design considered in this study incorporates all these features as well as state-of-the-art measurement technologies for materials accounting. Key elements of International Atomic Energy Agency (IAEA) safeguards for such a fuel-cycle facility have been identified in this report, and several issues of primary importance to materials accountancy and IAEA verifications have been examined. We have calculated detection sensitivities for abrupt and protracted diversions of plutonium assuming a single materials balance area for all processing areas. To help achieve optimal use of limited IAEA inspection resources, we have calculated sampling plans for attributes/variables verification. In addition, we have demonstrated the usefulness of calculating sigma/sub (MUF-D)/ and detection probabilities corresponding to specified material-loss scenarios and resource allocations. The data developed and the analyses performed during this study can assist both the facility operator and the IAEA in formulating necessary safeguards approaches and verification procedures to implement international safeguards for special nuclear materials.

Study of the Effect of (U0.8Pu0.2O2 Uranium–Plutonium Mixed Fuel Fission Products on a Living Organism

Directory of Open Access Journals (Sweden)

Ayagoz Baimukhanova

2016-08-01

Full Text Available The article describes the results of experiments conducted on pigs to determine the effect of plutonium, which is the most radiotoxic and highly active element in the range of mixed fuel (U0.8Pu0.2O2 fission products, on living organisms. The results will allow empirical prediction of the emergency plutonium radiation dose for various organs and tissues of humans in case of an accident in a reactor running on mixed fuel (U0.8Pu0.2O2.

Performance of fast reactor mixed-oxide fuels pins during extended overpower transients

International Nuclear Information System (INIS)

Tsai, H.; Neimark, L.A.; Asaga, T.; Shikakura, S.

1991-02-01

The Operational Reliability Testing (ORT) program, a collaborative effort between the US Department of Energy and the Power Reactor and Nuclear Fuel Development Corp. (PNC) of Japan, was initiated in 1982 to investigate the behavior of mixed-oxide fuel pin under various slow-ramp transient and duty-cycle conditions. In the first phase of the program, a series of four extended overpower transient tests, with severity sufficient to challenge the pin cladding integrity, was conducted. The objectives of the designated TOPI-1A through -1D tests were to establish the cladding breaching threshold and mechanisms, and investigate the thermal and mechanical effects of the transient on pin behavior. The tests were conducted in EBR-2, a normally steady-state reactor. The modes of transient operation in EBR-2 were described in a previous paper. Two ramp rates, 0.1%/s and 10%/s, were selected to provide a comparison of ramp-rate effects on fuel behavior. The test pins chosen for the series covered a range of design and pre-test irradiation parameters. In the first test (1A), all pins maintained their cladding integrity during the 0.1%/s ramp to 60% peak overpower. Fuel pins with aggressive designs, i.e., high fuel- smear density and/or thin cladding, were, therefore, included in the follow-up 1B and 1C tests to enhance the likelihood of achieving cladding breaching. In the meantime, a higher pin overpower capability, to greater than 100%, was established by increasing the reactor power limit from 62.5 to 75 MWt. In this paper, the significant results of the 1B and 1C tests are presented. 4 refs., 5 figs., 1 tab

Study of behavior of cermet fuel elements on IGR reactor under RIA type accident condition

International Nuclear Information System (INIS)

Vasil'ev, Yu.S.; Vurim, A.D.; Koltyshev, S.M.; Pakhnits, V.A.; Tukhvatulin, Sh.T.; Popov, V.V.; Ryzhkov, A.N.

1996-01-01

In 1993 December in IGR reactor of Inst. of Atomic Energy of National Nuclear Center of Republic of Kazakstan the second batch of in-pile testing of perspective cermet fuel elements under the condition, simulating RIA type accident was conducted. In the second batch of testing during eight start-ups 10 cermet fuel elements were examined. Among which 8 of monolith type and 2 fuel elements with false jacket beside cladding (FJF), as well as, 6 standard fuel elements of WWER-1000 type reactor with dioxide fuel were tested. 2 fuel elements - cermet and standard were placed into capsule filled with water. To measure energy release for the each start-up two fission monitor and inside core control gauge were placed. In all the start-ups operation mode of IGR was neutron pulse. Power of fuel element kept changing from 151 to 336 k W; energy release was 38-93 kJ/gr m 235 U; maximum temperature of cermet fuel was 1943-2173 K, of dioxide fuel - 1923-2843 K. The testing has demonstrated that operability of cermet fuel elements under reactivity accident condition with pulse width of 0,2 s is, at least, not less that operability of dioxide fuel elements, through advantages of cermet fuel under these conditions are revealed to the least extent

Core conversion from rod to plate type fuel elements in research reactors

International Nuclear Information System (INIS)

Khattab, M.S.; Mina, A.R.

1997-01-01

Core thermalhydraulic analysis have been performed for rod and plate types fuel elements without altering the core bundles square grid spacer (68 mm, side) and coolant mass flow rate. The U O 2 -Mg, 10% enrichment rod type fuel elements are replaced by the MTR plate type, U-Al alloy of 20% enrichment. Coolant mass flux increased from 2000 kg/m 2 S to 5000 kg/m 2 S. Reactor power could be upgraded from 2 to 10 MW without significantly altering the steady state, thermal-hydraulic safety margins. Fuel, clad and coolant transient temperatures are determined inside the core hot channel during flow coast down using paret code. Residual heat removal system of 20% coolant capacity is necessary for upgrading reactor power to encounter the case of pumps off at 10 MW nominal operation. 6 figs., 2 tabs

Air/fuel ratio visualization in a diesel spray

Science.gov (United States)

Carabell, Kevin David

1993-01-01

To investigate some features of high pressure diesel spray ignition, we have applied a newly developed planar imaging system to a spray in an engine-fed combustion bomb. The bomb is designed to give flow characteristics similar to those in a direct injection diesel engine yet provide nearly unlimited optical access. A high pressure electronic unit injector system with on-line manually adjustable main and pilot injection features was used. The primary scalar of interest was the local air/fuel ratio, particularly near the spray plumes. To make this measurement quantitative, we have developed a calibration LIF technique. The development of this technique is the key contribution of this dissertation. The air/fuel ratio measurement was made using biacetyl as a seed in the air inlet to the engine. When probed by a tripled Nd:YAG laser the biacetyl fluoresces, with a signal proportional to the local biacetyl concentration. This feature of biacetyl enables the fluorescent signal to be used as as indicator of local fuel vapor concentration. The biacetyl partial pressure was carefully controlled, enabling estimates of the local concentration of air and the approximate local stoichiometry in the fuel spray. The results indicate that the image quality generated with this method is sufficient for generating air/fuel ratio contours. The processes during the ignition delay have a marked effect on ignition and the subsequent burn. These processes, vaporization and pre-flame kinetics, very much depend on the mixing of the air and fuel. This study has shown that poor mixing and over-mixing of the air and fuel will directly affect the type of ignition. An optimal mixing arrangement exists and depends on the swirl ratio in the engine, the number of holes in the fuel injector and the distribution of fuel into a pilot and main injection. If a short delay and a diffusion burn is desired, the best mixing parameters among those surveyed would be a high swirl ratio, a 4-hole nozzle and a

Operation and Licensing of Mixed Cores in Water Cooled Reactors

International Nuclear Information System (INIS)

2013-11-01

Nuclear fuel is a highly complex material that is subject to continuous development and is produced by a range of manufacturers. During operation of a nuclear power plant, the nuclear fuel is subject to extreme conditions of temperature, corroding environment and irradiation, and many different designs of fuel have been manufactured with differing fuel materials, cladding materials and assembly structure to ensure these conditions. The core of an operating power plant can contain hundreds of fuel assemblies, and where there is more than a single design of a fuel assembly in the core, whether through a change of fuel vendor, introduction of an improved design or for some other reason, the core is described as a mixed core. The task of ensuring that the different assembly types do not interact in a harmful manner, causing, for example, differing flow resistance resulting in under cooling, is an important part of ensuring nuclear safety. This report has compiled the latest information on the operational experience of mixed cores and the tools and techniques that are used to analyse the core operation and demonstrate that there are no safety related problems with its operation. This publication is a result of a technical meeting in 2011 and a series of consultants meetings

Manufacturing method for fuel assembly

International Nuclear Information System (INIS)

Yamaguchi, Takashi.

1997-01-01

In an FBR type reactor, uranium/plutonium mixed oxide fuels (MOX fuels) are used. Nuclear fuel materials containing uranium and plutonium are filled to a portion or all of a plurality of fuel rods. In this case, an equivalent fissile coefficient (B) based on a combustion guarantee method defined by the formula: (B) = (M) · (F) is determined. (M) is a combustion matrix constituted based on the solution of equation of combustion which is a differential equation representing change with time of each of nuclear fuel materials during combustion. (F) is an equivalent fissile coefficient based on a reactivity keeping method which is a coefficient representing a reactivity worth equivalent with plutonium-239. The content of each of the nuclear fuel materials is determined so that the effective multiplication factor at the final stage of the operation cycle is substantially constant by using the equivalent fissile coefficient (B) based on the combustion guarantee method. (I.N.)

In situ bioremediation of JP-5 jet fuel

International Nuclear Information System (INIS)

Eisman, M.P.; Dorwin, E.; Barnes, D.; Nelson, B.

1991-01-01

Fuel leaks and spills of the jet fuel JP-5 at various Naval installations are required by law to be remediated. Use of microorganisms for fuel spill remediation is the focus of this paper, which examines biodegradation of JP-5 by means of CO 2 evolution in batch cultures. In particular, the aerobic biodegradation of fresh and weathered JP-5, along with a representative fuel mix of three pure compounds, is examined. Since microorganisms exist in aqueous environments, the solubility in water of fuels and fuel components is also examined. Other chemical properties of the complex mixture of hydrocarbons in JP-5 may affect bioavailability. This paper will also attempt to relate biodegradation to these properties, particularly water solubility and type of hydrocarbon

Cavitation problems in mixing devices of SNR-300 fuel elements

International Nuclear Information System (INIS)

Benemann, A.

1976-01-01

Because of a complex flow path within the mixing device developed for the fuel elements of the SNR-300, in order to determine the minimum allowable interval to the beginning of cavitation, experimental tests with the original geometry are necessary. These conclusions show that for cavitation values CV>=1,3 - in the model and prototype - no cavitation zones can form. For reactor conditions a maximum velocity of Vsub(max)=4,7m/sec is therefore allowable in the free annular space of the compensator unit which corresponds to a massflow of M=22,5kg/sec. A cavitation value of CV=1,5 can be figured for the 120% load factor (M=20,4kg/sec,T=560 0 C). The mixing device developed is free of cavitation under the present conditions in the SNR-300. The condition of the fully developed cavitation is evidenced by a white noise with frequencies of at least 2.000 - 300.000cps and a signal/noise ratio S/R>40dB. The pressure amplitudes dependent on frequency are propagated in the streaming fluid and are severely damped by the locally existing two-phase flow. The unstable range at the beginning of cavitation is characterized by frequencies of about f=15.000cps

A device for quantitative plutonium testing in mixed fuel by its neutron emission

International Nuclear Information System (INIS)

Gadzhiev, G.I.; Gorobets, A.K.; Golushko, V.V.; Dunaev, E.S.; Leshchenko, Yu.I.

1987-01-01

A device for quantitative plutonium testing in mixed fuel by its neutron emission is described. The method of ''assigned dead time'' for isolation of neutrons of spontaneous fission is used in the device. The main characteristics of the registrating equipment specifying the regime of measuring and affecting testing errors are presented. The results of spontaneous fission neutrons detection in the range up to 100 g of plutonium linearly depend on 240 Pu. Sensitivity of testing makes up about 28 pul./s per 1 g of 240 Pu

Performance evaluation of the Loviisa advanced type fuel rods

International Nuclear Information System (INIS)

Ranta-Puska, K.; Pihlatie, M.

2001-01-01

The fuel vendor TVEL has supplied to Loviisa WWER-440 power plant six lead assemblies of an advanced type which have profiling of the fuel enrichment, demountability of the assembly and a reduced shroud wall thickness. The pool side examination programme of these assemblies is underway including visual inspections, diameter and length measurements between operation cycles, and end-of-life fission gas release measurements, determined from 85 Kr activity in the plenum. Complementary evaluations and testing of models are done with the ENIGMA fuel performance code. The diameters of the corner rods have decreased to 30 μm during the first cycle and 40 to 70 μm after two cycles (with rod burnups of 24-30 MWd/kgU). The extent of creep-down is generally as expected, and agrees with the creep model adjusted for Russian Zr1%Nb cladding type and the Loviisa coolant and neutron flux conditions. The gap closure and reversed hoop strain are to be awaited during the third cycle so the new data will be an interesting validation exercise for the model and ENIGMA. Calculated temperatures stay low, and therefore low fission gas release fractions are anticipated as well

Mixed U/Pu oxide fuel fabrication facility co-processed feed, pelletized fuel

International Nuclear Information System (INIS)

1978-09-01

Two conceptual MOX fuel fabrication facilities are discussed in this study. The first facility in the main body of the report is for the fabrication of LWR uranium dioxide - plutonium dioxide (MOX) fuel using co-processed feed. The second facility in the addendum is for the fabrication of co-processed MOX fuel spiked with 60 Co. Both facilities produce pellet fuel. The spiked facility uses the same basic fabrication process as the conventional MOX plant but the fuel feed incorporates a high energy gamma emitter as a safeguard measure against diversion; additional shielding is added to protect personnel from radiation exposure, all operations are automated and remote, and normal maintenance is performed remotely. The report describes the fuel fabrication process and plant layout including scrap and waste processing; and maintenance, ventilation and safety measures

CORRIDOR-TYPE BAFFLED MIXING BASIN WITH CROSS POROUS BARRIERS

Directory of Open Access Journals (Sweden)

S. M. Epoyan

2018-02-01

Full Text Available Purpose. The paper hightlights the increase in operational efficiency of corridor-type baffled mixing basin by installing of cross porous barriers made of gravel (or other materials and epoxy resin, grade ED-20 (ED-16 with the hardener polyethylenepolyamine (PEPA, approved by Ukrainian Ministry of Health in systems of utility and drinking water supply. Methodology. The first stage of the experiments was performed on the model of the proposed mixer in scale 1:4 in order to determine the local resistance of the porous barrier, which is made of gravel with a size of 10-15 mm (average diameter 12.5 mm and thickness of 50 mm. The local resistance of the barrier was measured using piezometers installed before and after the porous barrier. The velocity of water motion in the corridor of the mixer was determined depending on the water consumption, incoming on the mixer accordingly to the water meter and by the volumetric method. Findings. In accordance with researches when the water flows at a velocity of 0.1 m/s in the corridor of the mixer, the head losses in the porous barrier is 17 cm (0.17 m, and at a velocity of 0.2 m/s–0.68 m. The resistance coefficient (ξ, which is equal to 333.2 for the investigated barrier, was determined experimentally. It allows determining the head losses in the porous barrier at other velocities of water motion. When the velocity of water motion in the corridors of the mixer is from 0.7 up to 0.5 m/s, head losses increase almost fourfold. The conducted researches allowed to develop a calculation methodology for corridor-type baffled mixing basin with porous polymer-concrete barriers. Originality. Authors developed and investigated the corridor-type baffled mixing basin with porous polymer-concrete barriers. These barriers allow increasing and regulating the intensity and time of reagents mixing with the initial water exactly in the barriers, improving the distribution of the flow through the section of the mixer

An improved characterization method for international accountancy measurements of fresh and irradiated mixed oxide (MOX) fuel: helping achieve continual monitoring and safeguards through the fuel cycle

Energy Technology Data Exchange (ETDEWEB)

Evans, Louise G [Los Alamos National Laboratory; Croft, Stephen [Los Alamos National Laboratory; Swinhoe, Martyn T [Los Alamos National Laboratory; Tobin, S. J. [Los Alamos National Laboratory; Menlove, H. O. [Los Alamos National Laboratory; Schear, M. A. [Los Alamos National Laboratory; Worrall, Andrew [U.K. NNL

2011-01-13

Nuclear fuel accountancy measurements are conducted at several points through the nuclear fuel cycle to ensure continuity of knowledge (CofK) of special nuclear material (SNM). Non-destructive assay (NDA) measurements are performed on fresh fuel (prior to irradiation in a reactor) and spent nuclear fuel (SNF) post-irradiation. We have developed a fuel assembly characterization system, based on the novel concept of 'neutron fingerprinting' with multiplicity signatures to ensure detailed CofK of nuclear fuel through the entire fuel cycle. The neutron fingerprint in this case is determined by the measurement of the various correlated neutron signatures, specific to fuel isotopic composition, and therefore offers greater sensitivity to variations in fissile content among fuel assemblies than other techniques such as gross neutron counting. This neutron fingerprint could be measured at the point of fuel dispatch (e.g. from a fuel fabrication plant prior to irradiation, or from a reactor site post-irradiation), monitored during transportation of the fuel assembly, and measured at a subsequent receiving site (e.g. at the reactor site prior to irradiation, or reprocessing facility post-irradiation); this would confirm that no unexpected changes to the fuel composition or amount have taken place during transportation and/ or reactor operations. Changes may indicate an attempt to divert material for example. Here, we present the current state of the practice of fuel measurements for both fresh mixed oxide (MOX) fuel and SNF (both MOX and uranium dioxide). This is presented in the framework of international safeguards perspectives from the US and UK. We also postulate as to how the neutron fingerprinting concept could lead to improved fuel characterization (both fresh MOX and SNF) resulting in: (a) assured CofK of fuel across the nuclear fuel cycle, (b) improved detection of SNM diversion, and (c) greater confidence in safeguards of SNF transportation.

Crossflow characteristics of flange type fuel element for very high temperature gas-cooled reactor

International Nuclear Information System (INIS)

Takizuka, Takakazu; Kaburaki, Hideo; Suzuki, Kunihiko; Nakamura, Masahide.

1987-01-01

Fuel element design incorporating mating flanges at block end faces has the potential to improve thermal hydraulic performance of a VHTR (very high temperature gas-cooled reactor) core. As part of research and development efforts to establish flange type fuel element design, experiments and analyses were carried out on crossflow through interface gap between elements. Air at atmospheric pressure and ambient temperature was used as a fluid. Crossflow loss coefficient factors were obtained with three test models, having different flange mating clearances, for various interface gap configurations, gap widths and block misalignments. It was found that crossflow loss coefficient factors for flange type fuel element were much larger than those for conventional flat-faced element. Numerical analyses were also made using a simple model devised to represent the crossflow path at the fuel element interface. The close agreement between numerical results and experimental data indicated that this model could predict well the crossflow characteristics of the flange type fuel element. (author)

Intense atmospheric pollution modifies weather: a case of mixed biomass burning with fossil fuel combustion pollution in eastern China

Science.gov (United States)

Ding, A. J.; Fu, C. B.; Yang, X. Q.; Sun, J. N.; Petäjä, T.; Kerminen, V.-M.; Wang, T.; Xie, Y.; Herrmann, E.; Zheng, L. F.; Nie, W.; Liu, Q.; Wei, X. L.; Kulmala, M.

2013-10-01

The influence of air pollutants, especially aerosols, on regional and global climate has been widely investigated, but only a very limited number of studies report their impacts on everyday weather. In this work, we present for the first time direct (observational) evidence of a clear effect of how a mixed atmospheric pollution changes the weather with a substantial modification in the air temperature and rainfall. By using comprehensive measurements in Nanjing, China, we found that mixed agricultural burning plumes with fossil fuel combustion pollution resulted in a decrease in the solar radiation intensity by more than 70%, a decrease in the sensible heat by more than 85%, a temperature drop by almost 10 K, and a change in rainfall during both daytime and nighttime. Our results show clear air pollution-weather interactions, and quantify how air pollution affects weather via air pollution-boundary layer dynamics and aerosol-radiation-cloud feedbacks. This study highlights cross-disciplinary needs to investigate the environmental, weather and climate impacts of the mixed biomass burning and fossil fuel combustion sources in East China.

Hexagonal type Ising nanowire with mixed spins: Some dynamic behaviors

International Nuclear Information System (INIS)

Kantar, Ersin; Kocakaplan, Yusuf

2015-01-01

The dynamic behaviors of a mixed spin (1/2–1) hexagonal Ising nanowire (HIN) with core–shell structure in the presence of a time dependent magnetic field are investigated by using the effective-field theory with correlations based on the Glauber-type stochastic dynamics (DEFT). According to the values of interaction parameters, temperature dependence of the dynamic magnetizations, the hysteresis loop areas and the dynamic correlations are investigated to characterize the nature (first- or second-order) of the dynamic phase transitions (DPTs). Dynamic phase diagrams, including compensation points, are also obtained. Moreover, from the thermal variations of the dynamic total magnetization, the five compensation types can be found under certain conditions, namely the Q-, R-, S-, P-, and N-types. - Highlights: • Dynamic behaviors of mixed spin HIN system are obtained within the EFT. • The system exhibits i, p and nm fundamental phases. • The dynamic phase diagrams are presented in (h, T), (D, T), (Δ S , T) and (r, T) planes. • The dynamic phase diagrams exhibit the dynamic tricritical point (TCP). • Different dynamic compensation types are obtained

NOT LOCAL PROBLEM OF TYPE OF THE PROBLEM BITSADZE – SAMARSKY FOR THE EQUATION THE MIXED TYPE IN UNLIMITED AREA

Directory of Open Access Journals (Sweden)

Zunnunov R.T.

2010-04-01

Full Text Available In this paper the existence and uniqueness of the solution of the nonlocal boundary value problem for the mixed type equation in unbounded domain are proved.In this paper the existence and uniqueness of the solution of the non-local boundary value problem for the mixed type equation in unbounded domain are proved.

Thermal hydraulic model validation for HOR mixed core fuel management

International Nuclear Information System (INIS)

Gibcus, H.P.M.; Vries, J.W. de; Leege, P.F.A. de

1997-01-01

A thermal-hydraulic core management model has been developed for the Hoger Onderwijsreactor (HOR), a 2 MW pool-type university research reactor. The model was adopted for safety analysis purposes in the framework of HEU/LEU core conversion studies. It is applied in the thermal-hydraulic computer code SHORT (Steady-state HOR Thermal-hydraulics) which is presently in use in designing core configurations and for in-core fuel management. An elaborate measurement program was performed for establishing the core hydraulic characteristics for a variety of conditions. The hydraulic data were obtained with a dummy fuel element with special equipment allowing a.o. direct measurement of the true core flow rate. Using these data the thermal-hydraulic model was validated experimentally. The model, experimental tests, and model validation are discussed. (author)

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

International Nuclear Information System (INIS)

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

2005-01-01

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

Sintering method for nuclear fuel pellet

International Nuclear Information System (INIS)

Omuta, Hirofumi; Nakabayashi, Shigetoshi.

1997-01-01

When sintering a compressed nuclear fuel powder in an atmosphere of a mixed gas comprising hydrogen and nitrogen, steams are added to the mixed gas to suppress the nitrogen content in sintered nuclear fuel pellets. In addition, the content of nitrogen impurities in the nuclear fuel pellets can be controlled by controlling the amount of steams to be added to the mixed gas, namely, by controlling the dew point as an index thereof. If the addition amount of steams to the mixed gas is determined by controlling the dew point as an index, the content of nitrogen impurities in the sintered nuclear fuel pellets can be controlled reliably to a specified value of 0.0075% or less. If ammonolyzed gas is used as the mixed gas, a more economical mixed gas can be obtained than in the case of forming mixed gas by mixing the hydrogen gas and the nitrogen gas. (N.H.)

Thorium Fuel Utilization Analysis on Small Long Life Reactor for Different Coolant Types

Science.gov (United States)

Permana, Sidik

2017-07-01

A small power reactor and long operation which can be deployed for less population and remote area has been proposed by the IAEA as a small and medium reactor (SMR) program. Beside uranium utilization, it can be used also thorium fuel resources for SMR as a part of optimalization of nuclear fuel as a “partner” fuel with uranium fuel. A small long-life reactor based on thorium fuel cycle for several reactor coolant types and several power output has been evaluated in the present study for 10 years period of reactor operation. Several key parameters are used to evaluate its effect to the reactor performances such as reactor criticality, excess reactivity, reactor burnup achievement and power density profile. Water-cooled types give higher criticality than liquid metal coolants. Liquid metal coolant for fast reactor system gives less criticality especially at beginning of cycle (BOC), which shows liquid metal coolant system obtains almost stable criticality condition. Liquid metal coolants are relatively less excess reactivity to maintain longer reactor operation than water coolants. In addition, liquid metal coolant gives higher achievable burnup than water coolant types as well as higher power density for liquid metal coolants.

Recent IAEA activities on CANDU-PHWR fuels and fuel cycles

International Nuclear Information System (INIS)

Inozemtsev, V.; Ganguly, C.

2005-01-01

Pressurized Heavy Water Reactors (PHWR), widely known as CANDU, are in operation in Argentina, Canada, China, India, Pakistan, Republic of Korea and Romania and account for about 6% of the world's nuclear electricity production. The CANDU reactor and its fuel have several unique features, like horizontal calandria and coolant tubes, on-power fuel loading, thin-walled collapsible clad coated with graphite on the inner surface, very high density (>96%TD) natural uranium oxide fuel and amenability to slightly enriched uranium oxide, mixed uranium plutonium oxide (MOX), mixed thorium plutonium oxide, mixed thorium uranium (U-233) oxide and inert matrix fuels. Several Technical Working Groups (TWG) of IAEA periodically discuss and review CANDU reactors, its fuel and fuel cycle options. These include TWGs on water-cooled nuclear power reactor Fuel Performance and Technology (TWGFPT), on Nuclear Fuel Cycle Options and spent fuel management (TWGNFCO) and on Heavy Water Reactors (TWGHWR). In addition, IAEA-INPRO project also covers Advanced CANDU Reactors (ACR) and DUPIC fuel cycles. The present paper summarises the Agency's activities in CANDU fuel and fuel cycle, highlighting the progress during the last two years. In the past we saw HWR and LWR technologies and fuel cycles separate, but nowadays their interaction is obviously growing, and their mutual influence may have a synergetic character if we look at the world nuclear fuel cycle as at an integrated system where the both are important elements in line with fast neutron, gas cooled and other advanced reactors. As an international organization the IAEA considers this challenge and makes concrete steps to tackle it for the benefit of all Member States. (author)