RA-3 core with uranium silicide fuel elements

International Nuclear Information System (INIS)

Abbate, Maximo J.; Sbaffoni, Maria M.

2000-01-01

Following on with studies on uranium silicide fuel elements, this paper reports some comparisons between the use of standard ECN [U 3 O 8 ] fuel elements and type P-06 [from U 3 Si 2 ] fuel elements in the RA-3 core.The first results showed that the calculated overall mean burn up is in agreement with that reported for the facility, which gives more confidence to the successive ones. Comparing the mentioned cores, the silicide one presents several advantages such as: -) a mean burn up increase of 18 %; -) an extraction burn up increase of 20 %; -) 37.4 % increase in full power days, for mean burn up. All this is meritorious for this fuel. Moreover, grouped and homogenized libraries were prepared for CITVAP code that will be used for planning experiments and other bidimensional studies. Preliminary calculations were also performed. (author)

Prospect of Uranium Silicide fuel element with hypostoichiometric (Si ≤3.7%)

International Nuclear Information System (INIS)

Suripto, A.; Sardjono; Martoyo

1996-01-01

An attempt to obtain high uranium-loading in silicide dispersion fuel element using the fabrication technology applicable nowadays can reach Uranium-loading slightly above 5 gU/cm 3 . It is difficult to achieve a higher uranium-loading than that because of fabricability constraints. To overcome those difficulties, the use of uranium silicide U 3 Si based is considered. The excess of U is obtained by synthesising U 3 Si 2 in Si-hypostoichiometric stage, without applying heat treatment to the ingot as it can generate undesired U 3 Si. The U U will react with the matrix to form U al x compound, that its pressure is tolerable. This experiment is to consider possibilities of employing the U 3 Si 2 as nuclear fuel element which have been performed by synthesising U 3 Si 2 -U with the composition of 3.7 % weigh and 3 % weigh U. The ingot was obtained and converted into powder form which then was fabricated into experimental plate nuclear fuel element. The interaction between free U and Al-matrix during heat-treatment is the rolling phase of the fuel element was observed. The study of the next phase will be conducted later

Irradiation behavior of uranium-silicide dispersion fuels

International Nuclear Information System (INIS)

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

1984-01-01

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

Irradiation of an uranium silicide prototype in RA-3 reactor; Irradiacion de un elemento combustible prototipo de siliciuro de uranio en el RA-3

Energy Technology Data Exchange (ETDEWEB)

Calabrese, R; Estrik, G; Notari, C [Comision Nacional de Energia Atomica, San Martin (Argentina). Unidad de Actividad Reactores y Centrales Nucleares

1997-12-31

The factibility of irradiation of an uranium silicide (U{sub 3} Si{sub 2}) prototype in the RA-3 reactor was studied. The standard RA-3 fuel element uses U{sub 3} O{sub 8} as fissible material. The enrichment of both standard and prototype is the same: 20% U{sub 235} and also the frame geometry and number of plates is identical. The differences are in the plate dimensions and the fissile content which is higher in the prototype. The cooling conditions of the core allow the insertion of the prototype in any core position, even near the water trap, if the overall power is kept below 5Mw. Nevertheless, the recommendation was to begin irradiation near the periphery and later on move the prototype towards more central positions in order to increase the burnup rate. The prototype was effectively introduced in a peripheral position and the thermal fluxes were measured between plates with the foil activation technique. These were also evaluated with the fuel management codes and a reasonable agreement was found. (author). 5 refs., 3 figs., 3 tabs.

Characterization of uranium silicide powder using XRD

International Nuclear Information System (INIS)

Garcia, Rafael H.L.; Saliba-Silva, Adonis M.; Carvalho, Elita F.U.; Lima, Nelson B.; Ichikawa, Rodrigo U.; Martinez, Luiz G.

2013-01-01

Uranium silicide (U 3 Si 2 ) is an intermetallic used as nuclear fuel in most modern MTR - Materials Test Reactor. Dispersed in aluminum, this fuel allows high uranium densities, up to 4.8 gU/cm 3 . At IPEN, the fabrication of fuel elements based on U 3 Si 2 for the IEA-R1 reactor is carried out in the Nuclear Fuel Center (CCN), by vacuum induction melting of uranium and silicon, followed by grinding. Before employed in a nuclear reactor, U 3 Si 2 must be submitted to a strict quality control, which includes granulometry, density, X-ray radiography for dispersion homogeneity, chemical and crystallographic characterization. Concerning phase composition for a qualified fuel, the fraction of U 3 Si 2 should be higher than 80wt.%. Aiming at the development of a routine methodology for quantification of phases via analysis of XRD data using the Rietved method, six samples from two production baths of CCN were submitted to X-ray diffraction. The data were analyzed using software GSAS and line profile analysis methods. The results suggest that fusion product have preferred orientation and grinding step is important for a better refinement. (author)

Information for irradiation and post-irradiation of the silicide fuel element prototype P-07

International Nuclear Information System (INIS)

Abbate, Maximo J.; Sbaffoni, Maria M.

2003-01-01

Included in the 'Silicides' Project, developed by the Nuclear Fuels Department of the National Atomic Energy Commission (CNEA), it is foreseen the qualification of this type of fuel for research reactors in order to be used in the Argentine RA-3 reactor and to confirm the CNEA as an international supplier. The paper presents basic information on several parameters corresponding to the new silicide prototype, called P-07, to be taken into account for its irradiation, postirradiation and qualification. (author)

Making of fission 99Mo from LEU silicide(s): A radiochemists' view

International Nuclear Information System (INIS)

Kolar, Z.I.; Wolterbeek, H.Th.

2005-01-01

The present-day industrial scale production of 99 Mo is fission based and involves thermal-neutron irradiation in research reactors of highly enriched uranium (HEU, > 20 % 235 U) containing targets, followed by radiochemical processing of the irradiated targets resulting in the final product: a 99 Mo containing chemical compound of molybdenum. In 1978 a program (RERTR) was started to develop a substitute for HEU reactor fuel i.e. a low enriched uranium (LEU, 235 U) one. In the wake of that program studies were undertaken to convert HEU into LEU based 99 Mo production. Both new targets and radiochemical treatments leading to 99 Mo compounds were proposed. One of these targets is based on LEU silicide, U 3 Si 2 . Present paper aims at comparing LEU U 3 Si 2 and LEU U 3 Si with another LEU target i.e. target material and arriving at some preferences pertaining to 99 Mo production. (author)

Milling uranium silicide powder for dispersion nuclear fuels

Energy Technology Data Exchange (ETDEWEB)

Vieira, E.; Silva, D.G.; Souza, J.A.B.; Durazzo, M. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Riella, H.G. [Universidade Federal de Santa Catarina (UFSC), Florianopolis, SC (Brazil)

2009-07-01

Full text: Uranium silicide (U3Si2) is presently considered the best fuel qualified so far in terms of uranium loading and performance. Stability of the U3Si2 fuel with uranium density of 4.8 g/cm3 was confirmed by burnup stability tests performed during the Reduced Enrichment for Research and Test Reactors (RERTR) program. This fuel was chosen to compose the first core of the new Brazilian Multipurpose Research Reactor (RMB), planned to be constructed in the next years. This new reactor will consume bigger quantities of U3Si2 powder, when compared with the small consumption of the IEA-R1 research reactor of IPEN-CNEN/SP, the unique MTR type research reactor operating in the country. At the present time, the milling operation of U3Si2 ingots is made manually. In order to increase the powder production capacity, the manual milling must be replaced by an automated procedure. This paper describes a new milling machine and procedure developed to produce U3Si2 powder with higher efficiency. (author)

Post-pulse detail metallographic examinations of low-enriched uranium silicide plate-type miniature fuel

International Nuclear Information System (INIS)

Yanagisawa, Kazuaki

1991-10-01

Pulse irradiation at Nuclear Safety Research Reactor (NSRR) was performed using low-enriched (19.89 w% 235 U) unirradiated silicide plate-type miniature fuel which had a density of 4.8 gU/cm 3 . Experimental aims are to understand the dimensional stability and to clarify the failure threshold of the silicide plate-type miniature fuel under power transient conditions through post-pulse detail metallographic examinations. A silicide plate-type miniature fuel was loaded into an irradiation capsule and irradiated by a single pulse. Deposited energies given in the experiments were 62, 77, 116 and 154 cal/g·fuel, which lead to corresponding peak fuel plate temperatures, 201 ± 28degC, 187 ± 10degC, 418 ± 74degC and 871 ± 74degC, respectively. Below 400degC, reliability and dimensional stability of the silicide plate fuel was sustained, and the silicide plate fuel was intact. Up to 540degC, wall-through intergranular crackings occurred in the Al-3%Mg alloy cladding. With the increase of the temperature, the melting of the aluminum cladding followed by recrystallization, the denudation of fuel core and the plate-through intergranular cracking were observed. With the increase of the temperature beyond 400degC, the bowing of fuel plate became significant. Above the temperature of 640degC molten aluminum partially reacted with the fuel core, partially flowed downward under the influence of surface tension and gravity, and partially formed agglomerations. Judging from these experimental observations, the fuel-plate above 400degC tends to reduce its dimensional stability. Despite of the apparent silicide fuel-plate failure, neither generation of pressure pulse nor that of mechanical energy occurred at all. (J.P.N.)

Solvent extraction studies of RERTR silicide fuels

International Nuclear Information System (INIS)

Gouge, Anthony P.

1983-01-01

Uranium silicide fuels, which are candidate RERTR fuel compositions, may require special considerations in solvent extraction reprocessing. Since Savannah River Plant may be reprocessing RERTR fuels as early as 1985, studies have been conducted at Savannah River Laboratory to demonstrate the solvent extraction behavior of this fuel. Results of solvent extraction studies with both unirradiated and irradiated fuel are presented along with the preliminary RERTR solvent extraction reprocessing flow sheet for Savannah River Plant. (author)

Trace metal assay of uranium silicide fuel

International Nuclear Information System (INIS)

Kulkarni, M.J.; Argekar, A.A.; Thulasidas, S.K.; Dhawale, B.A.; Rajeswari, B.; Adya, V.C.; Purohit, P.J.; Neelam, G.; Bangia, T.R.; Page, A.G.; Sastry, M.D.; Iyer, R.H.

1994-01-01

A comprehensive trace metal assay of uranium silicide, a fuel for nuclear research reactors that employs low-enrichment uranium, is carried out by atomic spectrometry. Of the list of specification elements, 21 metallic elements are determined by a direct current (dc) arc carrier distillation technique; the rare earths yttrium and zirconium are chemically separated from the major matrix followed by a dc arc/inductively coupled argon plasma (ICP) excitation technique in atomic emission spectrometry (AES); silver is determined by electrothermal atomization-atomic absorption spectrometry (ETA-AAS) without prior chemical separation of the major matrix. Gamma radioactive tracers are used to check the recovery of rare earths during the chemical separation procedure. The detection limits for trace metallics vary in the 0.1- to 40-ppm range. The precision of the determinations as evaluated from the analysis of the synthetic sample with intermediate range analyte concentration is better than 25% relative standard deviation (RSD) for most of the elements employing dc arc-AES, while that for silver determination by ETS-AAS is 10% RSD. The precision of the determinations for four crucially important rare earths by ICP-AES is better than 3% RSD

Irradiation mixing of Al into U3Si

International Nuclear Information System (INIS)

Birtcher, R.C.; Ding, F.R.; Kestel, B.J.; Baldo, P.M.; Zaluzec, N.J.

1995-11-01

Thermal and irradiation induced intermixing of uranium silicide reactor fuels with the aluminum cladding is an important consideration in understanding their fission gas and fuel swelling behavior. The authors have used Rutherford backscattering to follow the behavior of an Al thin film on U 3 Si and U 3 Si 2 during 1.5 MeV Kr ion irradiation at temperatures of 30 and 350 C. After an initial dose during which no intermixing occurs, the Al mixes quickly into U 3 Si. The threshold dose is believed to be associated with an oxide layer between the Al and the uranium silicide. At 300 C and doses greater than threshold, rates of mixing and aluminide phase growth are extracted

The fabrication and performance of Canadian silicide dispersion fuel for test reactors

International Nuclear Information System (INIS)

Sears, D.F.; Wood, J.C.; Berthiaume, L.C.; Herbert, L.N.; Schaefer, J.D.

1985-01-01

Fuel fabrication effort is now concentrated on the commissioning of large-scale process equipment, defining product specifications, developing a quality assurance plan, and setting up a mini-computer material accountancy system. In the irradiation testing program, full-size NRU assemblies containing 20% enriched silicide dispersion fuel have been Irradiated successfully to burnups in the range 65-80 atomic percent. Irradiations have also been conducted on mini-elements having 1.2 mm diameter holes In their mid-sections, some drilled before irradiation and others after irradiation to 22-83 atomic percent burnup. Uranium was lost to the coolant in direct proportion to the surface area of exposed core material. Pre-irradiation in the intact condition appeared to reduce in-reactor corrosion. Fuel cores developed for the NRU reactor are dimensionally very stable, swelling by only 6-8% at the very high burnup of 93 atomic percent. Two important factors contributing to this good performance are cylindrical clad restraint and coarse silicide particles. Thermal ramping tests were conducted on irradiated silicide aspersion fuels. Small segments of fuel cores released 85 Kr starting at about 520 deg. C and peaking at about 680 deg C. After a holding period of 1 hour at 720 deg. C a secondary 85 Kr peak occurred during cooling (at about 330 deg. C) probably due to thermal contraction cracking. Whole mini-elements irradiated to 93 atomic percent burnup were also ramped thermally, with encouraging results. After about 0.25 h at 530 deg. C the aluminum cladding developed very localized small blisters, some with penetrating pin-hole cracks preventing gross pillowing or ballooning. (author)

The fabrication and performance of Canadian silicide dispersion fuel for test reactors

Energy Technology Data Exchange (ETDEWEB)

Sears, D F; Wood, J C; Berthiaume, L C; Herbert, L N; Schaefer, J D

1985-07-01

Fuel fabrication effort is now concentrated on the commissioning of large-scale process equipment, defining product specifications, developing a quality assurance plan, and setting up a mini-computer material accountancy system. In the irradiation testing program, full-size NRU assemblies containing 20% enriched silicide dispersion fuel have been Irradiated successfully to burnups in the range 65-80 atomic percent. Irradiations have also been conducted on mini-elements having 1.2 mm diameter holes In their mid-sections, some drilled before irradiation and others after irradiation to 22-83 atomic percent burnup. Uranium was lost to the coolant in direct proportion to the surface area of exposed core material. Pre-irradiation in the intact condition appeared to reduce in-reactor corrosion. Fuel cores developed for the NRU reactor are dimensionally very stable, swelling by only 6-8% at the very high burnup of 93 atomic percent. Two important factors contributing to this good performance are cylindrical clad restraint and coarse silicide particles. Thermal ramping tests were conducted on irradiated silicide aspersion fuels. Small segments of fuel cores released {sup 85}Kr starting at about 520 deg. C and peaking at about 680 deg C. After a holding period of 1 hour at 720 deg. C a secondary {sup 85}Kr peak occurred during cooling (at about 330 deg. C) probably due to thermal contraction cracking. Whole mini-elements irradiated to 93 atomic percent burnup were also ramped thermally, with encouraging results. After about 0.25 h at 530 deg. C the aluminum cladding developed very localized small blisters, some with penetrating pin-hole cracks preventing gross pillowing or ballooning. (author)

Study of the recrystallisation of irradiated uranium; Etude sur l'uranium irradie

Energy Technology Data Exchange (ETDEWEB)

Bloch, J; Mustelier, J P; Bussy, P; Blin, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1958-07-01

1- Study of the recrystallisation of irradiated uranium. The recrystallisation of uranium irradiated to a burnup level of 220 MWj/t, at a temperature of the order of 350 deg. C, has been investigated. The observations were made chiefly by means of micrography an hardness measurements. If the irradiated metal is compared with a cold-drawn metal showing the same shearing of the twinned crystals, and therefore the same rate of plastic deformation, as the irradiated metal, it is noted that the restoring of the irradiated metal takes place at a considerably higher temperature than that of the cold-drawn metal. Pre-crystallisation is very much delayed. Only, a passage of the {alpha}-{beta} transformation point quickly wipes out irradiation effect. 2- Hardening of uranium by irradiation. Using hardness measurements we have studied more especially the effect of very weak irradiations on uranium (integrated flux < 10{sup 16} nvt). The hardness does not increase linearly with the flux, but a period of incubation is observed probably representing the time necessary for saturation of the dislocations. (author)Fren. [French] 1- Etude de la recristallisation de l'uranium irradie. On a etudie la recristallisation d'uranium irradie jusqu'a un taux de combustion de 220 MWj/t a une temperature de l'ordre de 350 deg. C. Les observations ont ete faites principalement a l'aide de la micrographie et de la durete. Si l'on compare le metal irradie avec un metal ecroui presentant le meme cisaillement des macles, donc le meme taux de deformation plastique que le metal irradie, on constate que la restauration du metal irradie se fait a une temperature notablement superieure a celle du metal ecroui. La recristallisation est tres retardee. Seul, un passage du point de transformation {alpha}-{beta} efface rapidement l'effet de l'irradiation. 2- Durcissement de l'uranium par irradiation. Nous avons, a l'aide de la durete, etudie plus particulierement l'effet de tres faibles irrtions sur l'uranium

Neutronic study on conversion of SAFARI-1 to LEU silicide fuel

International Nuclear Information System (INIS)

Ball, G.; Pond, R.; Hanan, N.; Matos, J.

1995-01-01

This paper marks the initial study into the technical and economic feasibility of converting the SAFARI-1 reactor in South Africa to LEU silicide fuel. Several MTR assembly geometries and LEU uranium densities have been studied and compared with MEU and HEU fuels. Two factors of primary importance for conversion of SAFARI-1 to LEU fuel are the economy of the fuel cycle and the performance of the incore and excore irradiation positions

A Study on Silicide Coatings as Diffusion barrier for U-7Mo Dispersion Fuel

Energy Technology Data Exchange (ETDEWEB)

Won, Ju Jin; Kim, Sung Hwan; Lee, Kyu Hong; Jeong, Yong Jin; Kim, Ki Nam; Park, Jong Man; Lee, Chong Tak [KAERI, Daejeon (Korea, Republic of)

2016-05-15

Gamma phase U-Mo alloys are regarded as one of the promising candidates for advanced research reactor fuel when it comes to the irradiation performance. However, it has been reported that interaction layer formation between the UMo alloys and Al matrix degrades the irradiation performance of U-Mo dispersion fuel. The excessive interaction between the U-Mo alloys and their surrounding Al matrix lead to excessive local swelling called 'pillowing'. For this reason, KAERI suggested several remedies such as alloying U-Mo with Al matrix with Si. In addition, silicide or nitride coatings on the surface of U-Mo particles have also been proposed to hinder the growth of the interaction layer. In this study, centrifugally atomized U-7Mo alloy powders were coated with silicide layers at 900 .deg. C for 1hr. U-Mo alloy powder was mixed with MoSi{sub 2}, Si and ZrSi{sub 2} powders and subsequently heat-treated to form uranium-silicide coating layers on the surface of U-Mo alloy particles. Silicide coated U-Mo powders and characterized using scanning electron microscopy (SEM), energy dispersive x-ray spectroscopy (EDS) and X-ray diffractometer (XRD). The ZrSi{sub 2} coating layers has a thickness of about 1∼ 2μm. The surface of a silicide coated particle was very rough and silicide powder attached to the surface of the coating layer. 3. The XRD analysis of the coating layers showed that, they consisted of compounds such as U3Si{sub 2}, USi{sub 2}.

Behaviour of irradiated uranium silicide fuel revisited

International Nuclear Information System (INIS)

Finlay, M. Ross; Hofman, Gerard L.; Rest, Jeffrey; Snelgrove, James L.

2002-01-01

Irradiated U 3 Si 2 dispersion fuels demonstrate very low levels of swelling, even at extremely high burn-up. This behaviour is attributed to the stability of fission gas bubbles that develop during irradiation. The bubbles remain uniformly distributed throughout the fuel and show no obvious signs of coalescence. Close examination of high burn-up samples during the U 3 Si 2 qualification program revealed a bimodal distribution of fission gas bubbles. Those observations suggested that an underlying microstructure was responsible for the behaviour. An irradiation induced recrystallisation model was developed that relied on the presence of sufficient grain boundary surface to trap and pin fission gas bubbles and prevent coalescence. However, more recent work has revealed that the U 3 Si 2 becomes amorphous almost instantaneously upon irradiation. Consequently, the recrystallisation model does not adequately explain the nucleation and growth of fission gas bubbles in U 3 Si 2 . Whilst it appears to work well within the range of measured data, it cannot be relied on to extrapolate beyond that range since it is not mechanistically valid. A review of the mini-plates irradiated in the Oak Ridge Research Reactor from the U 3 Si 2 qualification program has been performed. This has yielded a new understanding of U 3 Si 2 behaviour under irradiation. (author)

Basic research on high-uranium density fuels for research and test reactors

International Nuclear Information System (INIS)

Ugajin, M.; Itoh, A.; Akabori, M.

1992-01-01

High-uranium density fuels, uranium silicides (U 3 Si 2 , U 3 Si) and U 6 Me-type uranium alloys (Me = Fe, Mn, Ni), were prepared and examined metallurgically as low-enriched uranium (LEU) fuels for research and test reactors. Miniature aluminum-dispersion plate-type fuel (miniplate) and aluminum-clad disk-type fuel specimens were fabricated and subjected to the neutron irradiation in JMTR (Japan Materials Testing Reactor). Fuel-aluminum compatibility tests were conducted to elucidate the extent of reaction and to identify reaction products. The relative stability of the fuels in an aluminum matrix was established at 350degC or above. Experiments were also performed to predict the chemical form of the solid fission-products in the uranium silicide (U 3 Si 2 ) simulating a high burnup anticipated for reactor service. (author)

Surface morphology of erbium silicide

International Nuclear Information System (INIS)

Lau, S.S.; Pai, C.S.; Wu, C.S.; Kuech, T.F.; Liu, B.X.

1982-01-01

The surface of rare-earth silicides (Er, Tb, etc.), formed by the reaction of thin-film metal layers with a silicon substrate, is typically dominated by deep penetrating, regularly shaped pits. These pits may have a detrimental effect on the electronic performance of low Schottky barrier height diodes utilizing such silicides on n-type Si. This study suggests that contamination at the metal-Si or silicide-Si interface is the primary cause of surface pitting. Surface pits may be reduced in density or eliminated entirely through either the use of Si substrate surfaces prepared under ultrahigh vacuum conditions prior to metal deposition and silicide formation or by means of ion irradiation techniques. Silicide layers formed by these techniques possess an almost planar morphology

Irradiated uranium reprocessing

International Nuclear Information System (INIS)

Gal, I.

1961-12-01

Task concerned with reprocessing of irradiated uranium covered the following activities: implementing the method and constructing the cell for uranium dissolving; implementing the procedure for extraction of uranium, plutonium and fission products from radioactive uranium solutions; studying the possibilities for using inorganic ion exchangers and adsorbers for separation of U, Pu and fission products

Status of the atomized uranium silicide fuel development at KAERI

Energy Technology Data Exchange (ETDEWEB)

Kim, C.K.; Kim, K.H.; Park, H.D.; Kuk, I.H. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1997-08-01

While developing KMRR fuel fabrication technology an atomizing technique has been applied in order to eliminate the difficulties relating to the tough property of U{sub 3}Si and to take advantage of the rapid solidification effect of atomization. The comparison between the conventionally comminuted powder dispersion fuel and the atomized powder dispersion fuel has been made. As the result, the processes, uranium silicide powdering and heat treatment for U{sub 3}Si transformation, become simplified. The workability, the thermal conductivity and the thermal compatibility of fuel meat have been investigated and found to be improved due to the spherical shape of atomized powder. In this presentation the overall developments of atomized U{sub 3}Si dispersion fuel and the planned activities for applying the atomizing technique to the real fuel fabrication are described.

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

International Nuclear Information System (INIS)

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

2000-01-01

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

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

Science.gov (United States)

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

2011-12-01

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

BASIC program to compute uranium density and void volume fraction in laboratory-scale uranium silicide aluminum dispersion plate-type fuel

International Nuclear Information System (INIS)

Ugajin, Mitsuhiro

1991-05-01

BASIC program simple and easy to operate has been developed to compute uranium density and void volume fraction for laboratory-scale uranium silicide aluminum dispersion plate-type fuel, so called miniplate. An example of the result of calculation is given in order to demonstrate how the calculated void fraction correlates with the microstructural distribution of the void in a miniplate prepared in our laboratory. The program is also able to constitute data base on important parameters for miniplates from experimentally-determined values of density, weight of each constituent and dimensions of miniplates. Utility programs pertinent to the development of the BASIC program are also given which run in the popular MS-DOS environment. All the source lists are attached and brief description for each program is made. (author)

Analysis of reactivity accidents of the RSG-GAS core with silicide fuel

International Nuclear Information System (INIS)

Tukiran

2002-01-01

The fuels of RSG-GAS reactor is changed from uranium oxide to uranium silicide. For time being, the fuel of RSG-GAS core are mixed up between oxide and silicide fuels with 250 gr of loading and 2.96 g U/cm 3 of density, respectively. While, silicide fuel with 300 gr of loading is still under research. The advantages of silicide fuels are can be used in high density, so that, it can be stayed longer in the core at higher burn-up, therefore, the length of cycle is longer. The silicide fuel in RSG-GAS core is used in step-wise by using mixed up core. Firstly, it is used silicide fuel with 250 gr of loading and then, silicide fuel with 300 gr of loading (3.55 g U/cm 3 of density). In every step-wise of fuel loading must be analysed its safety margin. In this occasion, it is analysed the reactivity accident of RSG-GAS core with 300 gr of silicide fuel loading. The calculation was done by using POKDYN code which available at P2TRR. The calculation was done by reactivity insertion at start up and power rangers. From all cases which were have been done, the results of analysis showed that there is no anomaly and safety margin break at RSG-GAS core with 300 gr silicide fuel loading

Development of high uranium-density fuels for use in research reactors

International Nuclear Information System (INIS)

Ugajin, Mitsuhiro; Akabori, Mitsuo; Itoh, Akinori

1996-01-01

The uranium silicide U 3 Si 2 possesses uranium density 11.3 gU/cm 3 with a congruent melting point of 1665degC, and is now successfully in use as a research reactor fuel. Another uranium silicide U 3 Si and U 6 Me-type uranium alloys (Me=Fe,Mn,Ni) have been chosen as new fuel materials because of the higher uranium densities 14.9 and 17.0 gU/cm 3 , respectively. Experiments were carried out to fabricate miniature aluminum-dispersion plate-type and aluminum-clad disk-type fuels by using the conventional picture-frame method and a hot-pressing technique, respectively. These included the above-mentioned new fuel materials as well as U 3 Si 2 . Totally 14 miniplates with uranium densities from 4.0 to 6.3 gU/cm 3 of fuel meat were prepared together with 28 disk-type fuel containing structurally-modified U 3 Si, and subjected to the neutron irradiation in JMTR (Japan Materials Testing Reactor). Some results of postirradiation examinations are presented. (author)

Rate Theory Modeling and Simulations of Silicide Fuel at LWR Conditions

Energy Technology Data Exchange (ETDEWEB)

Miao, Yinbin [Argonne National Lab. (ANL), Argonne, IL (United States); Ye, Bei [Argonne National Lab. (ANL), Argonne, IL (United States); Mei, Zhigang [Argonne National Lab. (ANL), Argonne, IL (United States); Hofman, Gerard [Argonne National Lab. (ANL), Argonne, IL (United States); Yacout, Abdellatif [Argonne National Lab. (ANL), Argonne, IL (United States)

2015-12-10

Uranium silicide (U₃Si₂) fuel has higher thermal conductivity and higher uranium density, making it a promising candidate for the accident-tolerant fuel (ATF) used in light water reactors (LWRs). However, previous studies on the fuel performance of U₃Si₂, including both experimental and computational approaches, have been focusing on the irradiation conditions in research reactors, which usually involve low operation temperatures and high fuel burnups. Thus, it is important to examine the fuel performance of U₃Si₂ at typical LWR conditions so as to evaluate the feasibility of replacing conventional uranium dioxide fuel with this silicide fuel material. As in-reactor irradiation experiments involve significant time and financial cost, it is appropriate to utilize modeling tools to estimate the behavior of U₃Si₂ in LWRs based on all those available research reactor experimental references and state-of-the-art density functional theory (DFT) calculation capabilities at the early development stage. Hence, in this report, a comprehensive investigation of the fission gas swelling behavior of U₃Si₂ at LWR conditions is introduced. The modeling efforts mentioned in this report was based on the rate theory (RT) model of fission gas bubble evolution that has been successfully applied for a variety of fuel materials at devious reactor conditions. Both existing experimental data and DFT-calculated results were used for the optimization of the parameters adopted by the RT model. Meanwhile, the fuel-cladding interaction was captured by the coupling of the RT model with simplified mechanical correlations. Therefore, the swelling behavior of U₃Si₂ fuel and its consequent interaction with cladding in LWRs was predicted by the rate theory modeling, providing valuable information for the development of U₃Si₂ fuel as an accident

Evaluation Of Oxide And Silicide Mixed Fuels Of The RSG-GAS Core

International Nuclear Information System (INIS)

Tukiran; Sembiring, Tagor Malem; Suparlina, Lily

2000-01-01

Fuel exchange of the RSG-GAS reactor core from uranium oxide to uranium silicide in the same loading, density, and enrichment, that is 250 gr, 2.98 gr/cm 3 , and 19.75%, respectively, will be performed in-step wise. In every cycle of exchange with 5/1 mode, it is needed to evaluate the parameter of reactor core operation. The parameters of the reactor operation observed are criticality mass of fuels, reactivity balance, and fuel reactivity that give effect to the reactor operation. The evaluation was done at beginning of cycle of the first and second transition core with compared between experiment and calculation results. The experiments were performed at transition core I and II, BOC, and low power. At transition core I, there are 2 silicide fuels (RI-224 and R1-225) in the core and then, added five silicide fuels (R1-226, R1-252, R1-263, and R1-264) to the core, so that there are seven silicide fuels in the transition core II. The evaluation was done based on the experiment of criticality, control rod calibration, fuel reactivity of the RSG-GAS transition core. For inserting 2 silicide fuels in the transition core I dan 7 fuels in the transition core II, the operation of RSG-GAS core fulfilled the safety margin and the parameter of reactor operation change is not occur drastically in experiment and calculation results. So that, the reactor was operated during 36 days at 15 MW, 540 MWD at the first transition core. The general result showed that the parameter of reactor operation change is small so that the fuel exchange from uranium oxide to uranium silicide in the next step can be done

Behaviour of uranium under irradiation; Comportement de l'uranium sous irradiation

Energy Technology Data Exchange (ETDEWEB)

Adda, Y; Mustelier, J P; Quere, Y [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

1964-07-01

The main results obtained in a study of the formation of defects caused in uranium by fission at low temperature are reported. By irradiation at 20 K. it was possible to determine the number of Frenkel pairs produced by one fission. An analysis of the curves giving the variations in electrical resistivity shows the size of the displacement spikes and the mechanism of defect creation due to fission. Irradiations at 77 K gave additional information, showing behaviour differences in the case of recrystallised and of cold worked uranium. The diffusion of rare gases was studied using metal-rare gas alloys obtained by electrical discharge, and samples of irradiated uranium. Simple diffusion is only responsible for the release of the rare gases under vacuum in cases where the rare gas content is very low (very slightly irradiated U). On the other hand when the concentration is higher (samples prepared by electrical discharge) the gas is given off by the formation, growth and coalescence of bubbles; the apparent diffusion coefficient is then quite different from the true coefficient and cannot be used in calculations on swelling. The various factors governing the phenomenon of simple diffusion were examined. It was shown in particular that a small addition of molybdenum could reduce the diffusion coefficient by a factor of 100. The precipitation of gas in uranium (Kr), in silver (Kr) and in Al-Li alloy (He) have been followed by measurement of the crystal parameter and of the electrical resistivity, and by electron microscope examination of thin films. The important part played by dislocations in the generation and growth of bubbles has been demonstrated, and it has been shown also that precipitation of bubbles on the dislocation lattice could block the development of recrystallisation. The results of these studies were compared with observations made on the swelling of uranium and uranium alloys U Mo and U Nb strongly irradiated between 400 and 700 C. In the case of Cubic

Preparation of U-Si/U-Me (Me = Fe, Ni, Mn) aluminum-dispersion plate-type fuel (miniplates) for capsule irradiation

International Nuclear Information System (INIS)

Ugajin, Mitsuhiro; Itoh, Akinori; Akabori, Mitsuo

1993-06-01

Details of equipment installed, method adopted and final products were described on the preparation of uranium silicides and other fuels for capsule irradiation. Main emphasis was placed on the preparation of laboratory-scale aluminum-dispersion plate-type fuel (miniplates) loaded to the first and second JMTR silicide capsules. Fuels contained in the capsules are as follows: (A) uranium-silicide base alloys U 3 Si 2 , Mo- added U 3 Si 2 , U 3 Si 2 +U 3 Si, U 3 Si 2 +USi, U 3 Si, U 3 (Si 0.8 Ge 0.2 ), U 3 (Si 0.6 Ge 0.4 ) (B) U 6 Me-type alloys with higher uranium density U 6 Mn, U 6 Ni, U 6 (Fe 0.4 Ni 0.6 ), U 6 (Fe 0.6 Mn 0.4 ) The powder-metallurgical picture-frame method was adopted and laboratory-scale technique was established for the preparation of miniplates. As a result of inspection for capsule irradiation, miniplates were prepared to meet the requirements of specification. (author)

Stability of uranium silicides during high energy ion irradiation

International Nuclear Information System (INIS)

Birtcher, R.C; Wang, L.M.

1991-11-01

Changes induced by 1.5 MeV Kr ion irradiation of both U 3 Si and U 3 Si 2 have been followed by in situ transmission electron microscopy. When irradiated at sufficiently low temperatures, both alloys transform from the crystalline to the amorphous state. When irradiated at temperatures above the temperature limit for ion beam amorphization, both compounds disorder with the Martensite twin structure in U 3 Si disappearing from view in TEM. Prolonged irradiation of the disordered crystalline phases results in nucleation of small crystallites within the initially large crystal grains. The new crystallites increase in number during continued irradiation until a fine grain structure is formed. Electron diffraction yields a powder-like diffraction pattern that indicates a random alignment of the small crystallites. During a second irradiation at lower temperatures, the small crystallizes retard amorphization. After 2 dpa at high temperatures, the amorphization dose is increased by over twenty times compared to that of initially unirradiated material

Determination of irradiated reactor uranium in soil samples in Belarus using 236U as irradiated uranium tracer.

Science.gov (United States)

Mironov, Vladislav P; Matusevich, Janna L; Kudrjashov, Vladimir P; Boulyga, Sergei F; Becker, J Sabine

2002-12-01

This work presents experimental results on the distribution of irradiated reactor uranium from fallout after the accident at Chernobyl Nuclear Power Plant (NPP) in comparison to natural uranium distribution in different soil types. Oxidation processes and vertical migration of irradiated uranium in soils typical of the 30 km relocation area around Chernobyl NPP were studied using 236U as the tracer for irradiated reactor uranium and inductively coupled plasma mass spectrometry as the analytical method for uranium isotope ratio measurements. Measurements of natural uranium yielded significant variations of its concentration in upper soil layers from 2 x 10(-7) g g(-1) to 3.4 x 10(-6) g g(-1). Concentrations of irradiated uranium in the upper 0-10 cm soil layers at the investigated sampling sites varied from 5 x 10(-12) g g(-1) to 2 x 10(-6) g g(-1) depending on the distance from Chernobyl NPP. In the majority of investigated soil profiles 78% to 97% of irradiated "Chernobyl" uranium is still contained in the upper 0-10 cm soil layers. The physical and chemical characteristics of the soil do not have any significant influence on processes of fuel particle destruction. Results obtained using carbonate leaching of 236U confirmed that more than 60% of irradiated "Chernobyl" uranium is still in a tetravalent form, ie. it is included in the fuel matrix (non-oxidized fuel UO2). The average value of the destruction rate of fuel particles determined for the Western radioactive trace (k = 0.030 +/- 0.005 yr(-1)) and for the Northern radioactive trace (k = 0.035 + 0.009 yr(-1)) coincide within experimental errors. Use of leaching of fission products in comparison to leaching of uranium for study of the destruction rate of fuel particles yielded poor coincidence due to the fact that use of fission products does not take into account differences in the chemical properties of fission products and fuel matrix (uranium).

Development of high uranium-density fuels for use in research reactors

Energy Technology Data Exchange (ETDEWEB)

Ugajin, Mitsuhiro; Akabori, Mitsuo; Itoh, Akinori [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1996-02-01

The uranium silicide U{sub 3}Si{sub 2} possesses uranium density 11.3 gU/cm{sup 3} with a congruent melting point of 1665degC, and is now successfully in use as a research reactor fuel. Another uranium silicide U{sub 3}Si and U{sub 6}Me-type uranium alloys (Me=Fe,Mn,Ni) have been chosen as new fuel materials because of the higher uranium densities 14.9 and 17.0 gU/cm{sup 3}, respectively. Experiments were carried out to fabricate miniature aluminum-dispersion plate-type and aluminum-clad disk-type fuels by using the conventional picture-frame method and a hot-pressing technique, respectively. These included the above-mentioned new fuel materials as well as U{sub 3}Si{sub 2}. Totally 14 miniplates with uranium densities from 4.0 to 6.3 gU/cm{sup 3} of fuel meat were prepared together with 28 disk-type fuel containing structurally-modified U{sub 3}Si, and subjected to the neutron irradiation in JMTR (Japan Materials Testing Reactor). Some results of postirradiation examinations are presented. (author)

First-principles investigations of the physical properties of binary uranium silicide alloys

International Nuclear Information System (INIS)

Yang, Jin; Long, Jianping; Yang, Lijun; Li, Dongmei

2013-01-01

Graphical abstract: Total density of states for USi 2 . Display Omitted -- Abstract: The structural, elastic properties and the Debye temperature of binary Uranium Silicide (U-Si) alloys are investigated by using the first-principles plane-wave pseudopotential density function theory within the generalized gradient approximation (GGA). The ground states properties are found to agree with the available experimental data. The mechanical properties like shear modulus, Young’s modulus, Poisson’s ratio σ and ratio B/G are also calculated. Finally, The averaged sound velocity (v m ), the longitudinal sound velocity (v l ), transverse sound velocity (v t ) and the Debye temperature (θ D ) are obtained. However, the theoretical values are slightly different from few existed experiment data because the latter was obtained at room temperature while the former one at 0 K

Reprocessing of LEU silicide fuel at Dounreay

International Nuclear Information System (INIS)

Cartwright, P.

1996-01-01

UKAEA have recently reprocessed two LEU silicide fuel elements in their MTR fuel reprocessing plant at Dounreay. The reprocessing was undertaken to demonstrate UKAEA's commitment to the world-wide research reactor communities future needs. Reprocessing of LEU silicide fuel is seen as a waste treatment process, resulting in the production of a liquid feed suitable for conditioning in a stable form of disposal. The uranium product from the reprocessing can be used as a blending feed with the HEU to produce LEU for use in the MTR cycle. (author)

Study of the recrystallisation of irradiated uranium

International Nuclear Information System (INIS)

Bloch, J.; Mustelier, J.P.; Bussy, P.; Blin, J.

1958-01-01

1- Study of the recrystallisation of irradiated uranium. The recrystallisation of uranium irradiated to a burnup level of 220 MWj/t, at a temperature of the order of 350 deg. C, has been investigated. The observations were made chiefly by means of micrography an hardness measurements. If the irradiated metal is compared with a cold-drawn metal showing the same shearing of the twinned crystals, and therefore the same rate of plastic deformation, as the irradiated metal, it is noted that the restoring of the irradiated metal takes place at a considerably higher temperature than that of the cold-drawn metal. Pre-crystallisation is very much delayed. Only, a passage of the α-β transformation point quickly wipes out irradiation effect. 2- Hardening of uranium by irradiation. Using hardness measurements we have studied more especially the effect of very weak irradiations on uranium (integrated flux 16 nvt). The hardness does not increase linearly with the flux, but a period of incubation is observed probably representing the time necessary for saturation of the dislocations. (author) [fr

Behaviour of uranium under irradiation

International Nuclear Information System (INIS)

Adda, Y.; Mustelier, J.P.; Quere, Y.; Commissariat a l'Energie Atomique, Fontenay-aux-Roses

1964-01-01

The main results obtained in a study of the formation of defects caused in uranium by fission at low temperature are reported. By irradiation at 20 K. it was possible to determine the number of Frenkel pairs produced by one fission. An analysis of the curves giving the variations in electrical resistivity shows the size of the displacement spikes and the mechanism of defect creation due to fission. Irradiations at 77 K gave additional information, showing behaviour differences in the case of recrystallised and of cold worked uranium. The diffusion of rare gases was studied using metal-rare gas alloys obtained by electrical discharge, and samples of irradiated uranium. Simple diffusion is only responsible for the release of the rare gases under vacuum in cases where the rare gas content is very low (very slightly irradiated U). On the other hand when the concentration is higher (samples prepared by electrical discharge) the gas is given off by the formation, growth and coalescence of bubbles; the apparent diffusion coefficient is then quite different from the true coefficient and cannot be used in calculations on swelling. The various factors governing the phenomenon of simple diffusion were examined. It was shown in particular that a small addition of molybdenum could reduce the diffusion coefficient by a factor of 100. The precipitation of gas in uranium (Kr), in silver (Kr) and in Al-Li alloy (He) have been followed by measurement of the crystal parameter and of the electrical resistivity, and by electron microscope examination of thin films. The important part played by dislocations in the generation and growth of bubbles has been demonstrated, and it has been shown also that precipitation of bubbles on the dislocation lattice could block the development of recrystallisation. The results of these studies were compared with observations made on the swelling of uranium and uranium alloys U Mo and U Nb strongly irradiated between 400 and 700 C. In the case of Cubic

Irradiated uranium reprocessing; Prerada ozracenog urana

Energy Technology Data Exchange (ETDEWEB)

Gal, I [Institute of Nuclear Sciences Boris Kidric, Laboratorijaza visoku aktivnost, Vinca, Beograd (Serbia and Montenegro)

1961-12-15

Task concerned with reprocessing of irradiated uranium covered the following activities: implementing the method and constructing the cell for uranium dissolving; implementing the procedure for extraction of uranium, plutonium and fission products from radioactive uranium solutions; studying the possibilities for using inorganic ion exchangers and adsorbers for separation of U, Pu and fission products.

Study of Irradiation Effect onto Uranium silicide Fuel

International Nuclear Information System (INIS)

Suparjo

1998-01-01

The irradiation effect onto the U 3 Si-Al and U 3 Si 2 -Al dispersion type of fuel element has been studied. The fuel material performs swelling during irradiation due to boehmite (Al 2 O 3 (H 2 O)) formation in which might occurs inside the meat and on the cladding surface, the interaction between the fuel and aluminium matrix that produce U(Al,Si) 3 phase, and the formation of fission gas bubble inside the fuel. At a constant fission density, the U 3 Si-Al fuel swelling is higher than that of U 3 Si 2 -Al fuel. The swellings of both fuels increase with the increasing of fission density. The difference of swelling behavior was caused by formation of large bubble gases generated from fission product of U 3 Si fuel and distributed non-uniformly over all of fuel zone. On the other hand, the U 3 Si 2 fission produced small bubble gases, and those were uniformly distributed. The growth rate of fission gas bubble in the U 3 Si fuel has shown high diffusivity, transformation into amorph material and thus decrease its mechanical strength

Irradiation testing of LEU fuels in the SILOE Reactor - Progress report

International Nuclear Information System (INIS)

Merchie, Francis; Baas, Claude; Martel, Patrick

1985-01-01

Irradiation of uranium-silicide fuels has continued in the SILOE reactor during the past year. Thickness vs. fission density data from four U 3 Si plates containing 5.5 and 6.0 g U/cm 3 have been analyzed, and the results are presented. The irradiation of a full 60 g U/cm 3 U 3 Si element has begun. In addition, four U 3 Si 2 plates containing 20 to 54 g U/cm 3 are now being irradiated. These irradiations and future plans are discussed in the paper. (author)

Neutronic Analysis and Radiological Safety of RSG-GAS Reactor on 300 Grams Uranium Silicide Core

International Nuclear Information System (INIS)

Pande Made Udiyani; Lily Suparlina; Rokhmadi

2007-01-01

As starting of usage silicide U 250 g fuel element in the core of RSG-GAS and will be continued with usage of silicide U 300 g fuel element, hence done beforehand neutronic analyse and radiological safety of RSG-GAS. Calculation done by ORIGEN2.1 code to calculate source term, and also by PC-COSYMA code to calculate radiological safety of radioactive dispersion from RSG-GAS. Calculation of radioactive dispersion done at condition of reactor is postulated be happened an accident of LOCA causing one fuel element to melt. Neutronic analysis indicate that silicide U 250 g full core shall to be operated beforehand during 625 MWD before converted to silicide U 300 g core. During operation of transition core with mixture of silicide U 250 g and 300 g, all parameter fulfill criterion of safety Designed Balance core of silicide U 300 g will be reached at the time of fifth full core. Result of calculation indicate that through mixture core of silicide U 250 and 300 g proposed can form silicide U 300 g balance core of reactor RSG-GAS safely. Calculation of radiology safety by deterministic for silicide U 300 g balance core, and accident postulation which is equal to core of silicide U 250 g yield output in the form of radiation activity (radionuclide concentration in the air and deposition on the ground), radiation dose (collective and individual), radiation effect (short- and long-range), which accepted by society in each perceived sector. Result of calculation indicated that dose accepted by society is not pass permitted boundary for public society if happened accident. (author)

The Accident Analysis Due to Reactivity Insertion of RSG GAS 3.55 g U/cc Silicide Core

International Nuclear Information System (INIS)

Endiah Puji-Hastuti; Surbakti, Tukiran

2004-01-01

The fuels of RSG-GAS reactor was changed from uranium oxide with 250 g U of loading or 2.96 g U/cc of fuel loading to uranium silicide with the same loading. The silicide fuels can be used in higher density, staying longer in the reactor core and hence having a longer cycle length. The silicide fuel in RSG-GAS core was made up in step-wise by using mixed up core Firstly, it was used silicide fuel with 250 g U of loading and then, silicide fuel with 300 g U of loading (3.55 g U/cc of fuel loading). In every step-wise of fuel loading, it must be analyzed its safety margin. In this occasion, the reactivity accident of RSG-GAS core with 300 g U of silicide fuel loading is analyzed. The calculation was done using EUREKA-2/RR code available at P2TRR. The calculation was done by reactivity insertion at start up and power rangers. The worst case accident is transient due to control rod with drawl failure at start up by means of lowest initial power (0.1 W), either in power range. From all cases which have been done, the results of analysis showed that there is no anomaly and safety margin break at RSG-GAS core with 300 g U silicide fuel loading. (author)

Simulated Fission Gas Behavior in Silicide Fuel at LWR Conditions

Energy Technology Data Exchange (ETDEWEB)

Miao, Yinbin [Argonne National Lab. (ANL), Argonne, IL (United States); Mo, Kun [Argonne National Lab. (ANL), Argonne, IL (United States); Yacout, Abdellatif [Argonne National Lab. (ANL), Argonne, IL (United States); Harp, Jason [Argonne National Lab. (ANL), Argonne, IL (United States)

2016-09-15

As a promising candidate for the accident tolerant fuel (ATF) used in light water reactors (LWRs), the fuel performance of uranium silicide (U₃Si₂) at LWR conditions needs to be well-understood. However, existing experimental post-irradiation examination (PIE) data are limited to the research reactor conditions, which involve lower fuel temperature compared to LWR conditions. This lack of appropriate experimental data significantly affects the development of fuel performance codes that can precisely predict the microstructure evolution and property degradation at LWR conditions, and therefore evaluate the qualification of U₃Si₂ as an AFT for LWRs. Considering the high cost, long timescale, and restrictive access of the in-pile irradiation experiments, this study aims to utilize ion irradiation to simulate the inpile behavior of the U₃Si₂ fuel. Both in situ TEM ion irradiation and ex situ high-energy ATLAS ion irradiation experiments were employed to simulate different types of microstructure modifications in U₃Si₂. Multiple PIE techniques were used or will be used to quantitatively analyze the microstructure evolution induced by ion irradiation so as to provide valuable reference for the development of fuel performance code prior to the availability of the in-pile irradiation data.

Effects of Silicide Coating on the Interdiffusion between U-7Mo and Al

Energy Technology Data Exchange (ETDEWEB)

Nam, Ji Min; Kim, Ji Hyun; Kim, Sunghwan; Lee, Kyu Hong; Park, Jong Man; Jeong, Yong Jin; Kim, Ki Nam [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-05-15

The excessive interaction between the U-Mo alloys and their surrounding Al matrix lead to and excessive local swelling called 'pillowing'. For this reason, KAERI suggested several remedies such as alloying U-Mo with Ti, or Al matrix with Si. In addition, silicide, or nitride coatings on the surface of U-Mo particles have also been proposed to hinder the growth of interaction layer. In this study, centrifugally atomized U-7Mo alloy powders were coated with silicide layers at varying T (T = 900 and 1000 .deg. C) for 30 min, respectively. U-Mo alloy powder was blended with Si powders and subsequently heat-treated to form uranium-silicide coating layers on the surface of U-Mo alloy particles. For an annealing test, silicide-coated U-Mo alloy powders were made into a compact, and Al powders were used as a matrix. From EDS results, transformed uranium aluminide intermetallic compounds were mainly U(Al,Si)3. U(Al,Si)3 phase left the silicide coating layer behind, and formed inside of U-7Mo particles, as shown in Fig. 3(a) and (b). In the case of sample B, Al could not penetrate the silicide coating layer and the coating layers were remained constant, as shown in Fig. 3(c) and (d). From the results, we made a comparison between the compacts of sample A and B, and it was shown that Al can easily diffuse into unreacted Si and U{sub 3}Si{sub 5} mixed layer while U{sub 3}Si{sub 2} acted as a good diffusion barrier at 550 .deg. C though those layers had the same thickness.

Effects of Silicide Coating on the Interdiffusion between U-7Mo and Al

International Nuclear Information System (INIS)

Nam, Ji Min; Kim, Ji Hyun; Kim, Sunghwan; Lee, Kyu Hong; Park, Jong Man; Jeong, Yong Jin; Kim, Ki Nam

2015-01-01

The excessive interaction between the U-Mo alloys and their surrounding Al matrix lead to and excessive local swelling called 'pillowing'. For this reason, KAERI suggested several remedies such as alloying U-Mo with Ti, or Al matrix with Si. In addition, silicide, or nitride coatings on the surface of U-Mo particles have also been proposed to hinder the growth of interaction layer. In this study, centrifugally atomized U-7Mo alloy powders were coated with silicide layers at varying T (T = 900 and 1000 .deg. C) for 30 min, respectively. U-Mo alloy powder was blended with Si powders and subsequently heat-treated to form uranium-silicide coating layers on the surface of U-Mo alloy particles. For an annealing test, silicide-coated U-Mo alloy powders were made into a compact, and Al powders were used as a matrix. From EDS results, transformed uranium aluminide intermetallic compounds were mainly U(Al,Si)3. U(Al,Si)3 phase left the silicide coating layer behind, and formed inside of U-7Mo particles, as shown in Fig. 3(a) and (b). In the case of sample B, Al could not penetrate the silicide coating layer and the coating layers were remained constant, as shown in Fig. 3(c) and (d). From the results, we made a comparison between the compacts of sample A and B, and it was shown that Al can easily diffuse into unreacted Si and U 3 Si 5 mixed layer while U 3 Si 2 acted as a good diffusion barrier at 550 .deg. C though those layers had the same thickness

Crystallographic study of the tempering by irradiation of cold-worked uranium (1960); Etude cristallographique du revenu de l'uranium ecroui par irradiation (1960)

Energy Technology Data Exchange (ETDEWEB)

Tardivon, D [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

1960-07-01

We have studied the phenomenon of the tempering of cold-worked uranium under the action of irradiation by observing the narrowing of the (114) and (133) X-ray diffraction lines as a function of the irradiation level. Simultaneously we have studied the broadening of the 114 line of a recrystallised uranium as a function of the irradiation level. The irradiation temperature was always less than 60 deg. C. Of these two processes, the first is the fastest. We have observed a saturation of the irradiation tempering for a flux of 10{sup 18} n/cm{sup 2}; we deduce from this the dimensions of the volume perturbed by one fission atom to be 10{sup -17} cm{sup 3}. (author) [French] Nous avons etudie le phenomene de revenu par irradiation d'echantillons d'uranium ecroui, en observant l'affinement des raies de diffraction de rayons X (114) et (133) en fonction du taux d'irradiation. Parallelement nous avons etudie l'elargissement de la raie (114 ) d'un uranium recristallise en fonction du taux d'irradiation. La temperature d'irradiation est toujours restee inferieure a 60 deg. C. De ces deux processus le premier est le plus rapide. Nous avons observe une saturation du revenu par irradiation pour un flux de 10{sup 18} n/cm{sup 2}; on en deduit une valeur du volume de la perturbation creee par un atome de fission egale a 10{sup -17} cm{sup 3}. (auteur)

Uranium briquettes for irradiation target

Energy Technology Data Exchange (ETDEWEB)

Saliba-Silva, Adonis Marcelo; Garcia, Rafael Henrique Lazzari; Martins, Ilson Carlos; Carvalho, Elita Fontenele Urano de; Durazzo, Michelangelo, E-mail: saliba@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2011-07-01

Direct irradiation on targets inside nuclear research or multiple purpose reactors is a common route to produce {sup 99}Mo-{sup 99m}Tc radioisotopes. Nevertheless, since the imposed limits to use LEU uranium to prevent nuclear armament production, the amount of uranium loaded in target meats has physically increased and new processes have been proposed for production. Routes using metallic uranium thin film and UAl{sub x} dispersion have been used for this purpose. Both routes have their own issues, either by bringing difficulties to disassemble the aluminum case inside hot cells or by generating great amount of alkaline radioactive liquid rejects. A potential route might be the dispersion of powders of LEU metallic uranium and nickel, which are pressed as a blend inside a die and followed by pulse electroplating of nickel. The electroplating provides more strength to the briquettes and creates a barrier for gas evolution during neutronic disintegration of {sup 235}U. A target briquette platted with nickel encapsulated in an aluminum case to be irradiated may be an alternative possibility to replace other proposed targets. This work uses pulse Ni-electroplating over iron powder briquette to simulate the covering of uranium by nickel. The following parameters were applied 10 times for each sample: 900Hz, -0.84A/square centimeters with duty cycle of 0.1 in Watts Bath. It also presented the optical microscopy analysis of plated microstructure section. (author)

Uranium briquettes for irradiation target

International Nuclear Information System (INIS)

Saliba-Silva, Adonis Marcelo; Garcia, Rafael Henrique Lazzari; Martins, Ilson Carlos; Carvalho, Elita Fontenele Urano de; Durazzo, Michelangelo

2011-01-01

Direct irradiation on targets inside nuclear research or multiple purpose reactors is a common route to produce 99 Mo- 99m Tc radioisotopes. Nevertheless, since the imposed limits to use LEU uranium to prevent nuclear armament production, the amount of uranium loaded in target meats has physically increased and new processes have been proposed for production. Routes using metallic uranium thin film and UAl x dispersion have been used for this purpose. Both routes have their own issues, either by bringing difficulties to disassemble the aluminum case inside hot cells or by generating great amount of alkaline radioactive liquid rejects. A potential route might be the dispersion of powders of LEU metallic uranium and nickel, which are pressed as a blend inside a die and followed by pulse electroplating of nickel. The electroplating provides more strength to the briquettes and creates a barrier for gas evolution during neutronic disintegration of 235 U. A target briquette platted with nickel encapsulated in an aluminum case to be irradiated may be an alternative possibility to replace other proposed targets. This work uses pulse Ni-electroplating over iron powder briquette to simulate the covering of uranium by nickel. The following parameters were applied 10 times for each sample: 900Hz, -0.84A/square centimeters with duty cycle of 0.1 in Watts Bath. It also presented the optical microscopy analysis of plated microstructure section. (author)

Enhancement of uranium-accumulating ability of microorganisms by irradiation

Energy Technology Data Exchange (ETDEWEB)

Sakaguchi, Takashi; Nakajima, Akira; Tsuruta, Takehiko [Miyazaki Medical Coll., Kiyotake (Japan)

1998-01-01

Some microorganisms having excellent ability to accumulate uranium were isolated, from soil and water systems in and around the Ningyo-toge Station of Power Reactor and Nuclear Fuel Development Corporation. The enhancement of uranium-accumulating ability of microorganisms by electron-beam irradiation was examined, and the ability of JW-046 was increased 3-5% by the irradiation. The irradiation affect the growth of some of microorganisms tested. (author)

Recovery of uranium from an irradiated solid target after removal of molybdenum-99 produced from the irradiated target

Science.gov (United States)

Reilly, Sean Douglas; May, Iain; Copping, Roy; Dale, Gregory Edward

2017-10-17

A process for minimizing waste and maximizing utilization of uranium involves recovering uranium from an irradiated solid target after separating the medical isotope product, molybdenum-99, produced from the irradiated target. The process includes irradiating a solid target comprising uranium to produce fission products comprising molybdenum-99, and thereafter dissolving the target and conditioning the solution to prepare an aqueous nitric acid solution containing irradiated uranium. The acidic solution is then contacted with a solid sorbent whereby molybdenum-99 remains adsorbed to the sorbent for subsequent recovery. The uranium passes through the sorbent. The concentrations of acid and uranium are then adjusted to concentrations suitable for crystallization of uranyl nitrate hydrates. After inducing the crystallization, the uranyl nitrate hydrates are separated from a supernatant. The process results in the purification of uranyl nitrate hydrates from fission products and other contaminants. The uranium is therefore available for reuse, storage, or disposal.

Radiation Re-solution Calculation in Uranium-Silicide Fuels

International Nuclear Information System (INIS)

Matthews, Christopher; Andersson, Anders David Ragnar; Unal, Cetin

2017-01-01

The release of fission gas from nuclear fuels is of primary concern for safe operation of nuclear power plants. Although the production of fission gas atoms can be easily calculated from the fission rate in the fuel and the average yield of fission gas, the actual diffusion, behavior, and ultimate escape of fission gas from nuclear fuel depends on many other variables. As fission gas diffuses through the fuel grain, it tends to collect into intra-granular bubbles, as portrayed in Figure 1.1. These bubbles continue to grow due to absorption of single gas atoms. Simultaneously, passing fission fragments can cause collisions in the bubble that result in gas atoms being knocked back into the grain. This so called ''re-solution'' event results in a transient equilibrium of single gas atoms within the grain. As single gas atoms progress through the grain, they will eventually collect along grain boundaries, creating inter-granular bubbles. As the inter-granular bubbles grow over time, they will interconnect with other grain-face bubbles until a pathway is created to the outside of the fuel surface, at which point the highly pressurized inter-granular bubbles will expel their contents into the fuel plenum. This last process is the primary cause of fission gas release. From the simple description above, it is clear there are several parameters that ultimately affect fission gas release, including the diffusivity of single gas atoms, the absorption and knockout rate of single gas atoms in intra-granular bubbles, and the growth and interlinkage of intergranular bubbles. Of these, the knockout, or re-solution rate has an particularly important role in determining the transient concentration of single gas atoms in the grain. The re-solution rate will be explored in the following sections with regards to uranium-silicide fuels in order to support future models of fission gas bubble behavior.

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

Science.gov (United States)

Travelli, Armando

1988-01-01

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

Burn-up analysis of uranium silicide fuels 20% 235U, in the LFR facility

International Nuclear Information System (INIS)

Amor, Ricardo A.; Bouza, Edgardo; Cabrejas, Julian L.; Devida, Claudio A.; Gil, Daniel A.; Stankevicius, Alejandro; Gautier, Eduardo; Garavaglia, Ricardo N.; Lobo, Alfredo

2003-01-01

The LFR Facility is a laboratory designed and constructed with a Hot-Cells line, a Globe-Box and a Fume-Hood, all of them suited to work with radioactive materials such as samples of irradiated silicide MTR fuel elements. A series of dissolutions of this material was performed. From the resulting solutions, two fractions were separated by HPLC. One contained U + Pu, and other the fission product Nd. The concentrations of these elements were obtained by isotopic dilution and mass spectrometry (IDMS). It is concluded that this technique is very powerful and accurate when properly applied, and makes the validation of burn-up calculation codes possible. It is worth remarking the Lfr capacity to carry on different Research and Development (R + D) tasks in the Nuclear Fuel Cycle field. (author)

Neutronic design of mixed oxide-silicide cores for the core conversion of rsg-gas reactor

International Nuclear Information System (INIS)

Sembiring, Tagor Malem; Tukiran; Pinem surian; Febrianto

2001-01-01

The core conversion of rsg-gas reactor from an all-oxide (U 3 O 8 -Al) core, through a series of mixed oxide-silicide core, to an all-silicide (U 3 Si 2 -Al) core for the same meat density of 2.96 g U/cc is in progress. The conversion is first step of the step-wise conversion and will be followed by the second step that is the core conversion from low meat density of silicide core, through a series of mixed lower-higher density of silicide core, to an all-higher meat density of 3.55 g/cc core. Therefore, the objectives of this work is to design the mixed cores on the neutronic performance to achieve safety a first full-silicide core for the reactor with the low uranium meat density of 2.96gU/cc. The neutronic design of the mixed cores was performed by means of Batan-EQUIL-2D and Batan-3DIFF computer codes for 2 and 3 dimension diffusion calculation, respectively. The result shows that all mixed oxide-silicide cores will be feasible to achieve safety a fist full-silicide core. The core performs the same neutronic core parameters as those of the equilibrium silicide core. Therefore, the reactor availability and utilization during the core conversion is not changed

The dangers of irradiate uranium in nuclear reactors; Les dangers de l'uranium irradie dans les reacteurs nucleaires

Energy Technology Data Exchange (ETDEWEB)

Jammet, H; Joffre, H [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1955-07-01

The danger of the uranium cans sur-activated by the use in the nuclear reactors is triple: - Irradiation from afar, during manipulations of the cans. - Contamination of air when decladding. - Contamination of air by fire of uranium in a reactor in function The first two dangers are usual and can be treated thanks to the rules of security in use in the atomic industry. The third has an accidental character and claimed for the use of special and exceptional rules, overflowing the industrial setting, to reach the surrounding populations. (authors) [French] Le danger des cartouches d'uranium suractive par utilisation dans les reacteurs nucleaires est triple: - Irradiation a distance, lors des manipulations des cartouches. - Contamination de l'air au moment de leur degainage. - Contamination de l'air par incendie d'uranium dans un reacteur en fonctionnement. Les deux premiers dangers sont habituels et peuvent etre traites grace aux regles de securite en usage dans l'industrie atomique. Le troisieme revet un caractere accidentel et reclame l'emploi de regles speciales et exceptionnelles, debordant le cadre industriel, pour atteindre celui des populations environnantes. (auteurs)

The Comparison Of Silicon Analysis For The Uranium Silicide Fuel Using Spectrophotometrical And Gravimetrical Methods

International Nuclear Information System (INIS)

Putro, P. K.; Suripto, A.; Putra, S.; Gunanjar

1996-01-01

The analysis of silicon content in the uranium silicide fuel spectro-photometrical and gravimetrical method have been performed. The nitrous oxide-acetylene was used in the atomic absorption spectrophotometry (AAS) on the wave length of 251.6 nm, and the mixture of ammonium hepta molybdate complexes and SnC1 2 as reductor were applied during analysis by UV-VIS spectrophotometry (UV-VIS) on the wave length of 757.5 mm. The reagent of HCLO 4 and HNO 3 were used for determining Si content by gravimetrical methods. The results of this comparison is as follows: the accuracy result is around 96.37 % + 0.24 % for the Si concentration up to 300 ppm (the AAS), is 138.60 % = 0.43 % for the Si concentration range between 0.1-1.5 ppm (UV-VIS), and is 51.13 % + 0.8 % for 1 gram of Si (gravimetry). The results also show that the lowest analytical error is obtained by AAS method

Potentiometric determination of hexavalent uranium in uranium silicide samples

International Nuclear Information System (INIS)

Arlegui, Oscar

1999-01-01

The Chilean Nuclear Energy Commission's Department of Nuclear Materials has among its projects the production of fuels elements for nuclear reactors, and, therefore, the Chemical Analysis Laboratory must have a rapid and reliable method for uranium analysis, to control the uranium concentration during each stage of the production process. For this reason the Chilean Nuclear Energy Commission's Chemical Analysis Laboratory has validated a potentiometric method, which is a modification of the Davies and Gray method proposed by A.R. Eberle. This method uses the Potentiometric Titration Technique and is based on the direct and rapid reduction of uranium (VI) to Uranium (IV), in a concentrated phosphoric acid medium, with excess iron (II) used as a reducing agent. In this medium the excess iron (II) selectively oxidizes to iron (III) with nitric acid, using molybdenum (IV) as a catalyzer, the nitrous acid that is produced is eliminated by adding amidosulfuric acid. The solution is diluted with 1M sulfuric acid and the uranium (IV) obtained is titrated potentiometrically with potassium dichromate in the presence of vanadilic sulfate to obtain a better defined final titration point. The samples were softened with hydrochloric acid and nitric acid and later 50 ml were estimated in a 20% sulfuric acid medium. The analytical method was validated by comparing it with Certified Reference Material (C.R.M.) from the New Brunswick Laboratory (NBL), Metallic Uranium, CRM 112-A. The F Test and the T Test show that the value calculated is less than the tabulated value so the result is traceable to the reference material. The quantification limit, sensitivity, precision and accuracy were quantified for the method

Fuel-cycle cost comparisons with oxide and silicide fuels

International Nuclear Information System (INIS)

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

1982-01-01

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

Refractory silicides for integrated circuits

International Nuclear Information System (INIS)

Murarka, S.P.

1980-01-01

Transition metal silicides have, in the past, attracted attention because of their usefulness as high temperature materials and in integrated circuits as Schottky barrier and ohmic contacts. More recently, with the increasing silicon integrated circuits (SIC) packing density, the line widths get narrower and the sheet resistance contribution to the RC delay increases. The possibility of using low resistivity silicides, which can be formed directly on the polysilicon, makes these silicides highly attractive. The usefulness of a silicide metallization scheme for integrated circuits depends, not only on the desired low resistivity, but also on the ease with which the silicide can be formed and patterned and on the stability of the silicides throughout device processing and during actual device usage. In this paper, various properties and the formation techniques of the silicides have been reviewed. Correlations between the various properties and the metal or silicide electronic or crystallographic structure have been made to predict the more useful silicides for SIC applications. Special reference to the silicide resistivity, stress, and oxidizability during the formation and subsequent processing has been given. Various formation and etching techniques are discussed

Irradiation Stability of Uranium Alloys at High Exposures

International Nuclear Information System (INIS)

McDonell, W.R.

2001-01-01

Postirradiation examinations were begun of a series of unrestrained dilute uranium alloy specimens irradiated to exposures up to 13,000 MWD/T in NaK-containing stainless steel capsules. This test, part of a program of development of uranium metal fuels for desalination and power reactors sponsored by the Division of Reactor Development and Technology, has the objective of defining the temperature and exposure limits of swelling resistance of the alloyed uranium. This paper discusses those test results

In situ atomic-level observation of the formation of platinum silicide at platinum-silicon oxide interfaces under electron irradiation

Directory of Open Access Journals (Sweden)

Takeshi Nagase

2018-05-01

Full Text Available In situ atomic-level observation of the formation of Pt2Si at Pt/SiOx interface by electronic excitation under electron irradiation was performed by using scanning transmission electron microscopy. Scanning of an electron-beam probe stimulates silicide formation at the Pt/SiOx interface; the change in the Pt column corresponding to Pt2Si formation with a crystallographic orientation of (001Pt//(001Pt2Si and [110]Pt//[110]Pt2Si was observed in high-angle annular dark-field images.

Preliminary study of the preparation of uranium 232 by irradiation of protactinium 231; Etude preliminaire a la preparation d'uranium 232 par irradiation de protactinium 231

Energy Technology Data Exchange (ETDEWEB)

Guillot, Ph. [Commissariat a l' Energie Atomique, Fontenay aux Roses (France). Centre d' Etudes Nucleaires

1965-07-01

A bibliography about preparation of uranium 232 is done. This even-even isotope of uranium is suitable for radioactive tracer, neutron source through {alpha},n reaction and heat source applications. The irradiation of protactinium 231, the chemical separation and the purification of uranium are studied. (author) [French] Une etude bibliographique de la preparation d'uranium 232 a ete effectuee. Cet isotope pair-pair de l'uranium peut etre utilise en tant que traceur, source d'energie et source de neutrons, lorsqu'il est melange a un element leger tel le beryllium. Une etude du taux de formation des isotopes produits, lors de l'irradiation du protactinium 231 - une des manieres d'obtenir l'uranium 232 - a ete faite a l'aide d'un programme passe sur ordinateur. Les problemes poses par la separation chimique et la purification de l'uranium ont ete egalement envisages dans ce rapport. (auteur)

Irradiation tests of 99Mo isotope production targets employing uranium metal foils

International Nuclear Information System (INIS)

Hofman, G.L.; Wiencek, T.C.; Wood, E.L.; Snelgrove, J.L.; Suripto, A.; Nasution, H.; Lufti-Amin, D.; Gogo, A.

1996-01-01

Most of the world's supply of 99 mTc for medical purposes is currently produced from the decay of 99 Mo derived from the fissioning of high-enriched uranium (HEU). Substitution of low-enriched uranium (LEU) metal foils for the HEU UO 2 used in current target designs will allow equivalent 99 Mo yields with little change in target geometries. Substitution of uranium metal for uranium alloy and aluminide in other target designs will also allow the conversion of HEU to LEU. Several uranium-metal-foil targets have been fabricated at ANL and irradiated to prototypic burnup in the Indonesian RSG-GAS reactor. Postirradiation examination of the initial test indicated that design modifications were required to allow the irradiated foil to be removed for chemical processing. The latest test has shown good irradiation behavior, satisfactory dismantling and foil removal when the U-foil is separated from its containment by metallic, fission-recoil absorbing barriers. (author)

Irradiation tests of 99Mo isotope production targets employing uranium metal foils

International Nuclear Information System (INIS)

Hofman, G.L.; Wiencek, T.C.; Wood, E.L.; Snelgrove, J.L.; Suripto, A.; Nasution, H.; Lufti-Amin, D.; Gogo, A.

1996-01-01

Most of the world's supply of 99m Tc for medical purposes is currently produced form the decay of 99 Mo derived from the fissioning of high-enriched uranium (HEU). Substitution of low-enriched uranium (LEU) metal foils for the HEU UO 2 used in current target designs will allow equivalent 99 Mo yields with little change in target geometries. Substitution of uranium metal for uranium alloy and aluminide in other target designs will also allow the conversion of HEU to LEU. Several uranium-metal-foil targets have been fabricated at ANL and irradiated to prototypic burnup in the Indonesian RSG-GAS reactor. Postirradiation examination of the initial test indicated that design modifications were required to allow the irradiated foil to be removed for chemical processing. The latest test has shown good irradiation behavior, satisfactory dismantling and foil removal when the U-foil is separated from its containment by metallic, fission-recoil absorbing barriers

Uranium doping and neutron irradiation of Bi-2223 superconduction tapes for improved critical current density

International Nuclear Information System (INIS)

Moss, S.D.; Wang, W.G.; Dou, S.X.; Weinstein, R.

1998-01-01

It is demonstrated that a combination of neutron irradiation with uranium doping introduce fission tracks through a Bi-2223 tape which act as effective pinning centres, leading to a substantial increase in critical current. Preliminary data suggests that the combination of uranium doping and neutron irradiation produces improved flux pinning in Bi-2223 tapes over neutron irradiation alone. Before irradiation, SEM, DTA and XRD analyses were performed on the tapes. Both before and after irradiation the trapped maximum magnetic flux was measured at 77K. Before neutron irradiation, uranium doping has no effect on critical current. Preliminary SEM data suggested that the uranium is homogeneously distributed throughout the oxide core of the tape. The presence of 2212 and other secondary phases in the doped tapes suggest further refinement of the sintering procedure is necessary. (authors)

Near surface silicide formation after off-normal Fe-implantation of Si(001) surfaces

Energy Technology Data Exchange (ETDEWEB)

Khanbabaee, B., E-mail: khanbabaee@physik.uni-siegen.de; Pietsch, U. [Solid State Physics, University of Siegen, D-57068 Siegen (Germany); Lützenkirchen-Hecht, D. [Fachbereich C - Physik, Bergische Universität Wuppertal, D-42097 Wuppertal (Germany); Hübner, R.; Grenzer, J.; Facsko, S. [Helmholtz-Zentrum Dresden-Rossendorf, 01314 Dresden (Germany)

2014-07-14

We report on formation of non-crystalline Fe-silicides of various stoichiometries below the amorphized surface of crystalline Si(001) after irradiation with 5 keV Fe{sup +} ions under off-normal incidence. We examined samples prepared with ion fluences of 0.1 × 10{sup 17} and 5 × 10{sup 17} ions cm{sup −2} exhibiting a flat and patterned surface morphology, respectively. Whereas the iron silicides are found across the whole surface of the flat sample, they are concentrated at the top of ridges at the rippled surface. A depth resolved analysis of the chemical states of Si and Fe atoms in the near surface region was performed by combining X-ray photoelectron spectroscopy and X-ray absorption spectroscopy (XAS) using synchrotron radiation. The chemical shift and the line shape of the Si 2p core levels and valence bands were measured and associated with the formation of silicide bonds of different stoichiometric composition changing from an Fe-rich silicides (Fe{sub 3}Si) close to the surface into a Si-rich silicide (FeSi{sub 2}) towards the inner interface to the Si(001) substrate. This finding is supported by XAS analysis at the Fe K-edge which shows changes of the chemical environment and the near order atomic coordination of the Fe atoms in the region close to surface. Because a similar Fe depth profile has been found for samples co-sputtered with Fe during Kr{sup +} ion irradiation, our results suggest the importance of chemically bonded Fe in the surface region for the process of ripple formation.

Fabrication and irradiation testing of LEU [low enriched uranium] fuels at CRNL status as of 1987 September

International Nuclear Information System (INIS)

Sears, D.F.; Berthiaume, L.C.; Herbert, L.N.

1987-01-01

The current status of Chalk River Nuclear Laboratories' (CRNL) program to develop and test low-enriched uranium (LEU), proliferation-resistant fuels for use in research reactors is reviewed. CRNL's fuel manufacturing process has been qualified by the successful demonstration irradiation of 7 full-size rods in the NRU reactor. Now industrial-scale production equipment has been commissioned, and a fuel-fabrication campaign for 30 NRU rods and a MAPLE-X core is underway. Excess capacity could be used for commercial fuel fabrication. In the irradiation testing program, mini-elements with deliberately included core surface defects performed well in-reactor, swelling by only 7 to 8 vol% at 93 atomic percent burnup of the original U-235. The additional restraint provided by the aluminium cladding which flowed into the defects during extrusion contributed to this good performance. Mini-elements containing a variety of particle size distributions were also successfully irradiated to 93 at% burnup in NRU, as part of a study to establish the optimum particle size distribution. Swelling was found to be proportional to the percentage of fines (<44μm particles) contained in the cores. The mini-elements containing the composition normally used at CRNL had swollen by 5.8 vol%, and mini-elements with a much higher percentage of fines had swollen by 6.8 vol%, at 93 at% burnup. Also, a program to develop LEU targets for Mo-99 production, via the technology developed to fabricate dispersed silicide fuel, has started, and preliminary scoping studies are underway. (Author)

Manhattan Project Technical Series The Chemistry of Uranium (I) Chapters 1-10

International Nuclear Information System (INIS)

Rabinowitch, E. I.; Katz, J. J.

1946-01-01

This constitutes Chapters 1 through 10. inclusive, of The Survey Volume on Uranium Chemistry prepared for the Manhattan Project Technical Series. Chapters are titled: Nuclear Properties of Uranium; Properties of the Uranium Atom; Uranium in Nature; Extraction of Uranium from Ores and Preparation of Uranium Metal; Physical Properties of Uranium Metal; Chemical Properties of Uranium Metal; Intermetallic Compounds and Alloy systems of Uranium; the Uranium-Hydrogen System; Uranium Borides, Carbides, and Silicides; Uranium Nitrides, Phosphides, Arsenides, and Antimonides.

Fuel loading and homogeneity analysis of HFIR design fuel plates loaded with uranium silicide fuel

International Nuclear Information System (INIS)

Blumenfeld, P.E.

1995-08-01

Twelve nuclear reactor fuel plates were analyzed for fuel loading and fuel loading homogeneity by measuring the attenuation of a collimated X-ray beam as it passed through the plates. The plates were identical to those used by the High Flux Isotope Reactor (HFIR) but were loaded with uranium silicide rather than with HFIR's uranium oxide fuel. Systematic deviations from nominal fuel loading were observed as higher loading near the center of the plates and underloading near the radial edges. These deviations were within those allowed by HFIR specifications. The report begins with a brief background on the thermal-hydraulic uncertainty analysis for the Advanced Neutron Source (ANS) Reactor that motivated a statistical description of fuel loading and homogeneity. The body of the report addresses the homogeneity measurement techniques employed, the numerical correction required to account for a difference in fuel types, and the statistical analysis of the resulting data. This statistical analysis pertains to local variation in fuel loading, as well as to ''hot segment'' analysis of narrow axial regions along the plate and ''hot streak'' analysis, the cumulative effect of hot segment loading variation. The data for all twelve plates were compiled and divided into 20 regions for analysis, with each region represented by a mean and a standard deviation to report percent deviation from nominal fuel loading. The central regions of the plates showed mean values of about +3% deviation, while the edge regions showed mean values of about -7% deviation. The data within these regions roughly approximated random samplings from normal distributions, although the chi-square (χ 2 ) test for goodness of fit to normal distributions was not satisfied

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

International Nuclear Information System (INIS)

Nazare, S.

1984-01-01

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

High pressure studies on uranium and thorium silicide compounds: Experiment and theory

DEFF Research Database (Denmark)

Yagoubi, S.; Heathman, S.; Svane, A.

2013-01-01

The actinide silicides ThSi, USi and USi2 have been studied under high pressure using both theory and experiment. High pressure synchrotron X-ray diffraction experiments were performed on polycrystalline samples in diamond anvil cells at room temperature and for pressures up to 54, 52 and 26 GPa...

Irradiation performance of uranium-molybdenum alloy dispersion fuels

International Nuclear Information System (INIS)

Almeida, Cirila Tacconi de

2005-01-01

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

Determination of irradiated uranium in far-field contaminated areas of Belarus

International Nuclear Information System (INIS)

Mironov, V.; Pribylev, S.; Hotchkis, M.; Child, D.

2006-01-01

The possibility of using U 236 as an indicator for irradiated uranium is shown. The sensitivity of AMS is high enough for measurements of 236 U/ 238 U ratios down to 10 -9 on micrograms of uranium and therefore for the detection of Chernobyl originated uranium in the remote regions of radioactive fallout. (authors)

Steady-state thermal hydraulic analysis and flow channel blockage accident analysis of JRR-3 silicide core

International Nuclear Information System (INIS)

Kaminaga, Masanori

1997-03-01

JRR-3 is a light water moderated and cooled, beryllium and heavy water reflected pool type research reactor using low enriched uranium (LEU) plate-type fuels. Its thermal power is 20 MW. The core conversion program from uranium-aluminum (UAl x -Al) dispersion type fuel (aluminide fuel) to uranium-silicon-aluminum (U 3 Si 2 -Al) dispersion type fuel (silicide fuel) is currently conducted at the JRR-3. This report describes about the steady-state thermal hydraulic analysis results and the flow channel blockage accident analysis result. In JRR-3, there are two operation mode. One is high power operation mode up to 20 MW, under forced convection cooling using the primary and the secondary cooling systems. The other is low power operation mode up to 200 kW, under natural circulation cooling between the reactor core and the reactor pool without the primary and the secondary cooling systems. For the analysis of the flow channel blockage accident, COOLOD code was used. On the other hand, steady-state thermal hydraulic analysis for both of the high power operation mode under forced convection cooling and low power operation under natural convection cooling, COOLOD-N2 code was used. From steady-state thermal hydraulic analysis results of both forced and natural convection cooling, fuel temperature, minimum DNBR etc. meet the design criteria and JRR-3 LEU silicide core has enough safety margin under normal operation conditions. Furthermore, flow channel blockage accident analysis results show that one channel flow blockage accident meet the safety criteria for accident conditions which have been established for JRR-3 LEU silicide core. (author)

Irradiation effects of the zirconium oxidation and the uranium diffusion in zirconia; Effets d'irradiation sur l'oxydation du zirconium et la diffusion de l'uranium dans la zircone

Energy Technology Data Exchange (ETDEWEB)

Bererd, N

2003-07-01

The context of this study is the direct storage of spent fuel assemblies after operation in reactor. In order to obtain data on the capacities of the can as the uranium diffusion barrier, a fundamental study has been carried out for modelling the internal cladding surface under and without irradiation. The behaviour of zirconium in reactor conditions has at first been studied. A thin uranium target enriched with fissile isotope has been put on a zirconium sample, the set being irradiated by a thermal neutrons flux leading to the fission of the deposited uranium. The energetic history of the formed fission products has revealed two steps: 1)the zirconium oxidation and 2)the diffusion of uranium in the zirconia formed at 480 degrees C. A diffusion coefficient under irradiation has been measured. Its value is 10{sup -15} cm{sup 2}.s{sup -1}. In order to be able to reveal clearly the effect of the irradiation by the fission products on the zirconium oxidation, measurements of thermal oxidation and under {sup 129}Xe irradiation have been carried out. They have shown that the oxidation is strongly accelerated by the irradiation and that the temperature is negligible until 480 degrees C. On the other hand, the thermal diffusion of the uranium in zirconium and in zirconia has been studied by coupling ion implantation and Rutherford backscattering spectroscopy. This study shows that the uranium diffuses in zirconium and is trapped in zirconia in a UO{sub 3} shape. (O.M.)

SHI induced irradiation effect on Mo/Si interface

International Nuclear Information System (INIS)

Agarwal, Garima; Agarwal, Shivani; Jain, Rajkumar; Lal, Chhagan; Jain, I.P.; Kabiraj, D.; Pandey, Akhilesh

2006-01-01

Present parametric study investigates the characteristics of SHI induced mixed molybdenum silicide film with various ion fluences. The deposition of molybdenum thin films onto the Silicon substrate was performed using e-beam evaporation, while the heavy Au ion irradiation with energy 120 MeV was subsequently applied to form molybdenum silicide. The samples have been characterized by grazing incidence X-ray diffraction (GIXRD) for the identification of phase formation at the interface. Formation of t-Mo 5 Si 3 mixed molybdenum silicide was observed on increasing the ion irradiation fluences. (author)

A method for the quantitative determination of uranium-233 in an irradiated thorium rod; Une methode de dosage de l'uranium 233 contenu dans un barreau de thorium irradie

Energy Technology Data Exchange (ETDEWEB)

Bathellier, A; Sontag, R; Chesne, A [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

1961-07-01

A rapid method for the quantitative determination of uranium-233 in irradiated thorium is described. A 30 per cent solution of trilaurylamine in xylene is used to extract the uranium from an aqueous hydrochloric acid solution and separate it from the thorium. This may be followed by {alpha} counting or fluorimetry. The practical operating conditions of the separation are discussed in detail. (author) [French] Une methode rapide de dosage de l'uranium-233 contenu dans le thorium irradie est decrite. Elle utilise la trilauryfamine a 30 pour cent dans le xylene pour extraire l'uranium d'une dissolution aqueuse chlorhydrique et le separer du thorium. Le comptage {alpha} ou la fluorimetrie sont alors possibles. Les conditions operatoires de la separation sont discutees et precisees. (auteur)

Neutronic calculations of PARR-1 cores using LEU silicide fuel

International Nuclear Information System (INIS)

Arshad, M.; Bakhtyar, S.; Hayat, T.; Salahuddin, A.

1991-08-01

Detailed neutronic calculations have been carried out for different PARR-1 cores utilizing low enriched uranium (LEU) silicide fuel and operating at an upgraded power of 9 MW. The calculations include the search for critical loadings in open and stall ends of the pool, neutronic analysis of the first full equilibrium core and calculations cores. The burnup study of inventory have also been carried out. Further, the reactivity coefficients of the first full power operation core are evaluated for use in the accident analysis. 14 figs. (author)

Tuning up and fabrication of U3Si2 nuclear material

International Nuclear Information System (INIS)

Pasqualini, Enrique E.; Echenique, Patricia N.; Rossi, Gustavo S.; Canil, Eduardo E.; Esteban, Adolfo; Lopez, Marisol; Adelfang, Pablo

2000-01-01

This work describes the tuning up and fabrication of uranium-silicide powder for its utilization as nuclear fuel in material testing reactors taking in account NUREG-1313 recommendations, the experience of several suppliers and the one acquired in this work.Several alloy compositions were melted with natural uranium at temperatures of about 1800 degree C for adjusting composition and ingot homogeneity. Alumina, magnesia and zirconia-5% stabilized yttria crucibles were tested to evaluate the degree of contamination introduced by chemical attack of molten uranium and silicon. The fabrication procedure of 20% enriched uranium-silicide powder was established for building up the P-06 fuel element that actually is being irradiated at the RA-3 reactor facility. The selected procedures of fabrication and the critical analysis for the interpretation of several specifications are discussed. Results are shown of the obtained ingots and powder produced with the enriched uranium-silicide. (author)

Fuel development at CERCA. Status of development - September 1984

International Nuclear Information System (INIS)

Fanjas, Y.; Dewez, Ph.; Savornin, B.

1985-01-01

Since 1978, CERCA has developed high density aluminide (UAl x ), oxide (U 3 O 8 ) and silicides (U 3 Si 2 , U 3 Si) fuels allowing the use of 19.75 enriched uranium in research and test reactors. An extensive irradiation program has been carried out to test the full size fuel plates and fuel elements fabricated by CERCA. So far, all the irradiation tests have given satisfactory results whatever the uranium density, the burn-up level and the type of fuel. In particular, silicides which cover the whole density range from 1 to 7 g U/cm 3 appear more and more as the standard fuels for the future. (author)

Preliminary study of the preparation of uranium 232 by irradiation of protactinium 231

International Nuclear Information System (INIS)

Guillot, Ph.

1965-01-01

A bibliography about preparation of uranium 232 is done. This even-even isotope of uranium is suitable for radioactive tracer, neutron source through α,n reaction and heat source applications. The irradiation of protactinium 231, the chemical separation and the purification of uranium are studied. (author) [fr

Radiation damage in uranium under electron irradiation of energies up to 20 MeV

International Nuclear Information System (INIS)

Emets, N.L.; Zelenskij, V.F.; Kuz'menko, V.A.; Ranyuk, Yu.N.; Reznichenko, Eh.A.; Shilyaev, B.A.; Yamnitskij, V.A.

1980-01-01

The problem of conservation of primary radiation-induced defects in uranium irradiated by electrons with the energy exceeding photo fission threshold is considered. Calculation of uranium burnout is carried out. Calculations are conducted by the method of mathematical simulation, using some nuclear models; development of electromagnetic cascade in uranium, photofission process, elastic and inelastic electron scattering, as well as some secondary processes are taken into account. Proved is the fact of anomalous growth of uranium under electron irradiation, registered earlier experimentally. It is shown, that in case of acquiring the value Ed=15 eV radiation uranium growth at low levels of burnout can be explained by the complete capture of all the primary radiationn-induced defects into dislocation loops [ru

Development of the uranium recovery process from rejected fuel plates in the fabrication of MTR type nuclear fuel

International Nuclear Information System (INIS)

Fleming Rubio, Peter Alex

2010-01-01

The current work was made in Conversion laboratory belonging to Chilean Nuclear Energy Commission, CCHEN. This is constituted by the development of three hydrometallurgical processes, belonging to the recovery of uranium from fuel plates based on uranium silicide (U_3Si_2) process, for nuclear research reactors MTR (Material Testing Reactor) type, those that come from the Fuel Elements Manufacture Plant, PEC. In the manufacturing process some of these plates are subjected to destructive tests by quality requirement or others are rejected for non-compliance with technical specifications, such as: lack of homogenization of the dispersion of uraniferous compound in the meat, as well as the appearance of the defects, such as blisters, so-called "dog bone", "fish tail", "remote islands", among others. Because the uranium used is enriched in 19.75% U_2_3_5 isotope, which explains the high value in the market, it must be recovered for reuse, returning to the production line of fuel elements. The uranium silicide, contained in the plates, is dispersed in an aluminum matrix and covered with plates and frames of ASTM 6061 Aluminum, as a sandwich coating, commonly referred to as 'meat' (sandwich meat). As aluminum is the main impurity, the process begins with this metal dissolution, present in meat and plates, by NaOH reaction, followed by a vacuum filtration, washing and drying, obtaining a powder of uranium silicide, with a small impurities percentage. Then, the crude uranium silicide reacts with a solution of hydrofluoric acid, dissolving the silicon and simultaneously precipitating UF_4 by reaction with HNO_3, obtaining an impure UO_2(NO_3)_2 solution. The experimental work was developed and implemented at laboratory scale for the three stages pertaining to the uranium recovery process, determining for each one the optimum operation conditions: temperature, molarity or concentration, reagent excess, among others (author)

Formation of uranium based nanoparticles via gamma-irradiation

Energy Technology Data Exchange (ETDEWEB)

Nenoff, Tina M., E-mail: tmnenof@sandia.gov [Nanoscale Sciences Department, Sandia National Laboratories, P.O. Box 5800, MS-1415, Albuquerque, NM 87185 (United States); Ferriera, Summer R. [Nanoscale Sciences Department, Sandia National Laboratories, P.O. Box 5800, MS-1415, Albuquerque, NM 87185 (United States); Huang, Jianyu [Center for Integrated Nanotechnology, Sandia National Laboratories, P.O. Box 5800, MS-1315, Albuquerque, NM 87185 (United States); Hanson, Donald J. [Department of Hot Cells and Gamma Facilities, Sandia National Laboratories, P.O. Box 5800, MS-1143, Albuquerque, NM 87185 (United States)

2013-11-15

Graphical abstract: TEM image of d-U nanoparticles formed in aqueous solution by gamma irradiation. Display Omitted -- Highlights: •d-U nanoparticles were grown in solution by gamma irradiation. •The reaction solution does not exceed 25 °C (room temperature). •Only after multiday exposure to air is there evidence of oxidation of the d-U nanoparticles. •Evidence of d-U alloy nanoparticle formation confirmed by TEM/energy-dispersive X-ray (EDS) analysis. -- Abstract: The ability to fabricate nuclear fuels at low temperatures allows for the production of complex Uranium metal and alloys with minimum volatility of alloy components in the process. Gamma irradiation is a valuable method for the synthesis of a wide range of metal-based nanoparticles. We report on the synthesis via room temperature radiolysis and characterization of uranium (depleted, d-U) metal and uranium–lathanide (d-ULn, Ln = lanthanide surrogates) alloy nanoparticles from aqueous acidic salt solutions. The lanthanide surrogates chosen include La and Eu due to their similarity in ionic size and charge in solution. Detailed characterization results including UV–vis, TEM/HR-TEM, and single particle EDX (elemental analyses) are presented for the room temperature formed nanoparticle products.

Formation of uranium based nanoparticles via gamma-irradiation

International Nuclear Information System (INIS)

Nenoff, Tina M.; Ferriera, Summer R.; Huang, Jianyu; Hanson, Donald J.

2013-01-01

Graphical abstract: TEM image of d-U nanoparticles formed in aqueous solution by gamma irradiation. Display Omitted -- Highlights: •d-U nanoparticles were grown in solution by gamma irradiation. •The reaction solution does not exceed 25 °C (room temperature). •Only after multiday exposure to air is there evidence of oxidation of the d-U nanoparticles. •Evidence of d-U alloy nanoparticle formation confirmed by TEM/energy-dispersive X-ray (EDS) analysis. -- Abstract: The ability to fabricate nuclear fuels at low temperatures allows for the production of complex Uranium metal and alloys with minimum volatility of alloy components in the process. Gamma irradiation is a valuable method for the synthesis of a wide range of metal-based nanoparticles. We report on the synthesis via room temperature radiolysis and characterization of uranium (depleted, d-U) metal and uranium–lathanide (d-ULn, Ln = lanthanide surrogates) alloy nanoparticles from aqueous acidic salt solutions. The lanthanide surrogates chosen include La and Eu due to their similarity in ionic size and charge in solution. Detailed characterization results including UV–vis, TEM/HR-TEM, and single particle EDX (elemental analyses) are presented for the room temperature formed nanoparticle products

Uranium recovering from slags generated in the metallic uranium by magnesiothermic reduction

International Nuclear Information System (INIS)

Fornarolo, F.; Carvalho, E.F. Urano de; Durazzo, M.; Riella, H.G.

2008-01-01

The Nuclear Fuel Center of IPEN/CNEN-SP has recent/y concluded a program for developing the fabrication technology of the nuclear fuel based on the U 3 Si 2 -Al dispersion, which is being used in the IEA-R1 research reactor. The uranium silicide (U 3 Si 2 ) fuel production starts with the uranium hexafluoride (UF 6 ) processing and uranium tetrafluoride (UF 4 ) precipitation. Then, the UF 4 is converted to metallic uranium by magnesiothermic reduction. The UF 4 reduction by magnesium generates MgF 2 slag containing considerable concentrations of uranium, which could reach 20 wt%. The uranium contained in that slag should be recovered and this work presents the results obtained in recovering the uranium from that slag. The uranium recovery is accomplished by acidic leaching of the calcined slag. The calcination transforms the metallic uranium in U 3 O 8 , promoting the pulverization of the pieces of metallic uranium and facilitating the leaching operation. As process variables, have been considered the nitric molar concentration, the acid excess regarding the stoichiometry and the leaching temperature. As result, the uranium recovery reached a 96% yield. (author)

Kr ion irradiation study of the depleted-uranium alloys

Science.gov (United States)

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

2010-12-01

Fuel development for the reduced enrichment research and test reactor (RERTR) program is tasked with the development of new low enrichment uranium nuclear fuels that can be employed to replace existing high enrichment uranium fuels currently used in some research reactors throughout the world. For dispersion type fuels, radiation stability of the fuel-cladding interaction product has a strong impact on fuel performance. Three depleted-uranium alloys are cast for the radiation stability studies of the fuel-cladding interaction product using Kr ion irradiation to investigate radiation damage from fission products. SEM analysis indicates the presence of the phases of interest: U(Al, Si) 3, (U, Mo)(Al, Si) 3, UMo 2Al 20, U 6Mo 4Al 43 and UAl 4. Irradiations of TEM disc samples were conducted with 500 keV Kr ions at 200 °C to ion doses up to 2.5 × 10 19 ions/m 2 (˜10 dpa) with an Kr ion flux of 10 16 ions/m 2/s (˜4.0 × 10 -3 dpa/s). Microstructural evolution of the phases relevant to fuel-cladding interaction products was investigated using transmission electron microscopy.

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

International Nuclear Information System (INIS)

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

1995-01-01

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

Recent irradiation tests of uranium-plutonium-zirconium metal fuel elements

International Nuclear Information System (INIS)

Pahl, R.G.; Lahm, C.E.; Villarreal, R.; Hofman, G.L.; Beck, W.N.

1986-09-01

Uranium-Plutonium-Zirconium metal fuel irradiation tests to support the ANL Integral Fast Reactor concept are discussed. Satisfactory performance has been demonstrated to 2.9 at.% peak burnup in three alloys having 0, 8, and 19 wt % plutonium. Fuel swelling measurements at low burnup in alloys to 26 wt % plutonium show that fuel deformation is primarily radial in direction. Increasing the plutonium content in the fuel diminishes the rate of fuel-cladding gap closure and axial fuel column growth. Chemical redistribution occurs by 2.1 at.% peak burnup and generally involves the inward migration of zirconium and outward migration of uranium. Fission gas release to the plenum ranges from 46% to 56% in the alloys irradiated to 2.9 at.% peak burnup. No evidence of deleterious fuel-cladding chemical or mechanical interaction was observed

Status of core conversion with LEU silicide fuel in JRR-4

Energy Technology Data Exchange (ETDEWEB)

Nakajima, Teruo; Ohnishi, Nobuaki; Shirai, Eiji [Japan Atomic Energy Research Institute, Ibaraki-ken (Japan)

1997-08-01

Japan Research Reactor No.4 (JRR-4) is a light water moderated and cooled, 93% enriched uranium ETR-type fuel used and swimming pool type reactor with thermal output of 3.5MW. Since the first criticality was achieved on January 28, 1965, JRR-4 has been used for shielding experiments, radioisotope production, neutron activation analyses, training for reactor engineers and so on for about 30 years. Within the framework of the RERTR Program, the works for conversion to LEU fuel are now under way, and neutronic and thermal-hydraulic calculations emphasizing on safety and performance aspects are being carried out. The design and evaluation for the core conversion are based on the Guides for Safety Design and Evaluation of research and testing reactor facilities in Japan. These results show that the JRR-4 will be able to convert to use LEU fuel without any major design change of core and size of fuel element. LEU silicide fuel (19.75%) will be used and maximum neutron flux in irradiation hole would be slightly decreased from present neutron flux value of 7x10{sup 13}(n/cm{sup 2}/s). The conversion works are scheduled to complete in 1998, including with upgrade of the reactor building and utilization facilities.

Status of core conversion with LEU silicide fuel in JRR-4

International Nuclear Information System (INIS)

Nakajima, Teruo; Ohnishi, Nobuaki; Shirai, Eiji

1997-01-01

Japan Research Reactor No.4 (JRR-4) is a light water moderated and cooled, 93% enriched uranium ETR-type fuel used and swimming pool type reactor with thermal output of 3.5MW. Since the first criticality was achieved on January 28, 1965, JRR-4 has been used for shielding experiments, radioisotope production, neutron activation analyses, training for reactor engineers and so on for about 30 years. Within the framework of the RERTR Program, the works for conversion to LEU fuel are now under way, and neutronic and thermal-hydraulic calculations emphasizing on safety and performance aspects are being carried out. The design and evaluation for the core conversion are based on the Guides for Safety Design and Evaluation of research and testing reactor facilities in Japan. These results show that the JRR-4 will be able to convert to use LEU fuel without any major design change of core and size of fuel element. LEU silicide fuel (19.75%) will be used and maximum neutron flux in irradiation hole would be slightly decreased from present neutron flux value of 7x10 13 (n/cm 2 /s). The conversion works are scheduled to complete in 1998, including with upgrade of the reactor building and utilization facilities

Corrosion testing of uranium silicide fuel specimens

International Nuclear Information System (INIS)

Bourns, W.T.

1968-09-01

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

Kr ion irradiation study of the depleted-uranium alloys

Energy Technology Data Exchange (ETDEWEB)

Gan, J., E-mail: Jian.Gan@inl.go [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Keiser, D.D. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Miller, B.D. [University of Wisconsin, 1500 Engineering Drive, Madison, WI 53706 (United States); Kirk, M.A.; Rest, J. [Argonne National Laboratory, 9700 South Cass Ave., Argonne, IL 60439 (United States); Allen, T.R. [University of Wisconsin, 1500 Engineering Drive, Madison, WI 53706 (United States); Wachs, D.M. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States)

2010-12-01

Fuel development for the reduced enrichment research and test reactor (RERTR) program is tasked with the development of new low enrichment uranium nuclear fuels that can be employed to replace existing high enrichment uranium fuels currently used in some research reactors throughout the world. For dispersion type fuels, radiation stability of the fuel-cladding interaction product has a strong impact on fuel performance. Three depleted-uranium alloys are cast for the radiation stability studies of the fuel-cladding interaction product using Kr ion irradiation to investigate radiation damage from fission products. SEM analysis indicates the presence of the phases of interest: U(Al, Si){sub 3}, (U, Mo)(Al, Si){sub 3}, UMo{sub 2}Al{sub 20}, U{sub 6}Mo{sub 4}Al{sub 43} and UAl{sub 4}. Irradiations of TEM disc samples were conducted with 500 keV Kr ions at 200 {sup o}C to ion doses up to 2.5 x 10{sup 19} ions/m{sup 2} ({approx}10 dpa) with an Kr ion flux of 10{sup 16} ions/m{sup 2}/s ({approx}4.0 x 10{sup -3} dpa/s). Microstructural evolution of the phases relevant to fuel-cladding interaction products was investigated using transmission electron microscopy.

Reactivity management and burn-up management on JRR-3 silicide-fuel-core

International Nuclear Information System (INIS)

Kato, Tomoaki; Araki, Masaaki; Izumo, Hironobu; Kinase, Masami; Torii, Yoshiya; Murayama, Yoji

2007-08-01

On the conversion from uranium-aluminum-dispersion-type fuel (aluminide fuel) to uranium-silicon-aluminum-dispersion-type fuel (silicide fuel), uranium density was increased from 2.2 to 4.8 g/cm 3 with keeping uranium-235 enrichment of 20%. So, burnable absorbers (cadmium wire) were introduced for decreasing excess reactivity caused by the increasing of uranium density. The burnable absorbers influence reactivity during reactor operation. So, the burning of the burnable absorbers was studied and the influence on reactor operation was made cleared. Furthermore, necessary excess reactivity on beginning of operation cycle and the time limit for restart after unplanned reactor shutdown was calculated. On the conversion, limit of fuel burn-up was increased from 50% to 60%. And the fuel exchange procedure was changed from the six-batch dispersion procedure to the fuel burn-up management procedure. The previous estimation of fuel burn-up was required for the planning of fuel exchange, so that the estimation was carried out by means of past operation data. Finally, a new fuel exchange procedure was proposed for effective use of fuel elements. On the procedure, burn-up of spent fuel was defined for each loading position. The average length of fuel's staying in the core can be increased by two percent on the procedure. (author)

Corrosion testing of uranium silicide fuel specimens

Energy Technology Data Exchange (ETDEWEB)

Bourns, W T

1968-09-15

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

Fuel cycle cost comparisons with oxide and silicide fuels

Energy Technology Data Exchange (ETDEWEB)

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

1983-09-01

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

Technology for manufacturing dispersion nuclear fuel at Instituto de Pesquisas Energeticas e Nucleares IPEN/CNEN-SP, Brazil

International Nuclear Information System (INIS)

Durazzo, M.; Carvalho, E.F. Urano de; Saliba-Silva, A.M.; Souza, J.A.B.; Riella, H.G.

2008-01-01

IPEN has been working for increasing radioisotope production in order to supply the expanding demand for radiopharmaceutical medicines requested by the Brazilian welfare. To reach this objective, the IEA-R1 research reactor power capacity was recently increased from 2 MW to 3.5 MW. Since 1988 IPEN has been manufacturing its own fuel element, initially based on U 3 O 8 -Al dispersion fuel plates with 2.3 g U/cm 3 . To support the reactor power increase, higher uranium density had to be achieved for better irradiation flux and also to minimize the irradiated fuel elements to be stored. Uranium silicide was the chosen option. This paper describes the results of this program and the current status of silicide fuel fabrication and qualification. (author)

Irradiation defects in clayey minerals in association with discordance-type uranium deposit; Les Defauts d'Irradiation dans les Mineraux argileux associes aux gisements d'Uranium de type Discordance

Energy Technology Data Exchange (ETDEWEB)

Morichon, E.; Beaufort, D. [Universite de Poitiers, Laboratoire HydrASA, CNRS-FRE 3114, 86 - Poitiers (France); Morichon, E.; Allard, Th. [IMPMC, UMR 7590, 75 - Paris (France)

2009-07-01

Radioactivity generates defects in minerals and these defects are the witnesses of the presence of radio-elements, and therefore represent an interesting potential for uranium prospecting. Investigations made in the Athabasca basin in Canada reveal irradiation defects in very old clays (kaolinite, illite and sudoite) in the alteration halo of discordance-type uranium deposits. The authors comment the defect concentration variation among the different drillings. These differences show that hexavalent uranium circulated in the whole geological system

Irradiation of diffusion couples U-Mo/Al. Thermal calculation

International Nuclear Information System (INIS)

Fortis, Ana M.; Mirandou, Monica; Denis, Alicia C.

2004-01-01

The development of new low enrichment fuel elements for research reactors has lead to obtaining a number of compounds and alloys where the decrease in the enrichment is compensated by a higher uranium density in the fuel material. This has been achieved in particular with the uranium silicides dispersed in an aluminum matrix, where uranium densities about 4.8 g/cm 3 have been reached. Among the diverse candidate alloys, those of U-Mo with molybdenum content in the range 6 to 10 w % can yield, upon dispersion, to uranium densities of about 8 g/cm 3 . The first irradiation experiments employing these alloys in fuel plates, either dispersed in Al or monolithic revealed certain phenomena which are worthy of further studies. Failures have been detected apparently due to the formation of reaction products between the fissile material and the aluminum matrix, which exhibit a poor irradiation behavior. An experiment was designed which final purpose is to irradiate diffusion couples U-Mo/Al in the RA-3 reactor and to analyze the interaction zone at the working temperatures of the fuel elements. A simple device was built consisting of two Al 6063 blocks which press the U-Mo sample in between, located in an Al capsule. The ensemble is placed in a tube, which can be filled with different gases and introduced in the reactor. For safety reasons temperature predictions are necessary before performing the experiment. To this end, the COSMOS code was used. As a preliminary step and in order to test to exactness of the numerical estimations, two irradiations were performed in the RA-1 reactor with He and N 2 as transference gases. The agreement between the measured and calculated temperatures was good, particularly in the case of He and, along with the numerical predictions for the RA-3 reactor, provides a reliable basis to proceed with the following steps. (author)

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

Directory of Open Access Journals (Sweden)

H.J. RYU

2013-12-01

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

Continuing investigations for technology assessment of 99Mo production from LEU [low enriched uranium] targets

International Nuclear Information System (INIS)

Vandegrift, G.F.; Kwok, J.D.; Marshall, S.L.; Vissers, D.R.; Matos, J.E.

1987-01-01

Currently much of the world's supply of 99m Tc for medical purposes is produced from 99 Mo derived from the fissioning of high enriched uranium (HEU). This paper presents the results of our continuing studies on the effects of substituting low enriched uranium (LEU) for HEU in targets for the production of fission product 99 Mo. Improvements in the electrodeposition of thin films of uranium metal continue to increase the appeal for the substitution of LEU metal for HEU oxide films in cylindrical targets. The process is effective for targets fabricated from stainless steel or zircaloy. Included is a cost estimate for setting up the necessary equipment to electrodeposit uranium metal on cylindrical targets. Further investigations on the effect of LEU substitution on processing of these targets are also reported. Substitution of uranium silicides for the uranium-aluminium alloy or uranium aluminide dispersed fuel used in current target designs will allow the substitution of LEU for HEU in these targets with equivalent 99 Mo-yield per target and no change in target geometries. However, this substitution will require modifications in current processing steps due to 1) the insolubility of uranium silicides in alkaline solutions and 2) the presence of significant quantities of silicate in solution. Results to date suggest that substitution of LEU for HEU can be achieved. (Author)

Determination of uranium concentration and burn-up of irradiated reactor fuel in contaminated areas in Belarus using uranium isotopic ratios in soil samples

International Nuclear Information System (INIS)

Mironov, V.P.; Matusevich, J.L.; Kudrjashov, V.P.; Ananich, P.I.; Zhuravkov, V.V.; Boulyga, S.F.; Becker, J.S.

2005-01-01

An analytical method is described for the estimation of uranium concentrations, of 235 U/ 238 U and 236 U/ 238 U isotope ratios and burn-up of irradiated reactor uranium in contaminated soil samples by inductively coupled plasma mass spectrometry. Experimental results obtained at 12 sampling sites situated on northern and western radioactive fallout tails 4 to 53 km distant from Chernobyl nuclear power plant (NPP) are presented. Concentrations of irradiated uranium in the upper 0-10 cm soil layers at the investigated sampling sites varied from 2.1 x 10 -9 g/g to 2.0 x 10 -6 g/g depending mainly on the distance from Chernobyl NPP. A slight variation of the degree of burn-up of spent reactor uranium was revealed by analyzing 235 U/ 238 U and 236 U/ 238 U isotope ratios and the average value amounted to 9.4±0.3 MWd/(kg U). (orig.)

Determination of uranium concentration and burn-up of irradiated reactor fuel in contaminated areas in Belarus using uranium isotopic ratios in soil samples

Energy Technology Data Exchange (ETDEWEB)

Mironov, V.P.; Matusevich, J.L.; Kudrjashov, V.P.; Ananich, P.I.; Zhuravkov, V.V. [Inst. of Radiobiology, Minsk Univ. (Belarus); Boulyga, S.F. [Inst. of Inorganic Chemistry and Analytical Chemistry, Johannes Gutenberg-Univ. Mainz, Mainz (Germany); Becker, J.S. [Central Div. of Analytical Chemistry, Research Centre Juelich, Juelich (Germany)

2005-07-01

An analytical method is described for the estimation of uranium concentrations, of {sup 235}U/{sup 238}U and {sup 236}U/{sup 238}U isotope ratios and burn-up of irradiated reactor uranium in contaminated soil samples by inductively coupled plasma mass spectrometry. Experimental results obtained at 12 sampling sites situated on northern and western radioactive fallout tails 4 to 53 km distant from Chernobyl nuclear power plant (NPP) are presented. Concentrations of irradiated uranium in the upper 0-10 cm soil layers at the investigated sampling sites varied from 2.1 x 10{sup -9}g/g to 2.0 x 10{sup -6}g/g depending mainly on the distance from Chernobyl NPP. A slight variation of the degree of burn-up of spent reactor uranium was revealed by analyzing {sup 235}U/{sup 238}U and {sup 236}U/{sup 238}U isotope ratios and the average value amounted to 9.4{+-}0.3 MWd/(kg U). (orig.)

Recent Advances in Nb-silicide in-situ composites

International Nuclear Information System (INIS)

Bewlay, B.P.; Jackson, M.R.; Subramanian, P.R.; Briant, C.L.

2001-01-01

In-situ composites based on Nb silicides have great potential for future high-temperature applications. These Nb-silicide composites combine a ductile Nb-based matrix with high-strength silicides. With the appropriate combination of alloying elements, such as Ti, Hf, Cr, AI, it is possible to achieve a promising balance of fracture toughness, high-temperature creep performance, and oxidation resistance. This paper will describe the effect of volume fraction of silicide on microstructure, high-temperature creep performance, and oxidation resistance. The ratio of Nb:(W+Ti) is critical in determining both creep rate and oxidation performance. If this ratio goes below ∼1.5, the creep rate increases substantially. In more complex silicide-based systems, other intermetallics, such as laves phases and a boron-rich T-2 phase, are added for oxidation resistance. To understand the role of each phase on the creep resistance and oxidation performance of these composites, we determined the creep and oxidation behavior of the individual phases and composites at temperatures up to 1200 o C. These data allow quantification of the load-bearing capability of the individual phases in the Nb-silicide based in-situ composites. (author)

The development of uranium foil farication technology utilizing twin roll method for Mo-99 irradiation target

CERN Document Server

Kim, C K; Park, H D

2002-01-01

MDS Nordion in Canada, occupying about 75% of global supply of Mo-99 isotope, has provided the irradiation target of Mo-99 using the rod-type UAl sub x alloys with HEU(High Enrichment Uranium). ANL (Argonne National Laboratory) through co-operation with BATAN in Indonesia, leading RERTR (Reduced Enrichment for Research and Test Reactors) program substantially for nuclear non-proliferation, has designed and fabricated the annular cylinder of uranium targets, and successfully performed irradiation test, in order to develop the fabrication technology of fission Mo-99 using LEU(Low Enrichment Uranium). As the uranium foils could be fabricated in laboratory scale, not in commercialized scale by hot rolling method due to significant problems in foil quality, productivity and economic efficiency, attention has shifted to the development of new technology. Under these circumstances, the invention of uranium foil fabrication technology utilizing twin-roll casting method in KAERI is found to be able to fabricate LEU or...

TiSi2 integrity within a doped silicide process step

International Nuclear Information System (INIS)

Crean, G.M.; Cole, P.D.; Stoemenos, J.

1993-01-01

Degradation of arsenic implanted titanium silicide (TiSi 2 ) thin films as a result of thermal processing for shallow junction formation is investigated. Significant arsenic diffusion from the silicide overlayer into the silicon substrate has been detected by Rutherford Backscattering Spectrometry at drive-in temperatures > 1,050 C. Cross-sectional transmission electron micrographs have shown the silicide film become increasingly non-uniform as the thermal budget increases, ultimately leading to discontinuities forming in the silicide film. This observed degradation of the titanium silicide film is also supported by sheet resistance measurements which show the film to degrade significantly above a threshold thermal budget

Subsurface contributions in epitaxial rare-earth silicides

Energy Technology Data Exchange (ETDEWEB)

Luebben, Olaf; Shvets, Igor V. [Centre for Research on Adaptive Nanostructures and Nanodevices (CRANN), School of Physics, Trinity College, Dublin (Ireland); Cerda, Jorge I. [Instituto de Ciencia de Materiales de Madrid, ICMM-CSIC, Cantoblanco, Madrid (Spain); Chaika, Alexander N. [Institute of Solid State Physics, RAS, Chernogolovka (Russian Federation)

2015-07-01

Metallic thin films of heavy rare-earth silicides epitaxially grown on Si(111) substrates have been widely studied in recent years because of their appealing properties: unusually low values of the Schottky barrier height, an abrupt interface, and a small lattice mismatch. Previous studies also showed that these silicides present very similar atomic and electronic structures. Here, we examine one of these silicides (Gd{sub 3}Si{sub 5}) using scanning tunneling microscopy (STM) image simulations that go beyond the Tersoff-Hamann approach. These simulations strongly indicate an unusual STM depth sensitivity for this system.

Shielding Studies for Reducing the associated Radiological Risks Due To Irradiated Low Enriched Uranium Foil

International Nuclear Information System (INIS)

Margeanu, C.A.

2011-01-01

Present work estimates the radiation dose rates corresponding to irradiated Low Enriched Uranium (20 wt % 235 U) foil as part of shielding studies for radiological risks reduction after irradiation inside TRIGA 14 MW Research Reactor in an investigation on 99 Mo production possibility. Post-Irradiation Examination Laboratory's cell shielding calculations have been performed; radiation source was obtained by using ORIGEN-S code with specific cross-sections libraries. Different post-irradiation cooling times have been considered, gamma dose rates being estimated by using MAVRIC module from Scale 6 programs package, for following exposure situations (relative to Pie cell): i) front side, ii) lateral side and iii) back side. Three different calculations were performed: a) without any protection shield between operator and cell, except for the cell stainless steel wall; b) with a Lead protection shield between operator and cell and c) with a depleted Uranium shield, located inside the cell in between the radiation source and cell window. Radiation dose rates to cell external wall surface and for other eight fixed distances from cell wall were estimated. To obtain a consistent set of solutions, the study was done for various Uranium foil weights and different Lead and depleted Uranium shields thicknesses. Calculations were focused to assure that the dose rate to an operator positioned at 60 cm working distance from the cell will not exceed 0.02 mSv/h, maximum allowed dose rate for professionally exposed personnel according to Romanian regulations.

Morphology of Si/tungsten-silicides/Si interlayers

International Nuclear Information System (INIS)

Theodore, N.; Secco d'Aragona, F.; Blackstone, S.

1992-01-01

Tungsten and tungsten-silicides are of interest for semiconductor technology because of their refractory nature, low electrical-resistivity and high electromigration-resistance. This paper presents the first formation of buried tungsten-silicide layers in silicon, by proximity adhesion. The interlayers, created by a combination of chemical vapor-deposition (CVD) and proximity-adhesion were studied using transmission electron-microscopy (TEM). The behavior of the layers in the presence and absence of an adjacent silicon-dioxide interlayer was also investigated. Buried silicide layers were successfully formed with or without the adjacent silicon-dioxide. The silicide formed continuous layers with single grains encompassing the width of the interlayer. Individual grains were globular, with cusps at grain boundaries. This caused interlayer-thicknesses to be non-uniform, with lower thickness values being present at the cusps. Occasional voids were observed at grain-boundary cusps. The voids were smaller and less frequent in the presence of an adjacent oxide-layer, due to flow of the oxide during proximity adhesion. Electron-diffraction revealed a predominance of tungsten-disilicide in the interlayers, with some free tungsten being present. Stresses in the silicide layers caused occasional glide dislocations to propagate into the silicon substrate beneath the interlayers. The dislocations propagate only ∼100 nm into the substrate and therefore should not be detrimental to use of the buried layers. Occasional precipitates were observed at the end of glide-loops. These possibly arise due to excess tungsten from the interlayer diffusion down the glide dislocation to finally precipitate out as tungsten-silicide

A theoretical model of processes of irradiation of workers in uranium mines

International Nuclear Information System (INIS)

Zettwoog, Pierre.

1980-11-01

A quantity caled mine 'irradiativity', which is the ratio of collective dose to the production of uranium 'metal', is introduced. A distinction is made between irradiativity due to external exposure and that due to radon. A mathematical model to predict irradiativity has been developed using specific data from a mine: a) amount and distribution of work in areas of access and production; b) technology of ore extraction, especially the productivity of ore extraction, which is a basic parameter; c) protection technologies, especially the ventilation system; and d) geologic conditions. A numerical example is presented [fr

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

International Nuclear Information System (INIS)

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

1979-01-01

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

Analyses on Silicide Coating for LOCA Resistant Cladding

Energy Technology Data Exchange (ETDEWEB)

Sweidan, Faris B.; Lee, You Ho; Ryu, Ho Jin [KAIST, Daejeon (Korea, Republic of)

2015-10-15

A particular focus of accident-tolerant fuel has been cladding due to the rapid high-temperature oxidation of zirconium-based cladding with the evolution of H2 when steam is a reactant. Some key features of the coated cladding include high-temperature resistance to oxidation, lower processing temperatures, and a high melting point of the coating. Zirconium alloys exhibit a reasonably high melting temperature, so a coating for the cladding is appealing if the coating increases the high-temperature resistance to oxidation. In this case, the cladding is protected from complete oxidation. The cladding coating involves the application of zirconium silicide onto Zr-based cladding. Zirconium silicide coating is expected to produce a glassy layer that becomes more protective at elevated temperature. For this reason, silicide coatings on cladding offer the potential for improved reliability at normal operating temperatures and at the higher transient temperatures encountered during accidents. Although ceramic coatings are brittle and may have weak points to be used as coating materials, several ceramic coatings were successful and showed adherent behavior and high resistance to oxidation. In this study, the oxidation behavior of zirconium silicide and its oxidation kinetics are analyzed. Zirconium silicide is a new suggested material to be used as coatings on existing Zr-based cladding alloys, the aim of this study is to evaluate if zirconium silicide is applicable to be used, so they can be more rapidly developed using existing cladding technology with some modifications. These silicide coatings are an attractive alternative to the use of coatings on zirconium claddings or to the lengthy development of monolithic ceramic or ceramic composite claddings and coatings.

Analyses on Silicide Coating for LOCA Resistant Cladding

International Nuclear Information System (INIS)

Sweidan, Faris B.; Lee, You Ho; Ryu, Ho Jin

2015-01-01

A particular focus of accident-tolerant fuel has been cladding due to the rapid high-temperature oxidation of zirconium-based cladding with the evolution of H2 when steam is a reactant. Some key features of the coated cladding include high-temperature resistance to oxidation, lower processing temperatures, and a high melting point of the coating. Zirconium alloys exhibit a reasonably high melting temperature, so a coating for the cladding is appealing if the coating increases the high-temperature resistance to oxidation. In this case, the cladding is protected from complete oxidation. The cladding coating involves the application of zirconium silicide onto Zr-based cladding. Zirconium silicide coating is expected to produce a glassy layer that becomes more protective at elevated temperature. For this reason, silicide coatings on cladding offer the potential for improved reliability at normal operating temperatures and at the higher transient temperatures encountered during accidents. Although ceramic coatings are brittle and may have weak points to be used as coating materials, several ceramic coatings were successful and showed adherent behavior and high resistance to oxidation. In this study, the oxidation behavior of zirconium silicide and its oxidation kinetics are analyzed. Zirconium silicide is a new suggested material to be used as coatings on existing Zr-based cladding alloys, the aim of this study is to evaluate if zirconium silicide is applicable to be used, so they can be more rapidly developed using existing cladding technology with some modifications. These silicide coatings are an attractive alternative to the use of coatings on zirconium claddings or to the lengthy development of monolithic ceramic or ceramic composite claddings and coatings

Mössbauer spectroscopy study of surfactant sputtering induced Fe silicide formation on a Si surface

Energy Technology Data Exchange (ETDEWEB)

Beckmann, C.; Zhang, K. [2nd Institute of Physics, University of Göttingen, Friedrich-Hund-Platz 1, 37077 Göttingen (Germany); Hofsäss, H., E-mail: hans.hofsaess@phys.uni-goettingen.de [2nd Institute of Physics, University of Göttingen, Friedrich-Hund-Platz 1, 37077 Göttingen (Germany); Brüsewitz, C.; Vetter, U. [2nd Institute of Physics, University of Göttingen, Friedrich-Hund-Platz 1, 37077 Göttingen (Germany); Bharuth-Ram, K. [Physics Department, Durban University of Technology, Durban 4001 (South Africa)

2015-12-01

Highlights: • We study the formation of self-organized nanoscale dot and ripple patterns on Si. • Patterns are created by keV noble gas ion irradiation and simultaneous {sup 57}Fe co-deposition. • Ion-induced phase separation and the formation of a-FeSi{sub 2} is identified as relevant process. - Abstract: The formation of Fe silicides in surface ripple patterns, generated by erosion of a Si surface with keV Ar and Xe ions and simultaneous co-deposition of Fe, was investigated with conversion electron Mössbauer spectroscopy, atomic force microscopy and Rutherford backscattering spectrometry. For the dot and ripple patterns studied, we find an average Fe concentration in the irradiated layer between 6 and 25 at.%. The Mössbauer spectra clearly show evidence of the formation of Fe disilicides with Fe content close to 33 at.%, but very little evidence of the formation of metallic Fe particles. The results support the process of ion-induced phase separation toward an amorphous Fe disilicide phase as pattern generation mechanism. The observed amorphous phase is in agreement with thermodynamic calculations of amorphous Fe silicides.

Long term developments in irradiated natural uranium processing costs. Optimal size and siting of plants; Perspectives a long terme des couts de traitement de l'uranium naturel irradie. Tailles et localisations optimales des usines

Energy Technology Data Exchange (ETDEWEB)

Thiriet, L [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Oger, C; Vaumas, P de [Saint-Gobain Nucleaire, 92 - Courbevoie (France)

1964-07-01

The aim of this paper is to help solve the problem of the selection of optimal sizes and sites for spent nuclear fuel processing plants associated with power capacity programmes already installed. Firstly, the structure of capital and running costs of irradiated natural uranium processing plants is studied, as well as the influence of plant sizes on these costs and structures. Shipping costs from the production site to the plant must also be added to processing costs. An attempt to reach a minimum cost for the production of a country or a group of countries must therefore take into account both the size and the location of the plants. The foreseeable shipping costs and their structure (freight, insurance, container cost and depreciation), for spent natural uranium are indicated. Secondly, for various annual spent fuel reprocessing programmes, the optimal sizes and locations of the plants are determined. The sensitivity of the results to the basic assumptions relative to processing costs, shipping costs, the starting up year of the plant programme and the length of period considered, is also tested. - this rather complex problem, of a combinative nature, is solved through dynamic programming methods. - It is shown that these methods can also be applied to the problem of selecting the optimal sizes and locations of processing plants for MTR type fuel elements, related to research reactor programmes, as well as to future plutonium element processing plants related to breeder reactors. Thirdly, the case where yearly extraction of the plutonium contained in the irradiated natural uranium is not compulsory is examined; some stockpiling of the fuel is then allowed some years, entailing delayed processing. The load factor of such plants is thus greatly improved with respect to that of plants where the annual plutonium demand is strictly satisfied. By including spent natural uranium stockpiling costs an optimal rhythm of introduction and optimal sizes for spent fuel

Room temperature ferromagnetic gadolinium silicide nanoparticles

Science.gov (United States)

Hadimani, Magundappa Ravi L.; Gupta, Shalabh; Harstad, Shane; Pecharsky, Vitalij; Jiles, David C.

2018-03-06

A particle usable as T1 and T2 contrast agents is provided. The particle is a gadolinium silicide (Gd5Si4) particle that is ferromagnetic at temperatures up to 290 K and is less than 2 .mu.m in diameter. An MRI contrast agent that includes a plurality of gadolinium silicide (Gd.sub.5Si.sub.4) particles that are less than 1 .mu.m in diameter is also provided. A method for creating gadolinium silicide (Gd5Si4) particles is also provided. The method includes the steps of providing a Gd5Si4 bulk alloy; grinding the Gd5Si4 bulk alloy into a powder; and milling the Gd5Si4 bulk alloy powder for a time of approximately 20 minutes or less.

Irradiated uranium reprocessing, Final report - I-IV, Part I; Prerada ozracenog urana, Zavrsni izvestaj- I-VI, I Deo

Energy Technology Data Exchange (ETDEWEB)

Gal, I [Institute of Nuclear Sciences Boris Kidric, Laboratorija za visoku aktivnost, Vinca, Beograd (Serbia and Montenegro)

1961-12-15

This volume of the final report describes the construction and functioning of the cell for dissolving the irradiated uranium. Annexes of this report describe the parts of the cell as follows: description of the metal frame, grid and stainless steel plate; container for irradiated uranium; small opening of the cell; vessel for dissolving; device for opening the cover; device for sampling of the radioactive solution, inner and outer parts; pneumatic taps.

Continuing investigations for technology assessment of 99Mo production from LEU [low enriched Uranium] targets

International Nuclear Information System (INIS)

Vandergrift, G.F.; Kwok, J.D.; Marshall, S.L.; Vissers, D.R.; Matos, J.E.

1987-01-01

Currently much of the world's supply of /sup 99m/Tc for medical purposes is produced from 99 Mo derived from the fissioning of high enriched uranium (HEU). The need for /sup 99m/Tc is continuing to grow, especially in developing countries, where needs and national priorities call for internal production of 99 Mo. This paper presents the results of our continuing studies on the effects of substituting low enriched Uranium (LEU) for HEU in targets for the production of fission product 99 Mo. Improvements in the electrodeposition of thin films of uranium metal are reported. These improvements continue to increase the appeal for the substitution of LEU metal for HEU oxide films in cylindrical targets. The process is effective for targets fabricated from stainless steel or hastaloy. A cost estimate for setting up the necessary equipment to electrodeposit uranium metal on cylindrical targets is reported. Further investigations on the effect of LEU substitution on processing of these targets are also reported. Substitution of uranium silicides for the uranium-aluminum alloy or uranium aluminide dispersed fuel used in other current target designs will allow the substitution of LEU for HEU in these targets with equivalent 99 Mo-yield per target and no change in target geometries. However, this substitution will require modifications in current processing steps due to (1) the insolubility of uranium silicides in alkaline solutions and (2) the presence of significant quantities of silicate in solution. Results to date suggest that both concerns can be handled and that substitution of LEU for HEU can be achieved

Burn-Up Determination by High Resolution Gamma Spectrometry: Spectra from Slightly-Irradiated Uranium and Plutonium between 400-830 keV

International Nuclear Information System (INIS)

Forsyth, R.S.; Ronqvist, N.

1966-08-01

Previously published studies of the short-cooled fission product spectra of irradiated uranium have been severely restricted by the poor energy resolution of the sodium iodide detectors used. In this report are presented fission product spectra of irradiated uranium and plutonium obtained by means of a lithium-drifted germanium detector. The resolved gamma peaks have been assigned to various fission products by correlation of measured energy and half-life values with published data. By simultaneous study of the spectra of two irradiated mixtures of plutonium and uranium, the possibility of using the activities of Ru-103 and Ru-106 as a measure of the relative fission rate in U-235 and Pu-239 has been briefly examined

Burn-Up Determination by High Resolution Gamma Spectrometry: Spectra from Slightly-Irradiated Uranium and Plutonium between 400-830 keV

Energy Technology Data Exchange (ETDEWEB)

Forsyth, R S; Ronqvist, N

1966-08-15

Previously published studies of the short-cooled fission product spectra of irradiated uranium have been severely restricted by the poor energy resolution of the sodium iodide detectors used. In this report are presented fission product spectra of irradiated uranium and plutonium obtained by means of a lithium-drifted germanium detector. The resolved gamma peaks have been assigned to various fission products by correlation of measured energy and half-life values with published data. By simultaneous study of the spectra of two irradiated mixtures of plutonium and uranium, the possibility of using the activities of Ru-103 and Ru-106 as a measure of the relative fission rate in U-235 and Pu-239 has been briefly examined.

Palladium silicide - a new contact for semiconductor radiation detectors

International Nuclear Information System (INIS)

Totterdell, D.H.J.

1981-11-01

Silicide layers can be used as low resistance contacts in semiconductor devices. The formation of a metal rich palladium silicide Pd 2 Si is discussed. A palladium film 100A thick is deposited at 300 0 C and the resulting silicide layer used as an ohmic contact in an n + p silicon detector. This rugged contact has electrical characteristics comparable with existing evaporated gold contacts and enables the use of more reproducible bonding techniques. (author)

Irradiation of TZM: Uranium dioxide fuel pin at 1700 K

Science.gov (United States)

Mcdonald, G. E.

1973-01-01

A fuel pin clad with TZM and containing solid pellets of uranium dioxide was fission heated in a static helium-cooled capsule at a maximum surface temperature of 1700 K for approximately 1000 hr and to a total burnup of 2.0 percent of the uranium-235. The results of the postirradiation examination indicated: (1) A transverse, intergranular failure of the fuel pin occurred when the fuel pin reached 2.0-percent burnup. This corresponds to 1330 kW-hr/cu cm, where the volume is the sum of the fuel, clad, and void volumes in the fuel region. (2) The maximum swelling of the fuel pin was less than 1.5 percent on the fuel-pin diameter. (3) There was no visible interaction between the TZM clad and the UO2. (4) Irradiation at 1700 K produced a course-grained structure, with an average grain diameter of 0.02 centimeter and with some of the grains extending one-half of the thickness of the clad. (5) Below approximately 1500 K, the irradiation of the clad produced a moderately fine-grained structure, with an average grain diameter of 0.004 centimeter.

A study of CoSix silicide formed by recoil implantation

International Nuclear Information System (INIS)

Kwok, H.L.

1989-01-01

This work investigated the formation of CoSi x silicides on n-Si by recoil implantation through a thin cobalt layer using an inert gas ion beam. The results suggest the formation of a very shallow (35 to 45 nm) silicide surface layer under the specific conditions of preparation. The surface layer resistivity was comparable to values reported for Co 2 Si and CoSi, although below the surface, the resistivity decreased. This appeared to suggest a change-over from cobalt-rich silicides near the surface to a more conducting silicide (CoSi 2 ) at the interface. (author)

Phase analyses of silicide or nitride coated U–Mo and U–Mo–Ti particle dispersion fuel after out-of-pile annealing

Energy Technology Data Exchange (ETDEWEB)

Kim, Woo Jeong [Korea Atomic Energy Research Institute, 989-111 Daedeok-daero, Yuseong, Daejeon 305-353 (Korea, Republic of); Palancher, Hervé [CEA, DEN, DEC, F-13108 Saint Paul Lez Durance Cedex (France); Ryu, Ho Jin, E-mail: hojinryu@kaist.ac.kr [Korea Advanced Institute of Science and Technology, 291 Daehak-ro, Yuseong, Daejeon 305-701 (Korea, Republic of); Park, Jong Man; Nam, Ji Min [Korea Atomic Energy Research Institute, 989-111 Daedeok-daero, Yuseong, Daejeon 305-353 (Korea, Republic of); Bonnin, Anne [CEA, DEN, DEC, F-13108 Saint Paul Lez Durance Cedex (France); ESRF, 6, rue J. Horowitz, F-38000 Grenoble Cedex (France); Honkimäki, Veijo [ESRF, 6, rue J. Horowitz, F-38000 Grenoble Cedex (France); Charollais, François [CEA, DEN, DEC, F-13108 Saint Paul Lez Durance Cedex (France); Lemoine, Patrick [CEA, DEN, DISN, 91191 Gif sur Yvette (France)

2014-03-15

Highlights: • Silicide or nitride layers were coated on atomized U–Mo or U–Mo–Ti powder. • The constituent phases after annealing were identified through high-energy XRD. • U{sub 3}Si{sub 5} and U{sub 4}Mo(Mo{sub x}Si{sub 1−x})Si{sub 2} were identified in the silicide coating layers. • UN was identified for U–Mo particles and UN and U{sub 4}N{sub 7} formed on U–Mo–Ti particles. -- Abstract: The coating of silicide or nitride layers on U–7 wt%Mo or U–7 wt%Mo–1 wt%Ti particles has been proposed for the minimization of the interaction phase growth in U–Mo/Al dispersion fuel during irradiation. Out-of-pile annealing tests show reduced inter-diffusion by forming silicide or nitride protective layers on U–Mo and U–Mo–Ti particles. To characterize the constituent phases of the coated layers on U–Mo and U–Mo–Ti particles and the interaction phases of coated U–Mo and U–Mo–Ti particle dispersed Al matrix fuel, synchrotron X-ray diffraction experiments have been performed. It was identified that silicide coating layers consisted mainly of U{sub 3}Si{sub 5} and U{sub 4}Mo(Mo{sub x}Si{sub 1−x})Si{sub 2}, and nitride coating layers were composed of mainly UN and U{sub 4}N{sub 7}. The interaction phases obtained after annealing of coated U–Mo and U–Mo–Ti particle dispersion samples were identical to those found in U–Mo/Al–Si and U–Mo/Al systems. Nitride-coated particles showed less interaction formation than silicide-coated particles after annealing at 580 °C for 1 h owing to the higher susceptibility to breakage of the silicide coating layers during hot extrusion.

On the changing petroleum generation properties of Alum Shale over geological time caused by uranium irradiation

Science.gov (United States)

Yang, Shengyu; Schulz, Hans-Martin; Horsfield, Brian; Schovsbo, Niels H.; Noah, Mareike; Panova, Elena; Rothe, Heike; Hahne, Knut

2018-05-01

An interdisciplinary study was carried out to unravel organic-inorganic interactions caused by the radiogenic decay of uranium in the immature organic-rich Alum Shale (Middle Cambrian-Lower Ordovician). Based on pyrolysis experiments, uranium content is positively correlated with the gas-oil ratios and the aromaticities of both the free hydrocarbons residing in the rock and the pyrolysis products from its kerogen, indicating that irradiation has had a strong influence on organic matter composition overall and hence on petroleum potential. The Fourier Transform Ion Cyclotron Resonance mass spectrometry data reveal that macro-molecules in the uranium-rich Alum Shale samples are less alkylated than less irradiated counterparts, providing further evidence for structural alteration by α-particle bombardment. In addition, oxygen containing-compounds are enriched in the uranium-rich samples but are not easily degradable into low-molecular-weight products due to irradiation-induced crosslinking. Irradiation has induced changes in organic matter composition throughout the shale's entire ca. 500 Ma history, irrespective of thermal history. This factor has to be taken into account when reconstructing petroleum generation history. The Alum Shale's kerogen underwent catagenesis in the main petroleum kitchen area 420-340 Ma bp. Our calculations suggest the kerogen was much more aliphatic and oil-prone after deposition than that after extensive exposure to radiation. In addition, the gas sorption capacity of the organic matter in the Alum Shale can be assumed to have been less developed during Palaeozoic times, in contrast to results gained by sorption experiments performed at the present day, for the same reason. The kerogen reconstruction method developed here precludes overestimations of gas generation and gas retention in the Alum Shale by taking irradiation exposure into account and can thus significantly mitigate charge risk when applied in the explorations for both

Study on the uranium-cerium extraction and his application to the treatment of irradiated uranium

International Nuclear Information System (INIS)

Lobao, Afonso dos Santos Tome

1979-01-01

It was made a study on the behavior of uranium and cerium(IV) extraction, using the latter element as a plutonium simulator in a flowsheet of the treatment of irradiated uranium. Cerium(IV) was used under the same conditions as a plutonium in the Purex process because the admitted similar properties. An experimental work was initiated to determine the equilibrium curves of uranium, under the following conditions: concentration of 1 to 20 g U/1 and acidity varying from 1 to 5M in HNO 3 . Other parameters studied were the volumetric ratio of the phases and the influence of the concentration of TBP (tri-n-butyl phosphate). To guarantee the cerium(IV) extraction, the diluent (varsol) was previously treated with 10% potassium dichromate in perchloric acid, potassium permanganate in 1M sulphuric acid and concentrated sulphuric acid at 70 deg to eliminate reducing compounds. The results obtained for cerium extraction, allowed a better understanding of its behavior in solution. The results permitted to conclude that the decontamination for cerium are very high in the first Purex extraction cycle. The easy as cerium(IV) is reduced to the trivalent state contributes a great deal to its decontamination. (author)

Irradiation behavior of uranium-molybdenum dispersion fuel: Fuel performance data from RERTR-1 and RERTR-2

International Nuclear Information System (INIS)

Meyer, M.K.; Clark, C.R.; Hayes, S.L.; Strain, R.V.; Hofman, G.L.; Snelgrove, J.L.; Park, J.M.; Kim, K.H.

1999-01-01

This paper presents quantitative data on the irradiation behavior of uranium-molybdenum fuels from the low temperature RERTR-1 and -2 experiments. Fuel swelling measurements of U-Mo fuels at ∼40% and ∼70% burnup are presented. The rate of fuel-matrix interaction layer growth is estimated. Microstructures of fuel in the pre- and postirradiation condition were compared. Based on these data, a qualitative picture of the evolution of the U-Mo fuel microstructure during irradiation has been developed. Estimates of uranium-molybdenum fuel swelling and fuel-matrix interaction under high-power research reactor operating conditions are presented. (author)

Radiation hazards in the neighbourhood of uranium reactors; Dangers des rayonnements aupres des piles a uranium

Energy Technology Data Exchange (ETDEWEB)

Joffre, H [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1956-07-01

Radiation hazards near uranium reactors may be divided in two groups. Hazards when the reactor is normally operating: {gamma} radiation from hot uranium or air contamination by fission gases, {gamma} radiation or contamination by the coolant (air, nitrogen, heavy-water), {gamma} radiation from radioisotopes. Hazards in the case of an accident: presence of hot uranium in the atmosphere, soil contamination. (author) [French] Les dangers d'irradiation aupres des piles a uranium sont a classer essentiellement en deux groupes. Les dangers existant aupres d'une pile exploitee normalement: irradiation {gamma} par l'uranium irradie ou contamination de l'air par des gaz de fission, irradiation {gamma} ou contamination par les fluides de refroidissement (air, azote, eau lourde), irradiation {gamma} par les radioelements fabriques. Les dangers en cas d'accident survenant a un reacteur en fonctionnement, ayant pour consequence : la presence dans l'air d'uranium irradie, la contamination du sol. (auteur)

Neutronic design of the RSG-GAS silicide core

Energy Technology Data Exchange (ETDEWEB)

Sembiring, T.M.; Kuntoro, I.; Hastowo, H. [Center for Development of Research Reactor Technology National Nuclear Energy Agency BATAN, PUSPIPTEK Serpong Tangerang, 15310 (Indonesia)

2002-07-01

The objective of core conversion program of the RSG-GAS multipurpose reactor is to convert the fuel from oxide, U{sub 3}O{sub 8}-Al to silicide, U{sub 3}Si{sub 2}-Al. The aim of the program is to gain longer operation cycle by having, which is technically possible for silicide fuel, a higher density. Upon constraints of the existing reactor system and utilization, an optimal fuel density in amount of 3.55 g U/cc was found. This paper describes the neutronic parameter design of the silicide equilibrium core and the design of its transition cores as well. From reactivity control point of view, a modification of control rod system is also discussed. All calculations are carried out by means of diffusion codes, Batan-EQUIL-2D, Batan-2DIFF and -3DIFF. The silicide core shows that longer operation cycle of 32 full power days can be achieved without decreasing the safety criteria and utilization capabilities. (author)

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)

Effects of heat and pressure on the swelling of irradiated uranium. Part I

Energy Technology Data Exchange (ETDEWEB)

Churcman, A. T.; Barnes, R. S.; Cottrell, A. H.

1956-09-15

Small pieces of a natural uranium fuel bar from the NRX reactor which had been irradiated to 0.3 - 0.4% burn up have been heat treated either in vacuo or at high pressure and changes in their density measured.

Room temperature ferromagnetic gadolinium silicide nanoparticles

Energy Technology Data Exchange (ETDEWEB)

Hadimani, Magundappa Ravi L.; Gupta, Shalabh; Harstad, Shane; Pecharsky, Vitalij; Jiles, David C.

2018-03-06

A particle usable as T₁ and T₂ contrast agents is provided. The particle is a gadolinium silicide (Gd₅Si₄) particle that is ferromagnetic at temperatures up to 290 K and is less than 2 .mu.m in diameter. An MRI contrast agent that includes a plurality of gadolinium silicide (Gd.sub.5Si.sub.4) particles that are less than 1 .mu.m in diameter is also provided. A method for creating gadolinium silicide (Gd₅Si₄) particles is also provided. The method includes the steps of providing a Gd₅Si₄ bulk alloy; grinding the Gd₅Si₄ bulk alloy into a powder; and milling the Gd₅Si₄ bulk alloy powder for a time of approximately 20 minutes or less.

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

Determination of accurate metal silicide layer thickness by RBS

International Nuclear Information System (INIS)

Kirchhoff, J.F.; Baumann, S.M.; Evans, C.; Ward, I.; Coveney, P.

1995-01-01

Rutherford Backscattering Spectrometry (RBS) is a proven useful analytical tool for determining compositional information of a wide variety of materials. One of the most widely utilized applications of RBS is the study of the composition of metal silicides (MSi x ), also referred to as polycides. A key quantity obtained from an analysis of a metal silicide is the ratio of silicon to metal (Si/M). Although compositional information is very reliable in these applications, determination of metal silicide layer thickness by RBS techniques can differ from true layer thicknesses by more than 40%. The cause of these differences lies in how the densities utilized in the RBS analysis are calculated. The standard RBS analysis software packages calculate layer densities by assuming each element's bulk densities weighted by the fractional atomic presence. This calculation causes large thickness discrepancies in metal silicide thicknesses because most films form into crystal structures with distinct densities. Assuming a constant layer density for a full spectrum of Si/M values for metal silicide samples improves layer thickness determination but ignores the underlying physics of the films. We will present results of RBS determination of the thickness various metal silicide films with a range of Si/M values using a physically accurate model for the calculation of layer densities. The thicknesses are compared to scanning electron microscopy (SEM) cross-section micrographs. We have also developed supporting software that incorporates these calculations into routine analyses. (orig.)

Contribution to the study of defects created by {alpha} particles in uranium at 4.2 K; Contribution a l'etude des defauts crees par irradiation a l'aide de particules {alpha} dans l'uranium a 4.2 K

Energy Technology Data Exchange (ETDEWEB)

Raharinaivo, A L [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

1969-07-01

A device is described for the irradiation, in liquid helium, of metallic strips with {alpha} particles produced by radioactive sources. It has thereby been possible to measure changes in resistivity of variously treated uranium samples (cold- worked, annealed, previously exposed to neutrons, etc. ) as a function of the irradiation flux. The annealings carried out after irradiation compare favorably to those effected after a quenching from 100 to 4 K (JOUSSET experiments). The results are discussed; it is concluded that a defect, very probably of the interstitial type, is mobile in uranium at temperatures below 5 K. (author) [French] On decrit un dispositif permettant d'irradier, dans l'helium liquide, des lames metalliques par des particules {alpha} issues de sources radioactives. On a ainsi mesure les variations de resistivite, en fonction du flux d'irradiation, d'uranium diversement traite (ecroui, recuit, prealablement irradie par des neutrons...). Les recuits apres irradiation se comparent bien aux recuits apres trempe de 100 a 4 K (experiences de JOUSSET). L'ensemble des resultats est discute et il conduit a la conclusion qu'un defaut, tres vraisemblablement interstitiel, est mobile dans l'uranium a des temperatures inferieures a 5 K. (auteur)

Determination of uranium metal concentration in irradiated fuel storage basin sludge using selective dissolution

International Nuclear Information System (INIS)

Delegard, C.H.; Sinkov, S.I.; Chenault, J.W.; Schmidt, A.J.; Pool, K.N.; Welsh, T.L.

2014-01-01

Irradiated uranium metal fuel was stored underwater in the K East and K West storage basins at the US Department of Energy Hanford Site. The uranium metal under damaged cladding reacted with water to generate hydrogen gas, uranium oxides, and spalled uranium metal particles which intermingled with other particulates to form sludge. While the fuel has been removed, uranium metal in the sludge remains hazardous. An expeditious routine method to analyze 0.03 wt% uranium metal in the presence of >30 wt% total uranium was needed to support safe sludge management and processing. A selective dissolution method was designed based on the rapid uranium oxide dissolution but very low uranium metal corrosion rates in hot concentrated phosphoric acid. The uranium metal-bearing heel from the phosphoric acid step then is rinsed before the uranium metal is dissolved in hot concentrated nitric acid for analysis. Technical underpinnings of the selective dissolution method, including the influence of sludge components, were investigated to design the steps and define the reagents, quantities, concentrations, temperatures, and times within the selective dissolution analysis. Tests with simulant sludge proved the technique feasible. Tests with genuine sludge showed a 0.0028 ± 0.0037 wt% (at one standard deviation) uranium metal analytical background, a 0.011 wt% detection limit, and a 0.030 wt% quantitation limit in settled (wet) sludge. In tests using genuine K Basin sludge spiked with uranium metal at concentrations above the 0.030 wt% ± 25 % (relative) quantitation limit, uranium metal recoveries averaged 99.5 % with a relative standard deviation of 3.5 %. (author)

Si-Ge Nano-Structured with Tungsten Silicide Inclusions

Science.gov (United States)

Mackey, Jon; Sehirlioglu, Alp; Dynys, Fred

2014-01-01

Traditional silicon germanium high temperature thermoelectrics have potential for improvements in figure of merit via nano-structuring with a silicide phase. A second phase of nano-sized silicides can theoretically reduce the lattice component of thermal conductivity without significantly reducing the electrical conductivity. However, experimentally achieving such improvements in line with the theory is complicated by factors such as control of silicide size during sintering, dopant segregation, matrix homogeneity, and sintering kinetics. Samples are prepared using powder metallurgy techniques; including mechanochemical alloying via ball milling and spark plasma sintering for densification. In addition to microstructural development, thermal stability of thermoelectric transport properties are reported, as well as couple and device level characterization.

System for uranium superficial density measurement in U3Si2 MTR fuel plates using radiography

International Nuclear Information System (INIS)

Hey, Martin A.; Gomez Marlasca, Fernando

2003-01-01

The paper describes a method for measuring uranium superficial density in high density uranium silicide (U 3 Si 2 ) MTR fuel plates, through the use of industrial radiography, a set of patterns built for this purpose, a transmission optical densitometer, and a quantitative model of analysis and measurement. Our choice for this particular method responds to its high accuracy, low cost and easy implementation according to the standing quality control systems. (author)

Advances in the manufacturing and irradiation of reduced enrichment fuels for canadian research reactors

International Nuclear Information System (INIS)

Wood, J.C.; Foo, M.T.; Berthiaume, L.C.; Herbert, L.N.; Schaefer, J.D.

1984-01-01

The procedures for manufacturing fuel rods of uranium silicide dispersed in aluminum and clad in aluminum have been optimized to maximize production rates while minimizing scrap losses. Melting and casting, chip machining and core extrusion have all been re-evaluated to improve their efficiency and significant gains have been made, whilst maintaining high quality standards. The results of our irradiation program on mini-elements up to a burnup of 80 atomic percent continue to be encouraging. The upper bound curve of fuel core swelling versus burnup in the range 0-80 atomic percent represents 1% swelling per 10 atomic percent burnup. Fuel core swelling has now been measured directly on six mini-elements from which the clad surface oxide had been removed showing that previous calculated values of core swelling were marginally conservative. (author)

Internal irradiation by radon daughters in Bulgarian uranium mines over the period 1956-1990

International Nuclear Information System (INIS)

Dimitrov, M.; Presiyanov, D.

1993-01-01

The results of over 30-years' radiation monitoring in Bulgarian uranium mines are summed up in retrospect. The overall organization and the parameters monitored during the different periods are discussed. A radiological characteristic of the working environment is presented which includes radon and its decay products. The internal irradiation levels during different time periods and in different mines have been estimated. The health costs of uranium mining in Bulgaria are discussed and conclusions and recommendations for radiological control are made. (author)

Silicide formation by Ar/sup +/ ion bombardment of Pd/Si

Energy Technology Data Exchange (ETDEWEB)

Lee, R Y; Whang, C N; Kim, H K; Smith, R J

1988-08-01

Palladium films, 45 nm thick, evaporated on to Si(111) were irradiated to various doses with 78 keV Ar/sup +/ ions to promote silicide formation. Rutherford backscattering spectroscopy (RBS) shows that intermixing has occurred across the Pd/Si interface at room temperature. The mixing behaviour is increased with dose which coincides well with the theoretical model of cascade mixing. The absence of deep RBS tails for palladium and the small area of this for silicon spectra indicate that short-range mixing occurs. From the calculated damage profiles computed with TRIM code, the dominant diffusion species is found to be silicon atoms in the Pd/Si system. It is also found that the initial compound formed by Ar/sup +/ irradiation is Pd/sub 2/Si which increases with dose. At a dose of 1 x 10/sup 16/ Ar/sup +/ cm/sup -2/, a 48 nm thickness of Pd/sub 2/Si was formed by ion-beam mixing at room temperature.

Synthesis and design of silicide intermetallic materials

Energy Technology Data Exchange (ETDEWEB)

Petrovic, J.J.; Castro, R.G.; Butt, D.P. [Los Alamos National Lab., NM (United States)] [and others

1997-04-01

The overall objective of this program is to develop structural silicide-based materials with optimum combinations of elevated temperature strength/creep resistance, low temperature fracture toughness, and high temperature oxidation and corrosion resistance for applications of importance to the U.S. processing industry. A further objective is to develop silicide-based prototype industrial components. The ultimate aim of the program is to work with industry to transfer the structural silicide materials technology to the private sector in order to promote international competitiveness in the area of advanced high temperature materials and important applications in major energy-intensive U.S. processing industries. The program presently has a number of developing industrial connections, including a CRADA with Schuller International Inc. targeted at the area of MoSi{sub 2}-based high temperature materials and components for fiberglass melting and processing applications. The authors are also developing an interaction with the Institute of Gas Technology (IGT) to develop silicides for high temperature radiant gas burner applications, for the glass and other industries. Current experimental emphasis is on the development and characterization of MoSi{sub 2}-Si{sub 3}N{sub 4} and MoSi{sub 2}-SiC composites, the plasma spraying of MoSi{sub 2}-based materials, and the joining of MoSi{sub 2} materials to metals.

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

Energy Technology Data Exchange (ETDEWEB)

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

2012-10-15

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

Irradiation behaviour of mixed uranium-plutonium carbides, nitrides and carbonitrides; Comportement a l'irradiation de carbures, nitrures et carbonitrures mixtes d'uranium et de plutonium

Energy Technology Data Exchange (ETDEWEB)

Mikailoff, H; Mustelier, J P; Bloch, J; Leclere, J; Hayet, L [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

1967-07-01

In the framework of the research program of fast reactor fuels two irradiation experiments have been carried out on mixed uranium-plutonium carbides, nitrides and carbo-nitrides. In the first experiment carried out with thermal neutrons, the fuel consisted of sintered pellets sheathed in a stainless steel can with a small gap filled with helium. There were three mixed mono-carbide samples and the maximum linear power was 715 W/cm. After a burn-up slightly lower than 20000 MW day/tonne, a swelling of the fuel which had ruptured the cans was observed. In the second experiment carried out in the BR2 reactor with epithermal neutrons, the samples consisted of sintered pellets sodium bonded in a stainless steel tube. There were three samples containing different fuels and the linear power varies between 1130 and 1820 W/cm. Post-irradiation examination after a maximal burn-up of 1550 MW day/tonne showed that the behaviour of the three fuel elements was satisfactory. (authors) [French] Dans le cadre du programme d'etude des conibustiles pour reacteurs rapides, on a realise deux experiences d'irradiation de carbures, nitrures et carbonitrures mixtes d'uranium et de plutonium. Dans la premiere experience, faite en neutrons thermiques, le combustible etait constitue de,pastilles frittees gainees dans un tube d'acier inoxydable avec un faible jeu rempli d'helium. Il y avait trois echantillons de monocarbures mixtes, et la puissance lineaire maximale etait de 715 W/cm. Apres un taux de combustion legerement inferieur a 20 000 MWj/t, on a observe un gonflement des combustible qui a provoque, la rupture des gaines. Pans la seconde experience, realisee dans le reacteur BR2 en neutrons epithermiques, les echantillons etaient constitues de pastilles frittees gainees dans un tube d'acier avec un joint sodium. Il y avait trois echantillons contenant des combustibles differents, et la puissance lineaire variait de 1130 a 1820 W/cm. Les examens apres irradiation a un taux maximal de

Current status of irradiation facilities in JRR-3 and JRR-4

International Nuclear Information System (INIS)

Hori, Naohiko; Wada, Shigeru; Sasajima, Fumio; Kusunoki, Tsuyoshi

2006-01-01

The Department of Research Reactor has operated two research reactors, JRR-3 and JRR-4. These reactors were constructed in the Tokai Research Establishment. Many researchers and engineers use these joint-use facilities. JRR-3 is a light water moderated and cooled, pool type research reactor using low-enriched silicide fuel. JRR-3's maximum thermal power is 20MW. JRR-3 has nine vertical irradiation holes for RI production, nuclear fuels and materials irradiation at reactor core area. JRR-3 has many kinds of irradiation holes in a heavy water tank around the reactor core. These are two hydraulic rabbit irradiation facilities, two pneumatic rabbit irradiation facilities, one activation analysis irradiation facilities, one uniform irradiation facility, one rotating irradiation facility and one capsule irradiation facility. JRR-3 has nine horizontal experimental holes, that are used by many kinds of neutron beam experimental facilities using these holes. JRR-4 is a light water moderated and cooled, swimming pool type research reactor using low-enriched silicide fuel. JRR-4's maximum thermal power is 3.5MW. JRR-4 has five vertical irradiation tubes at reactor core area, three capsule irradiation facilities, one hydraulic rabbit irradiation facility, and one pneumatic rabbit irradiation facility. JRR-4 has a neutron beam hole, and it has used neutron beam experiments, irradiations for activation analysis and medical neutron irradiations. (author)

The uranium and thorium separation in the chemical reprocessing of the irradiated fuel of thorium and uranium mixed oxides

International Nuclear Information System (INIS)

Oliveira, E.F. de.

1984-09-01

A bibliographic research has been carried out for reprocessing techniques of irradiated thorium fuel from nuclear reactors. The Thorex/Hoechst process has been specially considered to establish a method for reprocessing thorium-uranium fuel from PWR. After a series of cold tests performed in laboratory it was possible to set the behavior of several parameters affecting the Thorex/Hoechst process. Some comments and suggestions are presented for modifications in the process flosheet conditions. A discussion is carried out for operational conditions such as the aqueous to organic flow ratio the acidity of strip and scrub solutions in the process steps for thorium and uranium recovery. The operation diagrams have been constructed using equilibrium experimental data which correspond to conditions observed in laboratory. (Author) [pt

Magnesium silicide production and silane synthesis on its basis

International Nuclear Information System (INIS)

Taurbaev, T.I.; Mukashev, F.A.; Manakov, S.M.; Francev, U.V.; Kalblanbekov, B.M.; Akhter, P.; Abbas, M.; Hussain, A.

2003-01-01

We had developed an alternative method of production of magnesium silicide with use of ferroalloys of silicon. Magnesium silicide is raw material for silane synthesis. The essence of the method consist of sintering FS -75 (ferrosilicium with 75 % of silicon and 25 % of iron, made by ferroalloy factories) with metal magnesium at temperature of 650 deg. C. The X-ray analysis has shown formation of magnesium silicide. That is further used for synthesis of silane. The output of silane is 60 % in respect of the contents of silicon. After removing the water vapors the mass-spectrometer analysis has estimated the purity of silane as 99.95 % with no detection of phosphine and diborane. (author)

ELECTROCHEMICAL STUDIES OF URANIUM METAL CORROSION MECHANISM AND KINETICS IN WATER

International Nuclear Information System (INIS)

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

2006-01-01

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

Thermal stability of Ni-Pt-Ta alloy silicides on epi-Si1-xCx

International Nuclear Information System (INIS)

Yoo, Jung-Ho; Chang, Hyun-Jin; Min, Byoung-Gi; Ko, Dae-Hong; Cho, Mann-Ho; Sohn, Hyunchul; Lee, Tae-Wan

2008-01-01

We investigated the silicide formation in Ni/epi-Si 1-x C x systems. Ni-Pt and Ni-Pt-Ta films were deposited on epi-Si 1-x C x /Si substrates by DC magnetron sputtering and processed at various temperatures. The sheet resistance of the silicide from the Ni alloy/epi-Si 1-x C x systems was maintained at low values compared to that from Ni/Si systems. By TEM and EDS analyses, we confirmed the presence of a Pt alloy layer at the top of the Ni-silicide layer. The stability of the silicide layer in the Ni alloy/epi-Si 1-x C x system is explained by not only the Pt rich layer on the top of the Ni-silicide layer, but also by the presence of a small amount of Pt in the Ni-silicide layer or at the grain boundaries. And both the thermal stability and the morphology of silicide were greatly improved by the addition of Ta in Ni-Pt films

Development of an alternative process for recovery of uranium from rejected plates in the manufacture of MTR type fuel elements

International Nuclear Information System (INIS)

Flores Gonzalez, Jocelyn Natalia

2011-01-01

This work discusses the recovery of enriched uranium in U 235 , from fuel plates rejected during the fuel elements manufacturing process for the La Reina Nuclear Studies Center, RECH-1, CCHEN. The plates have an aluminum based alloy coating, AISI-SAE 6061, with U 3 Si 2 powder distributed evenly inside and dispersed in an aluminum matrix. The high cost of enriched uranium means that it must be recovered from plates rejected in the production process because of non-compliance with the plate specifications, and also because some of them undergo destructive testing, to measure the aluminum coating's thickness on each side of the plate. The thickness of the uranium nucleus is measured as well and the size of the defects on the ends of the plate such as 'dog bone' and 'fish tail', that is, for the purposes of quality control. The first step in the process is carried out by dissolving the aluminum in a hot solution of NaOH in order to release the uranium silicide powder that is insoluble in the soda. A second step involves dissolving the uranium silicide in a hot HNO 3 solution, followed by washing and filtering, and then extracting the SX and analyzing its behavior during this stage. During the process 98.9% of the uranium is recovered together with a solution that is enough for the SX process given the experiences that were carried out in the extraction stage

Contribution to the study of defects created by {alpha} particles in uranium at 4.2 K; Contribution a l'etude des defauts crees par irradiation a l'aide de particules {alpha} dans l'uranium a 4.2 K

Energy Technology Data Exchange (ETDEWEB)

Raharinaivo, A.L. [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

1969-07-01

A device is described for the irradiation, in liquid helium, of metallic strips with {alpha} particles produced by radioactive sources. It has thereby been possible to measure changes in resistivity of variously treated uranium samples (cold- worked, annealed, previously exposed to neutrons, etc. ) as a function of the irradiation flux. The annealings carried out after irradiation compare favorably to those effected after a quenching from 100 to 4 K (JOUSSET experiments). The results are discussed; it is concluded that a defect, very probably of the interstitial type, is mobile in uranium at temperatures below 5 K. (author) [French] On decrit un dispositif permettant d'irradier, dans l'helium liquide, des lames metalliques par des particules {alpha} issues de sources radioactives. On a ainsi mesure les variations de resistivite, en fonction du flux d'irradiation, d'uranium diversement traite (ecroui, recuit, prealablement irradie par des neutrons...). Les recuits apres irradiation se comparent bien aux recuits apres trempe de 100 a 4 K (experiences de JOUSSET). L'ensemble des resultats est discute et il conduit a la conclusion qu'un defaut, tres vraisemblablement interstitiel, est mobile dans l'uranium a des temperatures inferieures a 5 K. (auteur)

Separation of rubidium from irradiated aluminum-encapsulated uranium

International Nuclear Information System (INIS)

Horwitz, E.P.; Schmitz, F.J.; Rokop, D.J.

1982-01-01

A procedure was developed for separating rubidium from irradiated aluminum encapsulated uranium. The separations procedure produces a final ultra-high purity RbCl product for subsequent high performance mass spectrometric analysis. The procedure involves first removing most of the macro-components and fission products by strong base anion exchange using, first, concentrated HCl, then oxalic acid media and second, selectively separating rubidium from alkaline-earth ions and other alkali-metal ions, including cesium, using Bio-Rex-40 cation-exchange resin. The resultant RbCl is then put through a final vacuum sublimation step. Ultra-pure reagents and specially clean glassware are used throughout the procedure to minimize contamination by naturally-occurring rubidium

Fuels for Canadian research reactors

International Nuclear Information System (INIS)

Feraday, M.A.

1993-01-01

This paper includes some statements and remarks concerning the uranium silicide fuels for which there is significant fabrication in AECL, irradiation and defect performance experience; description of two Canadian high flux research reactors which use high enrichment uranium (HEU) and the fuels currently used in these reactors; limited fabrication work done on Al-U alloys to uranium contents as high as 40 wt%. The latter concerns work aimed at AECL fast neutron program. This experience in general terms is applied to the NRX and NRU designs of fuel

Obtention of uranium tetrafluoride from effluents generated in the hexafluoride conversion process

International Nuclear Information System (INIS)

Silva Neto, J.B.; Urano de Carvalho, E.F.; Durazzo, M.; Riella, H.G.

2009-01-01

Full text: The uranium silicide (U3Si2) fuel is produced from uranium hexafluoride (UF6) as the primary raw material. The uranium tetrafluoride (UF4) and metallic uranium are the two subsequent steps. There are two conventional routes for UF4 production: the first one reduces the uranium from the UF6 hydrolysis solution by adding stannous chloride (SnCl2). The second one is based on the hydrofluorination of solid uranium dioxide (UO2) produced from the ammonium uranyl carbonate (AUC). This work introduces a third route, a dry way route which utilizes the recovering of uranium from liquid effluents generated in the uranium hexafluoride reconversion process adopted at IPEN/CNEN-SP. Working in the liquid phase, this route comprises the recovery of ammonium fluoride by NH4HF2 precipitation. The crystallized bifluoride is added to the solid UO2 to get UF4, which returns to the metallic uranium production process and, finally, to the U3Si2 powder production. The UF4 produced by this new route was chemically and physically characterized and will be able to be used as raw material for metallic uranium production by magnesiothermic reduction. (author)

Mechanoactivation of chromium silicide formation in the SiC-Cr-Si system

Directory of Open Access Journals (Sweden)

Vlasova M.

2002-01-01

Full Text Available The processes of simultaneous grinding of the components of a SiC-Cr-Si mixture and further temperature treatment in the temperature range 1073-1793 K were studied by X-ray phase analysis, IR spectroscopy, electron microscopy, and X-ray microanalysis. It was established that, during grinding of the mixture, chromium silicides form. A temperature treatment completes the process. Silicide formation proceeds within the framework of the diffusion of silicon into chromium. In the presence of SiO2 in the mixture, silicide formation occurs also as a result of the reduction of silica by silicon and silicon carbide. The sintering of synthesized composite SiC-chromium silicides powders at a high temperature under a high pressure (T = 2073 K, P = 5 GPa is accompanied by the destruction of cc-SiC particles, the cc/3 transition in silicon carbide and deformation distortions of the lattices of chromium silicides.

Thermoelectric characteristics of Pt-silicide/silicon multi-layer structured p-type silicon

International Nuclear Information System (INIS)

Choi, Wonchul; Jun, Dongseok; Kim, Soojung; Shin, Mincheol; Jang, Moongyu

2015-01-01

Electric and thermoelectric properties of silicide/silicon multi-layer structured devices were investigated with the variation of silicide/silicon heterojunction numbers from 3 to 12 layers. For the fabrication of silicide/silicon multi-layered structure, platinum and silicon layers are repeatedly sputtered on the (100) silicon bulk substrate and rapid thermal annealing is carried out for the silicidation. The manufactured devices show ohmic current–voltage (I–V) characteristics. The Seebeck coefficient of bulk Si is evaluated as 195.8 ± 15.3 μV/K at 300 K, whereas the 12 layered silicide/silicon multi-layer structured device is evaluated as 201.8 ± 9.1 μV/K. As the temperature increases to 400 K, the Seebeck coefficient increases to 237.2 ± 4.7 μV/K and 277.0 ± 1.1 μV/K for bulk and 12 layered devices, respectively. The increase of Seebeck coefficient in multi-layered structure is mainly attributed to the electron filtering effect due to the Schottky barrier at Pt-silicide/silicon interface. At 400 K, the thermal conductivity is reduced by about half of magnitude compared to bulk in multi-layered device which shows the efficient suppression of phonon propagation by using Pt-silicide/silicon hetero-junctions. - Highlights: • Silicide/silicon multi-layer structured is proposed for thermoelectric devices. • Electric and thermoelectric properties with the number of layer are investigated. • An increase of Seebeck coefficient is mainly attributed the Schottky barrier. • Phonon propagation is suppressed with the existence of Schottky barrier. • Thermal conductivity is reduced due to the suppression of phonon propagation

The fabrication of metal silicide nanodot arrays using localized ion implantation

International Nuclear Information System (INIS)

Han, Jin; Kim, Tae-Gon; Min, Byung-Kwon; Lee, Sang Jo

2010-01-01

We propose a process for fabricating nanodot arrays with a pitch size of less than 25 nm. The process consists of localized ion implantation in a metal thin film on a Si wafer using a focused ion beam (FIB), followed by chemical etching. This process utilizes the etching resistivity changes of the ion beam irradiated region that result from metal silicide formation by ion implantation. To control the nanodot diameter, a threshold ion dose model is proposed using the Gaussian distribution of the ion beam intensities. The process is verified by fabricating nanodots with various diameters. The mechanism of etching resistivity is investigated via x-ray photoelectron spectroscopy (XPS) and Auger electron spectroscopy (AES).

Gas cluster ion beam assisted NiPt germano-silicide formation on SiGe

Energy Technology Data Exchange (ETDEWEB)

Ozcan, Ahmet S., E-mail: asozcan@us.ibm.com [IBM Almaden Research Center, 650 Harry Road, San Jose, California 95120 (United States); Lavoie, Christian; Jordan-Sweet, Jean [IBM T. J. Watson Research Center, 1101 Kitchawan Road, Yorktown Heights, New York 10598 (United States); Alptekin, Emre; Zhu, Frank [IBM Semiconductor Research and Development Center, 2070 Route 52, Hopewell Junction, New York 12533 (United States); Leith, Allen; Pfeifer, Brian D.; LaRose, J. D.; Russell, N. M. [TEL Epion Inc., 900 Middlesex Turnpike, Bldg. 6, Billerica, Massachusetts 01821 (United States)

2016-04-21

We report the formation of very uniform and smooth Ni(Pt)Si on epitaxially grown SiGe using Si gas cluster ion beam treatment after metal-rich silicide formation. The gas cluster ion implantation process was optimized to infuse Si into the metal-rich silicide layer and lowered the NiSi nucleation temperature significantly according to in situ X-ray diffraction measurements. This novel method which leads to more uniform films can also be used to control silicide depth in ultra-shallow junctions, especially for high Ge containing devices, where silicidation is problematic as it leads to much rougher interfaces.

NMOS contact resistance reduction with selenium implant into NiPt silicide

Science.gov (United States)

Rao, K. V.; Khaja, F. A.; Ni, C. N.; Muthukrishnan, S.; Darlark, A.; Lei, J.; Peidous, I.; Brand, A.; Henry, T.; Variam, N.; Erokhin, Y.

2012-11-01

A 25% reduction in NMOS contact resistance (Rc) was achieved by Selenium implantation into NiPt silicide film in VIISta Trident high-current single-wafer implanter. The Trident implanter is designed for shallow high-dose implants with high beam currents to maintain high throughput (for low CoO), with improved micro-uniformity and no energy contamination. The integration of Se implant was realized using a test chip dedicated to investigating silicide/junction related electrical properties and testable after silicidation. The silicide module processes were optimized, including the pre-clean (prior to RF PVD NiPt dep) and pre- and post-implant anneals. A 270°C soak anneal was used for RTP1, whereas a msec laser anneal was employed for RTP2 with sufficient process window (800-850°C), while maintaining excellent junction characteristics without Rs degradation.

Technology CAD of silicided Schottky barrier MOSFET for elevated source-drain engineering

International Nuclear Information System (INIS)

Saha, A.R.; Chattopadhyay, S.; Bose, C.; Maiti, C.K.

2005-01-01

Technology CAD has been used to study the performance of a silicided Schottky barrier (SB) MOSFET with gate, source and drain contacts realized with nickel-silicide. Elevated source-drain structures have been used towards the S/D engineering of CMOS devices. A full process-to-device simulation has been employed to predict the performance of sub-micron SB n-MOSFETs for the first time. A model for the diffusion and alloy growth kinetics has been incorporated in SILVACO-ATLAS and ATHENA to explore the processing and design parameter space for the Ni-silicided MOSFETs. The temperature and concentration dependent diffusion model for NiSi have been developed and necessary material parameters for nickel-silicide and epitaxial-Si have been incorporated through the C-interpreter function. Two-dimensional (2D) process-to-device simulations have also been used to study the dc and ac (RF) performance of silicided Schottky barrier (SB) n-MOSFETs. The extracted sheet resistivity, as a function of annealing temperature of the silicided S/D contacts, is found to be lower than the conventional contacts currently in use. It is also shown that the Technology CAD has the full capability to predict the possible dc and ac performance enhancement of a MOSFET with elevated S/D structures. While the simulated dc performance shows a clear enhancement, the RF analyses show no performance degradation in the cut-off frequency/propagation delay and also improve the ac performance due to the incorporation of silicide contacts in the S/D region

Technology CAD of silicided Schottky barrier MOSFET for elevated source-drain engineering

Energy Technology Data Exchange (ETDEWEB)

Saha, A.R. [Department of Electronics and ECE, IIT, Kharagpur 721302 (India)]. E-mail: ars.iitkgp@gmail.com; Chattopadhyay, S. [Department of Electronics and ECE, IIT, Kharagpur 721302 (India); School of Electrical, Electronics and Computer Engineering, University of Newcastle, Newcastle upon Tyne (United Kingdom); Bose, C. [Department of Electronics and Telecommunication Engineering, Jadavpur University, Calcutta 700032 (India); Maiti, C.K. [Department of Electronics and ECE, IIT, Kharagpur 721302 (India)

2005-12-05

Technology CAD has been used to study the performance of a silicided Schottky barrier (SB) MOSFET with gate, source and drain contacts realized with nickel-silicide. Elevated source-drain structures have been used towards the S/D engineering of CMOS devices. A full process-to-device simulation has been employed to predict the performance of sub-micron SB n-MOSFETs for the first time. A model for the diffusion and alloy growth kinetics has been incorporated in SILVACO-ATLAS and ATHENA to explore the processing and design parameter space for the Ni-silicided MOSFETs. The temperature and concentration dependent diffusion model for NiSi have been developed and necessary material parameters for nickel-silicide and epitaxial-Si have been incorporated through the C-interpreter function. Two-dimensional (2D) process-to-device simulations have also been used to study the dc and ac (RF) performance of silicided Schottky barrier (SB) n-MOSFETs. The extracted sheet resistivity, as a function of annealing temperature of the silicided S/D contacts, is found to be lower than the conventional contacts currently in use. It is also shown that the Technology CAD has the full capability to predict the possible dc and ac performance enhancement of a MOSFET with elevated S/D structures. While the simulated dc performance shows a clear enhancement, the RF analyses show no performance degradation in the cut-off frequency/propagation delay and also improve the ac performance due to the incorporation of silicide contacts in the S/D region.

Structural and electronic properties of rare-earth silicide thin films at Si(111)

Energy Technology Data Exchange (ETDEWEB)

Dues, Christof; Schmidt, Wolf Gero; Sanna, Simone [Lehrstuhl fuer Theoretische Physik, Universitaet Paderborn (Germany)

2016-07-01

Rare-earth (RE) silicides thin films on silicon surfaces are currently of high interest. They grow nearly defect-free because of the small lattice mismatch, and exhibit very low Schottky-barriers on n-type silicon. They even give rise to the self-organized formation of RE silicide nanowires on the Si(001) and vicinal surfaces. Depending on the amount of deposited RE atoms, a plethora of reconstructions are observed for the RE silicide. While one monolayer leads to the formation of a 1 x 1-reconstruction, several monolayer thick silicides crystallize in a √(3) x √(3) R30 {sup circle} superstructure. Submonolayer RE deposition leads to different periodicities. In this work we investigate the formation of RE silicides thin films on Si(111) within the density functional theory. The energetically favored adsorption site for RE adatoms is determined calculating the potential energy surface. As prototypical RE, Dysprosium is used. Additional calculations are performed for silicides formed by different RE elements. We calculate structural properties, electronic band structures and compare measured and simulated STM images. We consider different terminations for the 5 x 2 reconstruction occurring in the submonolayer regime and investigate their stability by means of ab initio thermodynamics. The same method is employed to predict the stable silicide structure as a function of the deposited RE atoms.

Silicide/Silicon Heterointerfaces, Reaction Kinetics and Ultra-short Channel Devices

Science.gov (United States)

Tang, Wei

Nickel silicide is one of the electrical contact materials widely used on very large scale integration (VLSI) of Si devices in microelectronic industry. This is because the silicide/silicon interface can be formed in a highly controlled manner to ensure reproducibility of optimal structural and electrical properties of the metal-Si contacts. These advantages can be inherited to Si nanowire (NW) field-effect transistors (FET) device. Due to the technological importance of nickel silicides, fundamental materials science of nickel silicides formation (Ni-Si reaction), especially in nanoscale, has raised wide interest and stimulate new insights and understandings. In this dissertation, in-situ transmission electron microscopy (TEM) in combination with FET device characterization will be demonstrated as useful tools in nano-device fabrication as well as in gaining insights into the process of nickel silicide formation. The shortest transistor channel length (17 nm) fabricated on a vapor-liquid-solid (VLS) grown silicon nanowire (NW) has been demonstrated by controlled reaction with Ni leads on an in-situ transmission electron microscope (TEM) heating stage at a moderate temperature of 400 ºC. NiSi2 is the leading phase, and the silicide-silicon interface is an atomically sharp type-A interface. At such channel lengths, high maximum on-currents of 890 (microA/microm) and a maximum transconductance of 430 (microS/microm) were obtained, which pushes forward the performance of bottom-up Si NW Schottky barrier field-effect transistors (SB-FETs). Through accurate control over the silicidation reaction, we provide a systematic study of channel length dependent carrier transport in a large number of SB-FETs with channel lengths in the range of (17 nm -- 3.6 microm). Our device results corroborate with our transport simulations and reveal a characteristic type of short channel effects in SB-FETs, both in on- and off-state, which is different from that in conventional MOSFETs

Simultaneous determination of uranium and plutonium in dissolver solution of irradiated fuel, using ID-TIMS. IRP-11

International Nuclear Information System (INIS)

Shah, Raju; Sasi Bhushan, K.; Govindan, R.; Alamelu, D.; Khodade, P.S.; Aggarwal, S.K.

2007-01-01

A simple sample preparation and simultaneous analysis method to determine uranium and plutonium from dissolver solution, employing the technique of Isotope Dilution Mass spectrometry has been demonstrated. The method used, co-elusion of Uranium and Plutonium from anion exchanger column after initial elution of major part of uranium in 1:5 HNO 3 in order to reduce the initial U/Pu ratio from 1000 to about 100-200 in the co-eluted fraction. Due to the availability of variable multi-collector system, different Faraday cups were adjusted to collect the different ion intensities corresponding to the different masses, during the simultaneous analysis of Uranium and Plutonium, loaded on Re double filament assembly. 233 U and PR grade Plutonium were used as spikes to determine Uranium and Plutonium from dissolver solution of irradiated fuel from research reactor. The possibility of getting the isotopic composition of uranium from the simultaneous analysis of co-eluted purified fraction of U and Pu from spiked aliquots is also explained. (author)

EPR study of the production of OH radicals in aqueous solutions of uranium irradiated by ultraviolet light

Directory of Open Access Journals (Sweden)

MARKO DAKOVIĆ

2009-06-01

Full Text Available The aim of the study was to establish whether hydroxyl radicals (•OH were produced in UV-irradiated aqueous solutions of uranyl salts. The production of •OH was studied in uranyl acetate and nitrate solutions by an EPR spin trap method over a wide pH range, with variation of the uranium concentrations. The production of •OH in uranyl solutions irradiated with UV was unequivocally demonstrated for the first time using the EPR spin-trapping method. The production of •OH can be connected to speciation of uranium species in aqueous solutions, showing a complex dependence on the solution pH. When compared with the results of radiative de-excitation of excited uranyl (*UO22+ by the quenching of its fluorescence, the present results indicate that the generation of hydroxyl radicals plays a major role in the fluorescence decay of *UO22+. The role of the presence of carbonates and counter ions pertinent to environmental conditions in biological systems on the production of hydroxyl radicals was also assessed in an attempt to reveal the mechanism of *UO22+ de-excitation. Various mechanisms, including •OH production, are inferred but the main point is that the generation of •OH in uranium containing solutions must be considered when assessing uranium toxicity.

Silicide Schottky Contacts to Silicon: Screened Pinning at Defect Levels

Energy Technology Data Exchange (ETDEWEB)

Drummond, T.J.

1999-03-11

Silicide Schottky contacts can be as large as 0.955 eV (E{sub v} + 0.165 eV) on n-type silicon and as large as 1.05 eV (E{sub c} {minus} 0.07 eV) on p-type silicon. Current models of Schottky barrier formation do not provide a satisfactory explanation of occurrence of this wide variation. A model for understanding Schottky contacts via screened pinning at defect levels is presented. In the present paper it is shown that most transition metal silicides are pinned approximately 0.48 eV above the valence band by interstitial Si clusters. Rare earth disilicides pin close to the divacancy acceptor level 0.41 eV below the conduction band edge while high work function silicides of Ir and Pt pin close to the divacancy donor level 0.21 eV above the valence band edge. Selection of a particular defect pinning level depends strongly on the relative positions of the silicide work function and the defect energy level on an absolute energy scale.

133Xe release during post-irradiation annealing of uranium metal in the presence of a constant volume of air Pt. 1

International Nuclear Information System (INIS)

Marei, S.A.; El-Garhy, M.; El-Bayoumy, S.; Muenze, R.; Hladik, O.

1978-01-01

The fractional release of 133 Xe at different temperatures was studied as a function of time in the presence of air during post-irradiation annealing of uranium metal. The relation between the fractional release and tsup(1/2) was found to be irregular. There is an initial step in the annealing curves (at the temperature range of 400-710 deg C) which decreases by increasing temperature and totally disappears at the high temperature of 800-1000 deg C. The initial step was found to be due to the surface oxidation of uranium metal. The other two parts of the release curves are normal for 133 Xe release from uranium metal. Since in this work the irradiation temperature is low ( 133 Xe. (T.G.)

Kinetics of nickel silicide growth in silicon nanowires: From linear to square root growth

International Nuclear Information System (INIS)

Yaish, Y. E.; Beregovsky, M.; Katsman, A.; Cohen, G. M.

2011-01-01

The common practice for nickel silicide formation in silicon nanowires (SiNWs) relies on axial growth of silicide along the wire that is initiated from nickel reservoirs at the source and drain contacts. In the present work the silicide intrusions were studied for various parameters including wire diameter (25-50 nm), annealing time (15-120 s), annealing temperature (300-440 deg. C), and the quality of the initial Ni/Si interface. The silicide formation was investigated by high-resolution scanning electron microscopy, high-resolution transmission electron microscopy (TEM), and atomic force microscopy. The main part of the intrusion formed at 420 deg. C consists of monosilicide NiSi, as was confirmed by energy dispersive spectroscopy STEM, selected area diffraction TEM, and electrical resistance measurements of fully silicided SiNWs. The kinetics of nickel silicide axial growth in the SiNWs was analyzed in the framework of a diffusion model through constrictions. The model calculates the time dependence of the intrusion length, L, and predicts crossover from linear to square root time dependency for different wire parameters, as confirmed by the experimental data.

Investigations of uraniumsilicide-based dispersion fuels for the use of low enrichment uranium (LEU) in research and test reactors

International Nuclear Information System (INIS)

Nazare, S.

1982-07-01

The work presents at the outset, a review of the preparation and properties of uranium silicides (U 3 Si and U 3 Si 2 ) in so far as these are relevant for their use as dispersants in research reactor fuels. The experimental work deals with the preparation and powder metallurgical processing of Al-clad miniature fuel element plates with U 3 Si- und U 3 Si-Al up to U-densities of 6.0 g U/cm 3 . The compatibility of these silicides with the Al-matrix under equilibrium conditions (873 K) and the influence of the reaction on the dimensional stability of the miniplates is described and discussed. (orig.) [de

C.E.R.C.A. contribution to the RERTR program. Status of development, November 1982

International Nuclear Information System (INIS)

Savornin, B.; Fanjas, Y.R.

1983-01-01

For more than 20 years, CERCA has been manufacturing MTR fuel plates and fuel elements. 200.000 fuel plates have thus been produced and irradiated in reactors of 15 countries through the world, including the U.S.A. and Japan. On the way to the use of LEU in research reactors, we have adapted our fuel plate fabrication technology and developed new fuels in order to achieve high uranium densities in the meat. These technological developments have been performed on our own funds. Aluminide (U Al x ), Oxide (U 3 O 8 ) and Silicide (U 3 Si, U 3 Si 2 , U 3 Si Al) fuels have been successfully developed. Upon completion of the aluminide and oxide developments, we have intensified our effort on silicide fuels. After a brief review of the technological results obtained with U Al x and U 3 O 8 materials, silicide fuels results are examined in more detail. The irradiation status of these various fuels is discussed. It has been shown that CERCA is in position to provide on industrial scale UAl x , U 3 O 8 and U 3 Si 2 fuels with uranium densities up to 5 g/cm 3 . New equipment has been installed in MTR workshop in ROMANS for this purpose. For densities between and 74 g U/cm 3 , pilot technology is available. Development is carried out towards industrialization of U 3 Si fuel production. The commercial availability of these fuels depends on their qualification by post irradiation examination. It can be seen that aluminide fuel will be commercially available in 1983, oxide fuel in 1984. As far as silicide fuels are concerned, their availability will depend on the number of tests to be performed and the burn-up levels required by the Safety Commissions to consider the fuel qualified. These requirements may be different from one reactor to the other. This is why the corresponding dates stretch from 1985 to 1988. The goal of technological development has been reached. Densities above 7 g U/cm 3 have been achieved. Assuming silicide fuels reprocessing does not present major

The migration of intra-granular fission gas bubbles in irradiated uranium dioxide

International Nuclear Information System (INIS)

Baker, C.

1977-05-01

The mobility of intragranular fission gas bubbles in uranium dioxide irradiated at 1600-1800 0 C has been studied following isothermal annealing at temperatures below 1600 0 C. The intragranular fission gas bubbles, average diameter approximately 2nm, are virtually immobile at temperatures below 1500 0 C. The bubbles have clean surfaces with no solid fission product contamination and are faceted to the highest observed irradiation temperature of 1800 0 C. This bubble faceting is believed to be a major cause of bubble immobility. In fuel operating below 1500 0 C the predominant mechanism allowing the growth of intragranular bubbles and the subsequent gas release must be the diffusion of dissolved gas atoms rather than the movement of entire intragranular bubbles. (author)

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)

Prediction of barrier inhomogeneities and carrier transport in Ni-silicided Schottky diode

International Nuclear Information System (INIS)

Saha, A.R.; Dimitriu, C.B.; Horsfall, A.B.; Chattopadhyay, S.; Wright, N.G.; O'Neill, A.G.; Maiti, C.K.

2006-01-01

Based on Quantum Mechanical (QM) carrier transport and the effects of interface states, a theoretical model has been developed to predict the anomalous current-voltage (I-V) characteristics of a non-ideal Ni-silicided Schottky diode at low temperatures. Physical parameters such as barrier height, ideality factor, series resistance and effective Richardson constant of a silicided Schottky diode were extracted from forward I-V characteristics and are subsequently used for the simulation of both forward and reverse I-V characteristics using a QM transport model in which the effects of interface state and bias dependent barrier reduction are incorporated. The present analysis indicates that the effects of barrier inhomogeneity caused by incomplete silicide formation at the junction and the interface states may change the conventional current transport process, leading to anomalous forward and reverse I-V characteristics for the Ni-silicided Schottky diode

Reuse of ammonium fluoride generated in the uranium hexafluoride conversion

International Nuclear Information System (INIS)

Silva Neto, J.B.; Carvalho, E.F. Urano de; Durazzo, M.; Riella, H.G

2010-01-01

The Nuclear Fuel Centre of IPEN / CNEN - SP develops and manufactures dispersion fuel with high uranium concentration to meet the demand of the IEA-R1 reactor and future research reactors planned to be constructed in Brazil. The fuel uses uranium silicide (U 3 Si 2 ) dispersed in aluminum. For producing the fuel, the processes for uranium hexafluoride (UF 6 ) conversion consist in obtaining U 3 Si 2 and / or U 3 O 8 through the preparation of intermediate compounds, among them ammonium uranyl carbonate - AUC, ammonium diuranate - DUA and uranium tetrafluoride - UF 4 . This work describes a procedure for preparing uranium tetrafluoride by a dry route using as raw material the filtrate generated when producing routinely ammonium uranyl carbonate. The filtrate consists primarily of a solution containing high concentrations of ammonium (NH 4 + ), fluoride (F - ), carbonate (CO 3 -- ) and low concentrations of uranium. The procedure is basically the recovery of NH 4 F and uranium, as UF 4 , through the crystallization of ammonium bifluoride (NH 4 HF 2 ) and, in a later step, the addition of UO 2 , occurring fluoridation and decomposition. The UF 4 obtained is further diluted in the UF 4 produced routinely at IPEN / CNEN-SP by a wet route process. (author)

Analysis Of Core Management For The Transition Cores Of RSG-GAS Reactor To Full-Silicide Core

International Nuclear Information System (INIS)

Malem Sembiring, Tagor; Suparlina, Lily; Tukiran

2001-01-01

The core conversion of RSG-GAS reactor from oxide to silicide core with meat density of 2.96 g U/cc is still doing. At the end of 2000, the reactor has been operated for 3 transition cores which is the mixed core of oxide-silicide. Based on previous work, the calculated core parameter for the cores were obtained and it is needed 10 transition cores to achieve a full-silicide core. The objective of this work is to acquire the effect of the increment of the number of silicide fuel on the core parameters such as excess reactivity and shutdown margin. The measurement of the core parameters was carried out using the method of compensation of couple control rods. The experiment shows that the excess reactivity trends lower with the increment of the number of silicide fuel in the core. However, the shutdown margin is not change with the increment of the number of silicide fuel. Therefore, the transition cores can be operated safety to a full-silicide core

Far-infrared spectroscopy of thermally annealed tungsten silicide films

International Nuclear Information System (INIS)

Amiotti, M.; Borghesi, A.; Guizzetti, G.; Nava, F.; Santoro, G.

1991-01-01

The far-infrared transmittance spectrum of tungsten silicide has been observed for the first time. WSi 2 polycrystalline films were prepared by coevaporation and chemical-vapour deposition on silicon wafers, and subsequently thermally annealed at different temperatures. The observed structures are interpreted, on the basis of the symmetry properties of the crystal, such as infrared-active vibrational modes. Moreover, the marked lineshape dependence on annealing temperature enables this technique to analyse the formation of the solid silicide phases

Texture in thin film silicides and germanides: A review

International Nuclear Information System (INIS)

De Schutter, B.; De Keyser, K.; Detavernier, C.; Lavoie, C.

2016-01-01

Silicides and germanides are compounds consisting of a metal and silicon or germanium. In the microelectronics industry, silicides are the material of choice for contacting silicon based devices (over the years, CoSi_2, C54-TiSi_2, and NiSi have been adopted), while germanides are considered as a top candidate for contacting future germanium based electronics. Since also strain engineering through the use of Si_1_−_xGe_x in the source/drain/gate regions of MOSFET devices is an important technique for improving device characteristics in modern Si-based microelectronics industry, a profound understanding of the formation of silicide/germanide contacts to silicon and germanium is of utmost importance. The crystallographic texture of these films, which is defined as the statistical distribution of the orientation of the grains in the film, has been the subject of scientific studies since the 1970s. Different types of texture like epitaxy, axiotaxy, fiber, or combinations thereof have been observed in such films. In recent years, it has become increasingly clear that film texture can have a profound influence on the formation and stability of silicide/germanide contacts, as it controls the type and orientation of grain boundaries (affecting diffusion and agglomeration) and the interface energy (affecting nucleation during the solid-state reaction). Furthermore, the texture also has an impact on the electrical characteristics of the contact, as the orientation and size of individual grains influences functional properties such as contact resistance and sheet resistance and will induce local variations in strain and Schottky barrier height. This review aims to give a comprehensive overview of the scientific work that has been published in the field of texture studies on thin film silicide/germanide contacts.

Texture in thin film silicides and germanides: A review

Science.gov (United States)

De Schutter, B.; De Keyser, K.; Lavoie, C.; Detavernier, C.

2016-09-01

Silicides and germanides are compounds consisting of a metal and silicon or germanium. In the microelectronics industry, silicides are the material of choice for contacting silicon based devices (over the years, CoSi2, C54-TiSi2, and NiSi have been adopted), while germanides are considered as a top candidate for contacting future germanium based electronics. Since also strain engineering through the use of Si1-xGex in the source/drain/gate regions of MOSFET devices is an important technique for improving device characteristics in modern Si-based microelectronics industry, a profound understanding of the formation of silicide/germanide contacts to silicon and germanium is of utmost importance. The crystallographic texture of these films, which is defined as the statistical distribution of the orientation of the grains in the film, has been the subject of scientific studies since the 1970s. Different types of texture like epitaxy, axiotaxy, fiber, or combinations thereof have been observed in such films. In recent years, it has become increasingly clear that film texture can have a profound influence on the formation and stability of silicide/germanide contacts, as it controls the type and orientation of grain boundaries (affecting diffusion and agglomeration) and the interface energy (affecting nucleation during the solid-state reaction). Furthermore, the texture also has an impact on the electrical characteristics of the contact, as the orientation and size of individual grains influences functional properties such as contact resistance and sheet resistance and will induce local variations in strain and Schottky barrier height. This review aims to give a comprehensive overview of the scientific work that has been published in the field of texture studies on thin film silicide/germanide contacts.

Texture in thin film silicides and germanides: A review

Energy Technology Data Exchange (ETDEWEB)

De Schutter, B., E-mail: bob.deschutter@ugent.be; De Keyser, K.; Detavernier, C. [Department of Solid State Sciences, Ghent University, Ghent (Belgium); Lavoie, C. [IBM Research Division, T.J. Watson Research Center, P.O. Box 218, Yorktown Heights, New York 10598 (United States)

2016-09-15

Silicides and germanides are compounds consisting of a metal and silicon or germanium. In the microelectronics industry, silicides are the material of choice for contacting silicon based devices (over the years, CoSi{sub 2}, C54-TiSi{sub 2}, and NiSi have been adopted), while germanides are considered as a top candidate for contacting future germanium based electronics. Since also strain engineering through the use of Si{sub 1−x}Ge{sub x} in the source/drain/gate regions of MOSFET devices is an important technique for improving device characteristics in modern Si-based microelectronics industry, a profound understanding of the formation of silicide/germanide contacts to silicon and germanium is of utmost importance. The crystallographic texture of these films, which is defined as the statistical distribution of the orientation of the grains in the film, has been the subject of scientific studies since the 1970s. Different types of texture like epitaxy, axiotaxy, fiber, or combinations thereof have been observed in such films. In recent years, it has become increasingly clear that film texture can have a profound influence on the formation and stability of silicide/germanide contacts, as it controls the type and orientation of grain boundaries (affecting diffusion and agglomeration) and the interface energy (affecting nucleation during the solid-state reaction). Furthermore, the texture also has an impact on the electrical characteristics of the contact, as the orientation and size of individual grains influences functional properties such as contact resistance and sheet resistance and will induce local variations in strain and Schottky barrier height. This review aims to give a comprehensive overview of the scientific work that has been published in the field of texture studies on thin film silicide/germanide contacts.

Development of Silicide Coating on Molybdenum Alloy Cladding

International Nuclear Information System (INIS)

Lim, Woojin; Ryu, Ho Jin

2015-01-01

The molybdenum alloy is considered as one of the accident tolerant fuel (ATF) cladding materials due to its high temperature mechanical properties. However, molybdenum has a weak oxidation resistance at elevated temperatures. To modify the oxidation resistance of molybdenum cladding, silicide coating on the cladding is considered. Molybdenum silicide layers are oxidized to SiO 2 in an oxidation atmosphere. The SiO 2 protective layer isolates the substrate from the oxidizing atmosphere. Pack cementation deposition technique is widely adopted for silicide coating for molybdenum alloys due to its simple procedure, homogeneous coating quality and chemical compatibility. In this study, the pack cementation method was conducted to develop molybdenum silicide layers on molybdenum alloys. It was found that the Mo 3 Si layer was deposited on substrate instead of MoSi 2 because of short holding time. It means that through the extension of holding time, MoSi 2 layer can be formed on molybdenum substrate to enhance the oxidation resistance of molybdenum. The accident tolerant fuel (ATF) concept is to delay the process following an accident by reducing the oxidation rate at high temperatures and to delay swelling and rupture of fuel claddings. The current research for Atf can be categorized into three groups: First, modification of existing zirconium-based alloy cladding by improving the high temperature oxidation resistance and strength. Second, replacing Zirconium based alloys with alternative metallic materials such as refractory elements with high temperature oxidation resistance and strength. Third, designing alternative fuel structures using ceramic and composite systems

The shielded chemistry suite at BNL and its use in the study of the irradiated uranium-water reaction

International Nuclear Information System (INIS)

Bannister, G.H.; Tyfield, S.P.

1977-11-01

A versatile shielded facility for chemical studies is now in use at Berkeley Nuclear Laboratories. The suite was fully commissioned in 1976 and first utilised to study the aqueous corrosion of irradiated reactor grade uranium. The suite consists of three adjacent interconnected cells; a balance cell, a preparation cell and an experimental cell. The development, design and layout of the facilities are described, as are the details of the equipment installed to undertake a uranium corrosion study. The preparation cell is used for procedures that could results in high levels of contamination, such as cutting of uranium specimens. The uranium samples are corroded in sealed steel capsules installed in the experimental cell. The capsules are heated either isothermally or through a transient and the reaction monitored with a pressure transducer which records the rate of hydrogen evolution from the uranium-water reaction. The experimental results will be reported elsewhere. (author)

Oxidation behavior of molybdenum silicides and their composites

International Nuclear Information System (INIS)

Natesan, K.; Deevi, S. C.

2000-01-01

A key materials issue associated with the future of high-temperature structural silicides is the resistance of these materials to oxidation at low temperatures. Oxidation tests were conducted on Mo-based silicides over a wide temperature range to evaluate the effects of alloy composition and temperature on the protective scaling characteristics and testing regime for the materials. The study included Mo 5 Si 3 alloys that contained several concentrations of B. In addition, oxidation characteristics of MoSi 2 -Si 3 N 4 composites that contained 20--80 vol.% Si 3 N 4 were evaluated at 500--1,400 C

Gas Generation from K East Basin Sludges and Irradiated Metallic Uranium Fuel Particles Series III Testing

International Nuclear Information System (INIS)

Schmidt, Andrew J.; Delegard, Calvin H.; Bryan, Samuel A.; Elmore, Monte R.; Sell, Rachel L.; Silvers, Kurt L.; Gano, Susan R.; Thornton, Brenda M.

2003-01-01

The path forward for managing of Hanford K Basin sludge calls for it to be packaged, shipped, and stored at T Plant until final processing at a future date. An important consideration for the design and cost of retrieval, transportation, and storage systems is the potential for heat and gas generation through oxidation reactions between uranium metal and water. This report, the third in a series (Series III), describes work performed at the Pacific Northwest National Laboratory (PNNL) to assess corrosion and gas generation from irradiated metallic uranium particles (fuel particles) with and without K Basin sludge addition. The testing described in this report consisted of 12 tests. In 10 of the tests, 4.3 to 26.4 g of fuel particles of selected size distribution were placed into 60- or 800-ml reaction vessels with 0 to 100 g settled sludge. In another test, a single 3.72-g fuel fragment (i.e., 7150-mm particle) was placed in a 60 ml reaction vessel with no added sludge. The twelfth test contained only sludge. The fuel particles were prepared by crushing archived coupons (samples) from an irradiated metallic uranium fuel element. After loading the sludge materials (whether fuel particles, mixtures of fuel particles and sludge, or sludge-only) into reaction vessels, the solids were covered with an excess of K Basin water, the vessels closed and connected to a gas measurement manifold, and the vessels back-flushed with inert neon cover gas. The vessels were then heated to a constant temperature. The gas pressures and temperatures were monitored continuously from the times the vessels were purged. Gas samples were collected at various times during the tests, and the samples analyzed by mass spectrometry. Data on the reaction rates of uranium metal fuel particles with water as a function of temperature and particle size were generated. The data were compared with published studies on metallic uranium corrosion kinetics. The effects of an intimate overlying sludge layer

Irradiation performance of uranium-molybdenum alloy dispersion fuels; Desempenho sob irradiacao de elementos combustiveis do tipo U-Mo

Energy Technology Data Exchange (ETDEWEB)

Almeida, Cirila Tacconi de

2005-07-01

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

Method of preparation of uranium nitride

Science.gov (United States)

Kiplinger, Jaqueline Loetsch; Thomson, Robert Kenneth James

2013-07-09

Method for producing terminal uranium nitride complexes comprising providing a suitable starting material comprising uranium; oxidizing the starting material with a suitable oxidant to produce one or more uranium(IV)-azide complexes; and, sufficiently irradiating the uranium(IV)-azide complexes to produce the terminal uranium nitride complexes.

Microstructure and mechanical properties of metal/oxide and metal/silicide interfaces

International Nuclear Information System (INIS)

Shaw, L.; Miracle, D.; Abbaschian, R.

1995-01-01

Fracture energies of Al 2 O 3 /Nb interfaces and MoSi 2 /Nb interfaces with and without Al 2 O 3 coating were measured using sandwich-type chevron-notched specimens. The relations between the mechanical properties, microstructures, types of bonds at the interface and processing routes were explored. The fracture energy of the Al 2 O 3 /Nb interface was determined to be 9 J/m 2 and changed to 16 J/m 2 when Nb was pre-oxidized before the formation of the Al 2 O 3 /Nb interface. The fracture energy of the MoSi 2 /Nb interface could not be determined directly because of the formation of the interfacial compounds. However, the fracture energy at the MoSi 2 /Nb interfacial region was found to depend on the interfacial bond strength, roughness of interfaces and microstructure of interfacial compounds. The interfacial fracture energies of Al 2 O 3 with silicides, MoSi 2 , Nb 5 Si 3 , or (Nb, Mo)Si 2 were estimated to be about 16 J/m 2 , while the interfacial fracture energies between two silicides or between Nb and a silicide were larger than 34 J/m 2 . The measured fracture energies between two silicides or between Nb and a silicide were larger than 34 J/m 2 . The measured fracture energies of the various interfaces are discussed in terms of the interfacial microstructures and types of bonds at the interfaces

Evaluation of In-Core Fuel Management for the Transition Cores of RSG-GAS Reactor to Full-Silicide Core

International Nuclear Information System (INIS)

S, Tukiran; MS, Tagor; P, Surian

2003-01-01

The core conversion of RSG-GAS reactor from oxide to silicide core with meat density of 2.96 gU/cc has been done. The core-of RSG-GAS reactor has been operated full core of silicide fuels which is started with the mixed core of oxide-silicide start from core 36. Based on previous work, the calculated core parameter for the cores were obtained and it is needed 9 transition cores (core 36 - 44) to achieve a full-silicide core (core 45). The objective of this work is to acquire the effect of the increment of the number of silicide fuel on the core parameters. Conversion core was achieved by transition cores mixed oxide-silicide fuels. Each transition core is calculated and measured core parameter such as, excess reactivity and shutdown margin. Calculation done by Batan-EQUIL-2D code and measurement of the core parameters was carried out using the method of compensation of couple control rods. The results of calculation and experiment shows that the excess reactivity trends lower with the increment of the number of silicide fuel in the core. However, the shutdown margin is not change with the increment of the number of silicide fuel. Therefore, the transition cores can be operated safely to a full-silicide core

Test calculations of photoneutrons emission from surface of uranium sphere irradiated by 28 MeV electrons

International Nuclear Information System (INIS)

Blokhin, A.I.; Degtyarev, I.I.

2002-01-01

In this paper the results of physical verification for the BOFOD photonuclear data files are reported, available for the uranium isotopes U 235 , U 238 . These results were compared with calculated data by the parameterization driven model of photonuclear reaction and experimental data. Experimental data of photoneutron yields from surface of uranium sphere irradiated by 28 MeV electrons are used for a verification. Both calculations have been carried out with the RTS and T general purpose Monte Carlo code with detailed electron-photon-nucleon transport simulation using the ENDF/B-VI and EPDL evaluated data libraries

Thermal stability of Ni-Pt-Ta alloy silicides on epi-Si{sub 1-x}C{sub x}

Energy Technology Data Exchange (ETDEWEB)

Yoo, Jung-Ho; Chang, Hyun-Jin [Department of Ceramic Engineering, Yonsei University, Seoul 120-749 (Korea, Republic of); Min, Byoung-Gi [Department of Ceramic Engineering, Yonsei University, Seoul 120-749 (Korea, Republic of); Jusung Engineering Co., Ltd., 49, Neungpyeong-ri, Opo-eup, Gwangju-Si, Kyunggi-do 464-892 (Korea, Republic of); Ko, Dae-Hong [Department of Ceramic Engineering, Yonsei University, Seoul 120-749 (Korea, Republic of)], E-mail: dhko@yonsei.ac.kr; Cho, Mann-Ho [Institute of Physics and Applied Physics, Yonsei University, Seoul 120-749 (Korea, Republic of); Sohn, Hyunchul [Department of Ceramic Engineering, Yonsei University, Seoul 120-749 (Korea, Republic of); Lee, Tae-Wan [Jusung Engineering Co., Ltd., 49, Neungpyeong-ri, Opo-eup, Gwangju-Si, Kyunggi-do 464-892 (Korea, Republic of)

2008-12-05

We investigated the silicide formation in Ni/epi-Si{sub 1-x}C{sub x} systems. Ni-Pt and Ni-Pt-Ta films were deposited on epi-Si{sub 1-x}C{sub x}/Si substrates by DC magnetron sputtering and processed at various temperatures. The sheet resistance of the silicide from the Ni alloy/epi-Si{sub 1-x}C{sub x} systems was maintained at low values compared to that from Ni/Si systems. By TEM and EDS analyses, we confirmed the presence of a Pt alloy layer at the top of the Ni-silicide layer. The stability of the silicide layer in the Ni alloy/epi-Si{sub 1-x}C{sub x} system is explained by not only the Pt rich layer on the top of the Ni-silicide layer, but also by the presence of a small amount of Pt in the Ni-silicide layer or at the grain boundaries. And both the thermal stability and the morphology of silicide were greatly improved by the addition of Ta in Ni-Pt films.

X-ray spectroscopy study of electronic structure of laser-irradiated Au nanoparticles in a silica film

International Nuclear Information System (INIS)

Jonnard, P.; Bercegol, H.; Lamaignere, L.; Morreeuw, J.-P.; Rullier, J.-L.; Cottancin, E.; Pellarin, M.

2005-01-01

The electronic structure of gold nanoparticles embedded in a silica film is studied, both before and after irradiation at 355 nm by a laser. The Au 5d occupied valence states are observed by x-ray emission spectroscopy. They show that before irradiation the gold atoms are in metallic states within the nanoparticles. After irradiation with a fluence of 0.5 J/cm 2 , it is found that gold valence states are close to those of a metal-poor gold silicide; thanks to a comparison of the experimental Au 5d states with the calculated ones for gold silicides using the density-functional theory. The formation of such a compound is driven by the diffusion of the gold atoms into the silica film upon the laser irradiation. At higher fluence, 1 J/cm 2 , we find a higher percentage of metallic gold that could be attributed to annealing in the silica matrix

Preparation of uranium-230 as a new uranium tracer

International Nuclear Information System (INIS)

Hashimoto, T.; Kido, K.; Sotobayashi, T.

1977-01-01

A uranium isotope, 230 U(T=20.8 d), was produced from the 231 Pa(γ,n) 230 Pa→viaβ - decay 230 U process with a bremsstrahlung irradiation on a protactinium target. After standing for about one month to obtain a maximal growth of 230 U, the uranium was chemically purified, applying an ion-exchange method. The purity of the 230 U obtained was examined with alpha spectrometry and an intrinsic alpha peak due to 230 U as a new uranium tracer in an alpha spectrometric analysis of uranium isotopes is described. (author)

Exploitation of a self-limiting process for reproducible formation of ultrathin Ni1-xPtx silicide films

International Nuclear Information System (INIS)

Zhang Zhen; Zhu Yu; Rossnagel, Steve; Murray, Conal; Jordan-Sweet, Jean; Yang, Bin; Gaudet, Simon; Desjardins, Patrick; Kellock, Andrew J.; Ozcan, Ahmet; Zhang Shili; Lavoie, Christian

2010-01-01

This letter reports on a process scheme to obtain highly reproducible Ni 1-x Pt x silicide films of 3-6 nm thickness formed on a Si(100) substrate. Such ultrathin silicide films are readily attained by sputter deposition of metal films, metal stripping in wet chemicals, and final silicidation by rapid thermal processing. This process sequence warrants an invariant amount of metal intermixed with Si in the substrate surface region independent of the initial metal thickness, thereby leading to a self-limiting formation of ultrathin silicide films. The crystallographic structure, thickness, uniformity, and morphological stability of the final silicide films depend sensitively on the initial Pt fraction.

Development of uranium metal targets for 99Mo production

International Nuclear Information System (INIS)

Wiencek, T.C.; Hofman, G.L.

1993-10-01

A substantial amount of high enriched uranium (HEU) is used for the production of medical-grade 99 Mo. Promising methods of producing irradiation targets are being developed and may lead to the reduction or elimination of this HEU use. To substitute low enriched uranium (LEU) for HEU in the production of 99 Mo, the target material may be changed to uranium metal foil. Methods of fabrication are being developed to simplify assembly and disassembly of the targets. Removal of the uranium foil after irradiation without dissolution of the cladding is a primary goal in order to reduce the amount of liquid radioactive waste material produced in the process. Proof-of-concept targets have been fabricated. Destructive testing indicates that acceptable contact between the uranium foil and the cladding can be achieved. Thermal annealing tests, which simulate the cladding/uranium diffusion conditions during irradiation, are underway. Plans are being made to irradiate test targets

Further data of silicide fuel for the LEU conversion of JMTR

International Nuclear Information System (INIS)

Saito, M.; Futamura, Y.; Nakata, H.; Ando, H.; Sakurai, F.; Ooka, N.; Sakakura, A.; Ugajin, M.; Shirai, E.

1990-01-01

Silicide fuel data for the safety assessment of the JMTR LEU fuel conversion are being measured. The data include fission product release, thermal properties, behaviour under accident conditions, and metallurgical characteristics. The methods used in the experiments are discussed. Results of fission products release at high temperature are described. The release of iodine from the silicide fuel is considerably lower than for U-Al alloy fuel

Method for the chemical reprocessing of irradiated nuclear fuels, in particular nuclear fuels containing uranium

International Nuclear Information System (INIS)

Koch, G.

1976-01-01

In the chemical processing of irradiated uranium-containing nuclear fuels which are hydrolyzed with aqueous nitric acid, a suggestion is made to use as quaternary ammonium nitrate trialkyl-methyl ammonium nitrates as extracting agent, in which the sum of C atoms is greater than 16. In the illustrated examples, tricaprylmethylammonium nitrate, trilaurylmethylammonium nitrate and tridecylmethylammonium nitrate are named. (HPH/LH) [de

Improvement of Silicide Coating Method as Diffusion Barrier for U-Mo Dispersion Fuel

Energy Technology Data Exchange (ETDEWEB)

Nam, Ji Min; Kim, Sunghwan; Lee, Kyu Hong; Park, Jong Man [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-10-15

The excessive interaction between the U-Mo alloys and their surrounding Al matrix lead to excessive local swelling called 'pillowing'. For this reason, KAERI suggested several remedies such as alloying U-Mo with Ti, or Al matrix with Si. In addition, silicide or nitride coatings on the surface of U-Mo particles have also been proposed to hinder the growth of the interaction layer. In this study, centrifugally atomized U-Mo-Ti alloy powders were coated with silicide layers. The coating process was improved when compared to the previous coating in terms of the ball milling and heat treatment conditions. Subsequently, silicide coated U-Mo-Ti powders and pure aluminum powders were mixed and made into a compact for the annealing test. The compacts were annealed at 550 .deg. C for 2hr, and characterized using scanning electron microscopy (SEM) and energy dispersive x-ray spectroscopy (EDS). 1. Uniform, homogeneous, thickness controllable silicide layers were successfully coated on the surface of U-7wt%Mo-1wt%Ti powders. 2. U{sub 3}Si, U{sub 3}Si{sub 2} silicide layers formed on the surface of U-7wt%Mo-1wt%Ti powders, and were identified by XRD and EDS analyses.

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

International Nuclear Information System (INIS)

Rest, J.; Totev, T.

2007-01-01

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

Formation of silicides in a cavity applicator microwave system

International Nuclear Information System (INIS)

Thompson, D.C.; Kim, H.C.; Alford, T.L.; Mayer, J.W.

2003-01-01

Metal silicides of nickel and cobalt are formed in a cavity applicator microwave system with a magnetron power of 1200 W and a frequency of 2.45 GHz. X-ray diffraction, Rutherford backscattering spectrometry, and four-point-probe measurements are used to identify the silicide phase present and layer thicknesses. Additional processing confirmed that the products attained from heating by microwaves do not differ appreciably from those attained in heating by thermal processes. Materials properties are used to explain microwave power absorption and demonstrate how to tailor a robust process in which thin film reactions can be attained and specific products isolated

Method and apparatus for separating uranium isotopes

International Nuclear Information System (INIS)

Bernstein, E.R.

1977-01-01

A uranium compound in the solid phase (uranium borohydride four) is subjected to radiation of a first predetermined frequency that excites the uranium-235 isotope-bearing molecules but not the uranium-238 isotope-bearing molecules. The compound is simultaneously subjected to radiation of a second predetermined frequency which causes the excited uranium-235 isotope-bearing molecules to chemically decompose but which does not affect the uranium-238 isotope-bearing molecules. Sufficient heat is then applied to the irradiated compound in the solid phase to vaporize the non-decomposed uranium-238 isotope-bearing molecules but not the decomposed uranium-235 isotope-bearing molecules, thereby physically separating the uranium-235 isotope-bearing molecules from the uranium-238 isotope-bearing molecules. The uranium compound sample in the solid phase is deposited or grown in an elongated tube supported within a dewar vessel having a clear optical path tail section surrounded by a coolant. Two sources of radiation are focused on the uranium compound sample. A heating element is attached to the elongated tube to vaporize the irradiated compound

Comparison Between Calculated and Measured Cross Section Changes in Natural Uranium Irradiated in NRX

Energy Technology Data Exchange (ETDEWEB)

Ahlstroem, P E

1961-03-15

It is desirable to obtain an experimental check of the reliability of the methods currently used to determine reactivity changes in a reactor and, with a view to meeting this requirement to some extent, a preliminary comparison has been made between calculated and measured cross-section changes in rods of natural uranium irradiated in NRX. The measurements were made at Harwell in the GLEEP reactor and a description has been given by, inter alia, Ward and Craig. The theory of the calculations, which is briefly described in this report, has been indicated by Littler. The investigation showed that the methods for calculating burn up used at present provides a good illustration of the long-term variations in isotope contents. A satisfactory agreement is obtained with experimental results when calculating apparent cross-section changes in uranium rods due to irradiation if the fission cross- section for {sup 239}Pu is set to 780 b. This is 34 b higher than the figure quoted in BNL - 325 (1958). However, in order to get a good idea as to whether the calculated long-term variations in reactivity really correspond to reality, it is necessary to make further investigations. For this reason the results quoted in this report should be regarded as preliminary.

Technical report: technical development on the silicide plate-type fuel experiment at nuclear safety research reactor

International Nuclear Information System (INIS)

Yanagisawa, Kazuaki; Soyama, Kazuhiko; Ichikawa, Hiroki

1991-08-01

According to a reduction of fuel enrichment from 45 w/o 235 U to 20 w/o, an aluminide plate-type fuel used currently in the domestic research and material testing reactors will be replaced by a silicide plate-type one. One of the major concern arisen from this alternation is to understand the fuel behavior under simulated reactivity initiated accident (RIA) conditions, this is strongly necessary from the safety and licensing point of view. The in-core RIA experiments are, therefore, carried out at Nuclear Safety Research Reactor (NSRR) in Japan Atomic Energy Research Institute (JAERI). The silicide plate-type fuel consisted of the ternary alloy of U-Al-Si as a meat with uranium density up to 4.8 g/cm 3 having thickness by 0.51 mm and the binary alloy of Al-3%Mg as a cladding by thickness of 0.38 mm. Comparison of the physical properties of this metallic plate fuel with the UO 2 -zircaloy fuel rod used conventionally in commercial light water reactors shows that the heat conductivity of the former is of the order of about 13 times greater than the latter, however the melting temperature is only one-half (1570degC). Prior to in-core RIA experiments, there were some difficulties lay in our technical path. This report summarized the technical achievements obtained through our four years work. (J.P.N.)

Nanoscale investigation of the interface situation of plated nickel and thermally formed nickel silicide for silicon solar cell metallization

Energy Technology Data Exchange (ETDEWEB)

Mondon, A., E-mail: andrew.mondon@ise.fraunhofer.de [Fraunhofer ISE, Heidenhofst. 2, D-79110 Freiburg (Germany); Wang, D. [Karlsruhe Nano Micro Facility (KNMF), H.-von-Helmholz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Zuschlag, A. [Universität Konstanz FB Physik, Jacob-Burckhardt-Str. 27, D-78464 Konstanz (Germany); Bartsch, J.; Glatthaar, M.; Glunz, S.W. [Fraunhofer ISE, Heidenhofst. 2, D-79110 Freiburg (Germany)

2014-12-30

Highlights: • Adhesion of metallization of fully plated nickel–copper contacts on silicon solar cells can be achieved by formation of nickel silicide at the cost of degraded cell performance. • Understanding of silicide growth mechanisms and controlled growth may lead to high performance together with excellent adhesion. • Silicide formation is well known from CMOS production from PVD-Ni on flat surfaces. Yet the deposition methods and therefore layer characteristics and the surface topography are different for plated metallization. • TEM analysis is performed for differently processed samples. • A nickel silicide growth model is created for plated Ni on textured silicon solar cells. - Abstract: In the context of nickel silicide formation from plated nickel layers for solar cell metallization, there are several open questions regarding contact adhesion and electrical properties. Nanoscale characterization by transmission electron microscopy has been employed to support these investigations. Interfacial oxides and silicide phases were investigated on differently prepared samples by different analytical methods associated with transmission electron microscopy analysis. Processing variations included the pre-treatment of samples before nickel plating, the used plating solution and the thermal budget for the nickel–silicon solid-state reaction. It was shown that interface oxides of only few nm thickness on both silicon and nickel silicide are present on the samples, depending on the chosen process sequence, which have been shown to play an important role in adhesion of nickel on silicide in an earlier publication. From sample pretreatment variations, conclusions about the role of an interfacial oxide in silicide formation and its influence on phase formation were drawn. Such an oxide layer hinders silicide formation except for pinhole sites. This reduces the availability of Ni and causes a silicide with low Ni content to form. Without an interfacial oxide

Acidic aqueous uranium electrodeposition for target fabrication

International Nuclear Information System (INIS)

Saliba-Silva, A.M.; Oliveira, E.T.; Garcia, R.H.L.; Durazzo, M.

2013-01-01

Direct irradiation of targets inside nuclear research or multiple purpose reactors is a common route to produce 99 Mo- 99m Tc radioisotopes. The electroplating of low enriched uranium over nickel substrate might be a potential alternative to produce targets of 235 U. The electrochemistry of uranium at low temperature might be beneficial for an alternative route to produce 99 Mo irradiation LEU targets. Electrodeposition of uranium can be made using ionic and aqueous solutions producing uranium oxide deposits. The performance of uranium electrodeposition is relatively low because a big competition with H 2 evolution happens inside the window of electrochemical reduction potential. This work explores possibilities of electroplating uranium as UO 2 2+ (Uranium-VI) in order to achieve electroplating uranium in a sufficient amount to be commercially irradiated in the future Brazilian RMB reactor. Electroplated nickel substrate was followed by cathodic current electrodeposition from aqueous UO 2 (NO 3 ) 2 solution. EIS tests and modeling showed that a film formed differently in the three tested cathodic potentials. At the lower level, (-1.8V) there was an indication of a double film formation, one overlaying the other with ionic mass diffusion impaired at the interface with nickel substrate as showed by the relatively lower admittance of Warburg component. (author)

Analysis of impurity effect on Silicide fuels of the RSG-GAS core

International Nuclear Information System (INIS)

Tukiran-Surbakti

2003-01-01

Simulation of impurity effect on silicide fuel of the RSG-GAS core has been done. The aim of this research is to know impurity effect of the U-234 and U-236 isotopes in the silicide fuels on the core criticality. The silicide fuels of 250 g U loading and 19.75 of enrichment is used in this simulation. Cross section constant of fuels and non-structure material of core are generated by WIMSD/4 computer code, meanwhile impurity concentration was arranged from 0.01% to 2%. From the result of analysis can be concluded that the isotopes impurity in the fuels could make trouble in the core and the core can not be operated at critical after a half of its cycle length (350 MW D)

The influence of alloying on the phase formation sequence of ultra-thin nickel silicide films and on the inheritance of texture

Science.gov (United States)

Geenen, F. A.; Solano, E.; Jordan-Sweet, J.; Lavoie, C.; Mocuta, C.; Detavernier, C.

2018-05-01

The controlled formation of silicide materials is an ongoing challenge to facilitate the electrical contact of Si-based transistors. Due to the ongoing miniaturisation of the transistor, the silicide is trending to ever-thinner thickness's. The corresponding increase in surface-to-volume ratio emphasises the importance of low-energetic interfaces. Intriguingly, the thickness reduction of nickel silicides results in an abrupt change in phase sequence. This paper investigates the sequence of the silicides phases and their preferential orientation with respect to the Si(001) substrate, for both "thin" (i.e., 9 nm) and "ultra-thin" (i.e., 3 nm) Ni films. Furthermore, as the addition of ternary elements is often considered in order to tailor the silicides' properties, additives of Al, Co, and Pt are also included in this study. Our results show that the first silicide formed is epitaxial θ-Ni2Si, regardless of initial thickness or alloyed composition. The transformations towards subsequent silicides are changed through the additive elements, which can be understood through solubility arguments and classical nucleation theory. The crystalline alignment of the formed silicides with the substrate significantly differs through alloying. The observed textures of sequential silicides could be linked through texture inheritance. Our study illustrates the nucleation of a new phase drive to reduce the interfacial energy at the silicide-substrate interface as well as at the interface with the silicide which is being consumed for these sub-10 nm thin films.

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

International Nuclear Information System (INIS)

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

1983-01-01

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

Nanoscale investigation of the interface situation of plated nickel and thermally formed nickel silicide for silicon solar cell metallization

Science.gov (United States)

Mondon, A.; Wang, D.; Zuschlag, A.; Bartsch, J.; Glatthaar, M.; Glunz, S. W.

2014-12-01

In the context of nickel silicide formation from plated nickel layers for solar cell metallization, there are several open questions regarding contact adhesion and electrical properties. Nanoscale characterization by transmission electron microscopy has been employed to support these investigations. Interfacial oxides and silicide phases were investigated on differently prepared samples by different analytical methods associated with transmission electron microscopy analysis. Processing variations included the pre-treatment of samples before nickel plating, the used plating solution and the thermal budget for the nickel-silicon solid-state reaction. It was shown that interface oxides of only few nm thickness on both silicon and nickel silicide are present on the samples, depending on the chosen process sequence, which have been shown to play an important role in adhesion of nickel on silicide in an earlier publication. From sample pretreatment variations, conclusions about the role of an interfacial oxide in silicide formation and its influence on phase formation were drawn. Such an oxide layer hinders silicide formation except for pinhole sites. This reduces the availability of Ni and causes a silicide with low Ni content to form. Without an interfacial oxide a continuous nickel silicide of greater depth, polycrystalline modification and expected phase according to thermal budget is formed. Information about the nature of silicide growth on typical solar cell surfaces could be obtained from silicide phase and geometric observations, which were supported by FIB tomography. The theory of isotropic NiSi growth and orientation dependent NiSi2 growth was derived. By this, a very well performing low-cost metallization for silicon solar cells has been brought an important step closer to industrial introduction.

Evaluation of anomalies during nickel and titanium silicide formation using the effective heat of formation mode

CSIR Research Space (South Africa)

Pretorius, R

1993-11-01

Full Text Available , as well as the observed sequence of growth of different silicide phases, are not in agree- ment with thermodynamic considerations [26]. In the case of the nickel silicides Ni,Si is nearly always found to be the first... to determine how the oxygen content in the silicon affects phase formation. We also show how the anomalous behaviour of titanium and nickel silicide formation can be explained thermodynamically by using the ?effective heat...

Radiation damage of metal uranium

International Nuclear Information System (INIS)

Mihajlovic, A.

1965-01-01

This report is concerned with the role of dispersion second phase in uranium and burnup rate. The role of dispersion phases in radiation stability of metal uranium was studies by three methods: variation of electric conductivity dependent on the neutron flux and temperature of pure uranium for different states of dispersion second phase; influence of dispersion phase on the radiation creep; transmission electron microscopy of fresh and irradiated uranium

High U-density nuclear fuel development with application of centrifugal atomization technology

International Nuclear Information System (INIS)

Kim, Chang Kyu; Kim, Ki Hwan; Lee, Don Bae

1997-01-01

In order to simplify the preparation process and improve the properties of uranium silicide fuels prepared by mechanical comminution, a fuel fabrication process applying rotating-disk centrifugal atomization technology was invented in KAERI in 1989. The major characteristic of atomized U 3 Si and U 3 Si 2 powders have been examined. The out-pile properties, including the thermal compatibility between atomized particle and aluminum matrix in uranium silicide dispersion fuels, have generally showed a superiority to the comminuted fuels. Moreover, the RERTR (reduced enrichment for research and test reactors) program, which recently begins to develop very-high-density uranium alloy fuels, including U-Mo fuels, requires the centrifugal atomization process to overcome the contaminations of impurities and the difficulties of the comminution process. In addition, a cooperation with ANL in the U.S. has been performed to develop high-density fuels with an application of atomization technology since December 1996. If the microplate and miniplate irradiation tests of atomized fuels, which have been performed with ANL, demonstrated the stability and improvement of in-reactor behaviors, nuclear fuel fabrication technology by centrifugal atomization could be most-promising to the production method of very-high-uranium-loading fuels. (author). 22 refs., 2 tabs., 12 figs

Calculation of oxygen distribution in uranium-plutonium oxide fuels during irradiation (programme CODIF)

International Nuclear Information System (INIS)

Moreno, A.; Sari, C.

1978-01-01

Radial gradients of oxygen to metal ratio, O/M, in uranium-plutonium oxide fuel pins, during irradiation and at the end of life, have been calculated on the basis of solid-state thermal diffusion using measured values of the heat of transport. A detailed computer model which includes the calculation of temperature profiles and the variation of the average O/M ratio as a function of burn-up is given. Calculations show that oxygen profiles are affected by the isotopic composition of the fuel, by the temperature profiles and by fuel-cladding interactions

Separation and purification of short lived fission products from irradiated uranium

International Nuclear Information System (INIS)

Balasubramanian, K.R.; Rao, K.L.N.; Mathai, C.; Varma, R.N.; Dhiwar, V.I.; Saxena, S.K.

1991-01-01

Fission produced radioisotopes like 95 Zr, 140 Ba, 103 Ru, 89 Sr and 91 Y whose half lives are less than 1 year find a wide variety of applications in the fields of industry, medicine and research. Isotope Division, BARC has been supplying these isotopes in hudreds of mCi amounts during the past several years. A new method for the sequential separation of these isotopes from irradiated uranium has been developed based on synthetic inorganic exchangers like stannic phosphate, polyphospho antimonic acid, hydrous manganese dioxide, etc. This report describes the new flow sheet worked out and adopted for the regular processing of these isotopes at hundreds of milli curie amounts. (author). 19 refs., 4 tabs. , 1 fig

Pt silicide/poly-Si Schottky diodes as temperature sensors for bolometers

Energy Technology Data Exchange (ETDEWEB)

Yuryev, V. A., E-mail: vyuryev@kapella.gpi.ru; Chizh, K. V.; Chapnin, V. A.; Mironov, S. A.; Dubkov, V. P.; Uvarov, O. V.; Kalinushkin, V. P. [A. M. Prokhorov General Physics Institute of the Russian Academy of Sciences, 38 Vavilov Street, Moscow 119991 (Russian Federation); Senkov, V. M. [P. N. Lebedev Physical Institute of the Russian Academy of Sciences, 53 Leninskiy Avenue, Moscow 119991 (Russian Federation); Nalivaiko, O. Y. [JSC “Integral” – “Integral” Holding Management Company, 121A, Kazintsa I. P. Street, Minsk 220108 (Belarus); Novikau, A. G.; Gaiduk, P. I. [Belarusian State University, 4 Nezavisimosti Avenue, 220030 Minsk (Belarus)

2015-05-28

Platinum silicide Schottky diodes formed on films of polycrystalline Si doped by phosphorus are demonstrated to be efficient and manufacturable CMOS-compatible temperature sensors for microbolometer detectors of radiation. Thin-film platinum silicide/poly-Si diodes have been produced by a CMOS-compatible process on artificial Si{sub 3}N{sub 4}/SiO{sub 2}/Si(001) substrates simulating the bolometer cells. Layer structure and phase composition of the original Pt/poly-Si films and the Pt silicide/poly-Si films synthesized by a low-temperature process have been studied by means of the scanning transmission electron microscopy; they have also been explored by means of the two-wavelength X-ray structural phase analysis and the X-ray photoelectron spectroscopy. Temperature coefficient of voltage for the forward current of a single diode is shown to reach the value of about −2%/ °C in the temperature interval from 25 to 50 °C.

Uranium in aqueous solutions by colorimetry

International Nuclear Information System (INIS)

Anon.

1981-01-01

The method covers the quantitative determination of uranium in known volumes of aqueous solutions that contain radioactive nuclides. These solutions arise from processing of irradiated nuclear fuel and from laboratory studies on irradiated uranium. The method is applicable to solutions containing a minimum of 30 μg of uranium per sample although as little as 0.5 μg can be detected but with lower precision. Highest precision is obtained with 50 to 75 μg of uranium in the test sample. Dilutions must be made at concentrations above 750 μg/ml. The method includes a discussion of photometers and photometric practice, apparatus, reagents, cell matching, preparation of standard curves, calibration by the method of internal standards, procedure, calculation, and precision

Recovering and recycling uranium used for production of molybdenum-99

Science.gov (United States)

Reilly, Sean Douglas; May, Iain; Copping, Roy; Dale, Gregory Edward

2017-12-12

A processes for recycling uranium that has been used for the production of molybdenum-99 involves irradiating a solution of uranium suitable for forming fission products including molybdenum-99, conditioning the irradiated solution to one suitable for inducing the formation of crystals of uranyl nitrate hydrates, then forming the crystals and a supernatant and then separating the crystals from the supernatant, thus using the crystals as a source of uranium for recycle. Molybdenum-99 is recovered from the supernatant using an adsorbent such as alumina. Another process involves irradiation of a solid target comprising uranium, forming an acidic solution from the irradiated target suitable for inducing the formation of crystals of uranyl nitrate hydrates, then forming the crystals and a supernatant and then separating the crystals from the supernatant, thus using the crystals as a source of uranium for recycle. Molybdenum-99 is recovered from the supernatant using an adsorbent such as alumina.

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)

Molybdenum silicide based materials and their properties

International Nuclear Information System (INIS)

Yao, Z.; Stiglich, J.; Sudarshan, T.S.

1999-01-01

Molybdenum disilicide (MoSi 2 ) is a promising candidate material for high temperature structural applications. It is a high melting point (2030 C) material with excellent oxidation resistance and a moderate density (6.24 g/cm 3 ). However, low toughness at low temperatures and high creep rates at elevated temperatures have hindered its commercialization in structural applications. Much effort has been invested in MoSi 2 composites as alternatives to pure molybdenum disilicide for oxidizing and aggressive environments. Molybdenum disilicide-based heating elements have been used extensively in high-temperature furnaces. The low electrical resistance of silicides in combination with high thermal stability, electron-migration resistance, and excellent diffusion-barrier characteristics is important for microelectronic applications. Projected applications of MoSi 2 -based materials include turbine airfoils, combustion chamber components in oxidizing environments, missile nozzles, molten metal lances, industrial gas burners, diesel engine glow plugs, and materials for glass processing. On this paper, synthesis, fabrication, and properties of the monolithic and composite molybdenum silicides are reviewed

Effects of enriched uranium on developing brain damage of neonatal rats

International Nuclear Information System (INIS)

Gu Guixiong; Zhu Shoupeng; Wang Liuyi; Yang Shuqin; Zhu Lingli

2001-01-01

The model of irradiation-induced brain damage in vivo was settled first of all. The micro-auto-radiographic tracing showed that when the rat's brain at postnatal day after lateral ventricle injection with enriched uranium 235 U the radionuclides were mainly accumulated in the nucleus. At the same time autoradiographic tracks appeared in the cytoplasm and interval between cells. The effects of cerebrum exposure to alpha irradiation by enriched uranium on somatic growth and neuro-behavior development of neonatal rats were examined by determination of multiple parameters. In the growth and development of the neonatal rat's cerebrum exposure to enriched uranium, the somatic growth such as body weight and brain weight increase was lower significantly. The data indicated that the neonatal wistar rats having cerebrum exposure to alpha irradiation by enriched uranium showed delayed growth and abnormal neuro-behavior. The changes of neuron specific enolase (NSE), interleukin-1 β (IL- β), superoxide dismutase (SOD), and endothelin (ET) in cerebellum, cerebral cortex, hippocampus, diencephalons of the rat brain after expose to alpha irradiation by enriched uranium were examined with radioimmunoassay. The results showed that SOD and ET can be elevated by the low dose irradiation of enriched uranium, and can be distinctly inhibited by the high dose. The data in view of biochemistry indicated firstly that alpha irradiation from enriched uranium on the developing brain damage of neonatal rats were of sensibility, fragility and compensation in nervous cells

Effects of enriched uranium on developing brain damage of neonatal rats

Energy Technology Data Exchange (ETDEWEB)

Guixiong, Gu; Shoupeng, Zhu; Liuyi, Wang; Shuqin, Yang; Lingli, Zhu [Suzhou Medical College, Suzhou (China)

2001-04-01

The model of irradiation-induced brain damage in vivo was settled first of all. The micro-auto-radiographic tracing showed that when the rat's brain at postnatal day after lateral ventricle injection with enriched uranium {sup 235}U the radionuclides were mainly accumulated in the nucleus. At the same time autoradiographic tracks appeared in the cytoplasm and interval between cells. The effects of cerebrum exposure to alpha irradiation by enriched uranium on somatic growth and neuro-behavior development of neonatal rats were examined by determination of multiple parameters. In the growth and development of the neonatal rat's cerebrum exposure to enriched uranium, the somatic growth such as body weight and brain weight increase was lower significantly. The data indicated that the neonatal wistar rats having cerebrum exposure to alpha irradiation by enriched uranium showed delayed growth and abnormal neuro-behavior. The changes of neuron specific enolase (NSE), interleukin-1 {beta} (IL- {beta}), superoxide dismutase (SOD), and endothelin (ET) in cerebellum, cerebral cortex, hippocampus, diencephalons of the rat brain after expose to alpha irradiation by enriched uranium were examined with radioimmunoassay. The results showed that SOD and ET can be elevated by the low dose irradiation of enriched uranium, and can be distinctly inhibited by the high dose. The data in view of biochemistry indicated firstly that alpha irradiation from enriched uranium on the developing brain damage of neonatal rats were of sensibility, fragility and compensation in nervous cells.

Studies on the separation and purification of uranium from thoria irradiated in PHWR

International Nuclear Information System (INIS)

Kedari, C.S.; Sinalkar, N.; Dhami, P.S.; Achuthan, P.V.; Vijayan, K.; Moorthy, A.D.; Jambunathan, U.; Munshi, S.K.; Dey, P.K.

2005-10-01

About 5 g of thoria pellet extracted from the thoria bundle irradiated to 508 full power days in PHWR at KAPS and cooled for about 4.5 years was dissolved in 25 mL of 13 M HNO 3 containing 0.03 M HF and 0.1 M Al(NO 3 ) 3 under reflux condition in a shielded glove box. Thorium concentration and acidity of the dissolved solution was adjusted to ∼97 g/L and 4M HNO 3 , suitable for solvent extraction studies. The feed solution had a uranium concentration of 1.36 g/L and gross gamma activity of ∼6 Ci/L. Liquid-liquid extraction experiments were carried out in batch with this feed using 3% TBP in n-dodecane as extractant. The uranium product solution obtained from the solvent extraction was further purified by ion-exchange method using cation exchange resin Dowex 50 x 4. Fission product activity and the isotopic composition of the final recovered product were measured. The uranium product had an overall decontamination factor of 10 4 and 2 x 10 4 with respect to thorium and fission product activity respectively. A part of the solution was monitored for dose build-up due to the growth of the decay products of 232 U which is a vital input for designing the shielding of a 233 U separation facility. (author)

Status of research reactor fuel development in KAERI

International Nuclear Information System (INIS)

Kim, Chang-Kyu; Ryu, Woo-Seok; Park, Jong-Man; Lee, Don-Bae; Kim, Ki-Hwan; Kuk, Il-Hyun

1996-01-01

The development of uranium silicide dispersion fuel fabrication technology has been carried out in KAERI. LEU fuel bundle was prepared for irradiation test. In order to compare the performance of atomized and comminuted U 3 Si dispersed fuels, the bundle of two kinds of fuel elements were prepared. Irradiation test will be performed in the OR-hole of HANARO in the near future. U 3 Si 2 atomization technology has been improved by using ceramic crucible and nozzle. Irradiation test for atomized U 3 Si 2 plate type fuel will be carried out in cooperation with ANL by using HANARO in connection with RERTR advanced fuel development. (author)

The oxidation behaviour of uranium in air at 348-765 K

International Nuclear Information System (INIS)

Bennett, M.J.; Price, J.B.

1981-01-01

The oxidation behaviour of adjusted uranium has been examined in air, at atmospheric pressure, at 348-765 K. Particular emphasis has been directed to the role of swelling resulting from prior irradiation of the metal to a burn-up of 5600-9100 MWd/t and an addition of 2.5 x 10 4 vpm water vapour to the air. Pre-irradiation of uranium enhanced its attack by air at 348-523 K, the enhancement increasing progressively with percentage swelling. This effect resulted primarily from the break-up of the uranium surface during oxidation with the generation of a greater reaction surface area and was independent of the temperature of oxidation in dry air and also above 423 K in wet air. At lower temperatures, however, the water vapour addition increased the effective reaction rate, possibly by the transistory involvement of uranium hybride. The influence of the water vapour increased with swelling of the irradiated uranium and was greater than that exerted on the oxidation of unirradiated uranium at comparable temperatures. With increasing temperature above 623 K, swelling had a progressively decreasing influence upon the attack of irradiated uranium in both environments. (orig.)

Graphite Isotope Ratio Method Development Report: Irradiation Test Demonstration of Uranium as a Low Fluence Indicator

International Nuclear Information System (INIS)

Reid, B.D.; Gerlach, D.C.; Love, E.F.; McNeece, J.P.; Livingston, J.V.; Greenwood, L.R.; Petersen, S.L.; Morgan, W.C.

1999-01-01

This report describes an irradiation test designed to investigate the suitability of uranium as a graphite isotope ratio method (GIRM) low fluence indicator. GIRM is a demonstrated concept that gives a graphite-moderated reactor's lifetime production based on measuring changes in the isotopic ratio of elements known to exist in trace quantities within reactor-grade graphite. Appendix I of this report provides a tutorial on the GIRM concept

Results of Uranium Dioxide-Tungsten Irradiation Test and Post-Test Examination

Science.gov (United States)

Collins, J. F.; Debogdan, C. E.; Diianni, D. C.

1973-01-01

A uranium dioxide (UO2) fueled capsule was fabricated and irradiated in the NASA Plum Brook Reactor Facility. The capsule consisted of two bulk UO2 specimens clad with chemically vapor deposited tungsten (CVD W) 0.762 and 0.1016 cm (0.030-and 0.040-in.) thick, respectively. The second specimen with 0.1016-cm (0.040-in.) thick cladding was irradiated at temperature for 2607 hours, corresponding to an average burnup of 1.516 x 10 to the 20th power fissions/cu cm. Postirradiation examination showed distortion in the bottom end cap, failure of the weld joint, and fracture of the central vent tube. Diametral growth was 1.3 percent. No evidence of gross interaction between CVD tungsten or arc-cast tungsten cladding and the UO2 fuel was observed. Some of the fission gases passed from the fuel cavity to the gas surrounding the fuel specimen via the vent tube and possibly the end-cap weld failure. Whether the UO2 loss rates through the vent tube were within acceptable limits could not be determined in view of the end-cap weld failure.

Development of Self-Healing Zirconium-Silicide Coatings for Improved Performance Zirconium-Alloy Fuel Cladding

Energy Technology Data Exchange (ETDEWEB)

Sridharan, Kumar [University of Wisconsin-Madison; Mariani, Robert [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bai, Xianming [Idaho National Lab. (INL), Idaho Falls, ID (United States); Xu, Peng [Westinghouse Electric Company; Lahoda, Ed [Westinghouse Electric Company

2018-03-31

oxide film matrix which develops at high temperatures. Cold spray and thermal spray deposition of ZrSi2 have been performed by systematically varying deposition parameters. While deposition of coating was achieved, further process optimization is required to improve coating quality deposited using these methods. An optimal zirconium-silicide slurry coating for mechanical properties and corrosion resistance/oxide protection was developed based on a theoretical characterization. The slurry coating was investigated using a range of similar compositional coatings to accurately characterize the best coating with respect its mechanical integrity and environmental protection for zirconium alloy substrates. Distinctive oxide layers of ZrSi2 prepared at 1000°C and 1400°C in ambient air were subjected to a 3.9 MeV Si2+ ions irradiation at 305°C and their radiation responses were characterized and analyzed. Nanocomposite oxides consisting of ZrO2 nanocrystals embedded in amorphous SiO2 matrix formed on ZrSi2 surface after oxidation at 1000°C. Radiation-induced phase mixing of the oxide phases and amorphization of ZrO2 was observed up to ~ 820 nm in depth, about one-third of the total radiation damaged region (2.5 µm in thickness) as estimated by SRIM calculation. Polygonal crystalline ZrSiO4 grains in dual-layered oxide scale on ZrSi2 at 1400°C were completely amorphized under the ion-irradiation. Given the high corrosion resistance of ZrSiO4 and immobilization of Si in aqueous environments conclusively demonstrated in this study, it is anticipated that the irradiated oxide scale would have a superior corrosion resistance compared to unirradiated surface. The results can be applied to not only baseline data for radiation response of the potential neutron deflector in advanced reactors but also development of nanocomposite structural materials in future spacecraft, sensors, detector subjected to high radiation dose. A quench test facility was designed and built to study

The whole-core LEU silicide fuel demonstration in the JMTR

Energy Technology Data Exchange (ETDEWEB)

Aso, Tomokazu; Akashi, Kazutomo; Nagao, Yoshiharu [Japan Atomic Energy Research Institute, Ibaraki-ken (Japan)] [and others

1997-08-01

The JMTR was fully converted to LEU silicide (U{sub 3}Si{sub 2}) fuel with cadmium wires as burnable absorber in January, 1994. The reduced enrichment program for the JMTR was initiated in 1979, and the conversion to MEU (enrichment ; 45%) aluminide fuel was carried out in 1986 as the first step of the program. The final goal of the program was terminated by the present LEU conversion. This paper describes the results of core physics measurement through the conversion phase from MEU fuel core to LEU fuel core. Measured excess reactivities of the LEU fuel cores are mostly in good agreement with predicted values. Reactivity effect and burnup of cadmium wires, therefore, were proved to be well predicted. Control rod worth in the LEU fuel core is mostly less than that in the MEU fuel core. Shutdown margin was verified to be within the safety limit. There is no significant difference in temperature coefficient of reactivity between the MEU and LEU fuel cores. These results verified that the JMTR was successfully and safely converted to LEU fuel. Extension of the operating cycle period was achieved and reduction of spend fuel elements is expected by using the fuel with high uranium density.

Long term developments in irradiated natural uranium processing costs. Optimal size and siting of plants

International Nuclear Information System (INIS)

Thiriet, L.

1964-01-01

The aim of this paper is to help solve the problem of the selection of optimal sizes and sites for spent nuclear fuel processing plants associated with power capacity programmes already installed. Firstly, the structure of capital and running costs of irradiated natural uranium processing plants is studied, as well as the influence of plant sizes on these costs and structures. Shipping costs from the production site to the plant must also be added to processing costs. An attempt to reach a minimum cost for the production of a country or a group of countries must therefore take into account both the size and the location of the plants. The foreseeable shipping costs and their structure (freight, insurance, container cost and depreciation), for spent natural uranium are indicated. Secondly, for various annual spent fuel reprocessing programmes, the optimal sizes and locations of the plants are determined. The sensitivity of the results to the basic assumptions relative to processing costs, shipping costs, the starting up year of the plant programme and the length of period considered, is also tested. - this rather complex problem, of a combinative nature, is solved through dynamic programming methods. - It is shown that these methods can also be applied to the problem of selecting the optimal sizes and locations of processing plants for MTR type fuel elements, related to research reactor programmes, as well as to future plutonium element processing plants related to breeder reactors. Thirdly, the case where yearly extraction of the plutonium contained in the irradiated natural uranium is not compulsory is examined; some stockpiling of the fuel is then allowed some years, entailing delayed processing. The load factor of such plants is thus greatly improved with respect to that of plants where the annual plutonium demand is strictly satisfied. By including spent natural uranium stockpiling costs an optimal rhythm of introduction and optimal sizes for spent fuel

75 years ago. The discovery of nuclear fission during the neutron irradiation of uranium on 17 December 1938 in Berlin

International Nuclear Information System (INIS)

Kuczera, Bernhard

2013-01-01

A ground-breaking, scientific discovery made in the Berlin experiments of the year 1938 on neutron irradiation of uranium, was that atom nuclei can be fissured and, during the fission energy, so-called nuclear energy, is released. If further U-235 atoms are in the immediate vicinity, the released neutrons can trigger a chain reaction of the nuclear fission. The story of this ground-breaking discovery that led to the worldwide civil utilisation of nuclear energy is outlined here. The discovery of neutrons and their implementation in radiation experiments occurred in the 1930s. The research work followed the model of the Stockholm citizen Niels Bohr on the configuration of an atom. In 1930, Walther Bothe reported on the occurrence of a high-energy radiation when beryllium was irradiated with alpha-particles. James Chadwick could prove experimentally in 1932 that this is a current of high-energy, electrically neutral particles. He called these particles neutrons. In Berlin, the radio-chemist Otto Hahn is the Director at the Kaiser-Wilhelm-Institute for Chemistry and head of the chemical-radioactive department. In the years 1935 to 1938, the radiation experiments of uranium with neutrons were at the centre of the research in Berlin. In October 1938, Hahn and Strassmann responded with their own experiments to the scientific challenge of Irene Curie and Paul Savitch of 3.5-hours of activity when uranium was radiated with neutrons. The decisive discovery for the neutron radiation of uranium was made by Hahn and Strassmann on 17 December 1938. From the results of different experiments, they conclude: 'Our radium isotopes have the properties of barium.' On 19 December, Hahn informs Lise Meitner and speaking on the issue of 'Where does the barium come from?' addresses a possible 'bursting' of the uranium nucleus. That is Hahn's 'Heureka.' moment: He recognises that the neutron radiation does not only effect radiochemical transformations, but can also cause the uranium

Nickel silicide thin films as masking and structural layers for silicon bulk micro-machining by potassium hydroxide wet etching

International Nuclear Information System (INIS)

Bhaskaran, M; Sriram, S; Sim, L W

2008-01-01

This paper studies the feasibility of using titanium and nickel silicide thin films as mask materials for silicon bulk micro-machining. Thin films of nickel silicide were found to be more resistant to wet etching in potassium hydroxide. The use of nickel silicide as a structural material, by fabricating micro-beams of varying dimensions, is demonstrated. The micro-structures were realized using these thin films with wet etching using potassium hydroxide solution on (1 0 0) and (1 1 0) silicon substrates. These results show that nickel silicide is a suitable alternative to silicon nitride for silicon bulk micro-machining

Capital and operating costs of irradiated natural uranium reprocessing plants; Couts d'investissement et d'exploitation des usines de retraitement de l'uranium naturel irradie

Energy Technology Data Exchange (ETDEWEB)

Thiriet, L; Jouannaud, C; Couture, J; Duboz, J [Commissariat a l' Energie Atomique (France). Centre d' Etudes Nucleaires; Oger, C [Saint Gobain Nucleaire (France)

1966-07-01

This paper presents first a method of analysing natural uranium reprocessing plants investment costs (method similar to LANG and BACH well known in the fuel oil industry) and their operating costs (analysed according to their economic type). This method helps establishing standard cost structures for these plants, allowing thus comparisons between existing or planned industrial facilities. It also helps evaluating the foreseeable consequences of technical progress. Some results obtained are given, concerning: the investment costs sensitivity to the various technical parameters defining the fuel and their comparison according to the country or the economic area taken into account. Finally, the influence of the plants size on their investment costs is shown. (author) [French] La communication expose d'abord une methode d'analyse des couts d'investissement des usines de retraitement de l'uranium naturel irradie (inspiree de celles de LANG et de BACH, bien connues dans l'industrie petroliere) et de leurs couts d'exploitation (selon leur nature economique). Cette methode permet d'etablir des structures types de couts de ces usines et de comparer les realisations industrielles et les projets. Elle facilite l'exploration des consequences previsibles du progres technique. On indique un certain nombre de resultats obtenus, concernant la sensibilite des couts d'investissement de ces usines aux differents parametres techniques definissant le combustible et leur confrontation selon les pays ou aires economiques envisages. On montre enfin comment doit pouvoir s'exprimer l'influence de la taille des usines sur leur cout d'investissement. (auteur)

Self-organized patterns along sidewalls of iron silicide nanowires on Si(110) and their origin

Energy Technology Data Exchange (ETDEWEB)

Das, Debolina; Mahato, J. C.; Bisi, Bhaskar; Dev, B. N., E-mail: msbnd@iacs.res.in [Department of Materials Science, Indian Association for the Cultivation of Science, Kolkata 700032 (India); Satpati, B. [Surface Physics and Material Science Division, Saha Institute of Nuclear Physics, 1/AF Bidhannagar, Kolkata 700064 (India)

2014-11-10

Iron silicide (cubic FeSi{sub 2}) nanowires have been grown on Si(110) by reactive deposition epitaxy and investigated by scanning tunneling microscopy and scanning/transmission electron microscopy. On an otherwise uniform nanowire, a semi-periodic pattern along the edges of FeSi{sub 2} nanowires has been discovered. The origin of such growth patterns has been traced to initial growth of silicide nanodots with a pyramidal Si base at the chevron-like atomic arrangement of a clean reconstructed Si(110) surface. The pyramidal base evolves into a comb-like structure along the edges of the nanowires. This causes the semi-periodic structure of the iron silicide nanowires along their edges.

Core conversion study from silicide to molybdenum fuel in the Indonesian 30 MW multipurpose reactor G.A. Siwabessy (RSG-GAS)

International Nuclear Information System (INIS)

Sembiring, T.M.; Kuntoro, I.

2005-01-01

This paper describes the core conversion from silicide to molybdenum core through a series of silicide (2.96 gU cm -3 ) - molybdenum (3.55 gUcm -3 ) mixed transition cores for the Indonesian 30 MW-Multipurpose G.A. Siwabessy (RSGGAS) reactor. The core calculations are carried out using the two-dimensional multigroup neutron diffusion method code of Batan-EQUIL-2D. The calculated results showed that the proposed silicide-molybdenum mixed transition cores, using the same refueling/reshuffling scheme, meet the safety criteria and it can be used in safely converting from an all-silicide core to an all-molybdenum core. (author)

Material test reactor fuel research at the BR2 reactor

Energy Technology Data Exchange (ETDEWEB)

Dyck, Steven Van; Koonen, Edgar; Berghe, Sven van den [Institute for Nuclear Materials Science, SCK-CEN, Boeretang, Mol (Belgium)

2012-03-15

The construction of new, high performance material test reactor or the conversion of such reactors' core from high enriched uranium (HEU) to low enriched uranium (LEU) based fuel requires several fuel qualification steps. For the conversion of high performance reactors, high density dispersion or monolithic fuel types are being developed. The Uranium-Molybdenum fuel system has been selected as reference system for the qualification of LEU fuels. For reactors with lower performance characteristics, or as medium enriched fuel for high performance reactors, uranium silicide dispersion fuel is applied. However, on the longer term, the U-Mo based fuel types may offer a more efficient fuel alternative and-or an easier back-end solution with respect to the silicide based fuels. At the BR2 reactor of the Belgian nuclear research center, SCK-CEN in Mol, several types of fuel testing opportunities are present to contribute to such qualification process. A generic validation test for a selected fuel system is the irradiation of flat plates with representative dimensions for a fuel element. By flexible positioning and core loading, bounding irradiation conditions for fuel elements can be performed in a standard device in the BR2. For fuel element designs with curved plates, the element fabrication method compatibility of the fuel type can be addressed by incorporating a set of prototype fuel plates in a mixed driver fuel element of the BR2 reactor. These generic types of tests are performed directly in the primary coolant flow conditions of the BR2 reactor. The experiment control and interpretation is supported by detailed neutronic and thermal-hydraulic modeling of the experiments. Finally, the BR2 reactor offers the flexibility for irradiation of full size prototype fuel elements, as 200mm diameter irradiation channels are available. These channels allow the accommodation of various types of prototype fuel elements, eventually using a dedicated cooling loop to provide the

LEU-plate irradiation at FRJ-2 (DIDO) under the German AF-programme

Energy Technology Data Exchange (ETDEWEB)

Groos, E; Krug, W; Seferiadis, J; Thamm, G

1985-07-01

10 LEU fuel plates (8 with uranium silicides max. U-density 6.1 g/cm{sup 3}) have been irradiated at FRJ-2 (DIDO) of KFA-Juelich till end of October 1984 during 321 full power days up to max. burnup of 2.41x10{sup 27} fissions/m{sup 3} without major interruptions and troubles. PIE began recently in KFA hot cells. Visual inspections revealed no damage or greater deformation for the majority of the plates, but red/brown coloured layers (partially peeled off) on the cladding over the fuel. Aluminium (oxide) is the chief constituent of the layer with smaller portions of Ni and Fe the latter causing the red/brown colour. The major part of the layer ({approx}50 {mu}m) most probably has been formed during 20 h immediately after experiment start-up under abnormal conditions of the coolant water. Gamma scanning has been completed. Dimensional measurements are under way confirming first observations of severe swelling (pillowing) of 1 plate. Density and blister testing as well as metallography and burnup analysis remain to be accomplished end of 1985/beginning of 1986. (author)

Proceedings of the international symposium on uranium and electricity

International Nuclear Information System (INIS)

Talbot, K.H.; Lakshmanan, V.I.

1988-01-01

The first Canadian Nuclear Society 'Uranium and Electricity' Conference provided a forum for discussion on developments in all phases of the nuclear/electricity fuel cycle. It included both technical and management aspects in areas of mutual interest to those involved in mining, refining, fuel manufacture, reactor operation and irradiated fuel. The 60 papers offered in this volume present detailed discussions on uranium exploration and mining; waste management; uranium metallurgy, refining and byproducts; irradiated fuel management; effluent management, uranium chemistry, refining and the environment; environmental protection from low level waste; fuel design, manufacture and performance; regulation and control of radiological hazards; Canadian power reactor fuel and safety trends; and future developments

Multi-layered silicides coating for vanadium alloys for generation IV reactors

International Nuclear Information System (INIS)

Mathieu, S.; Chaia, N.; Vilasi, M.; Le Flem, M.

2012-01-01

The halide-activated pack-cementation technique was employed to fabricate a diffusion coating that is resistant both to isothermal and to cyclic oxidation in air at 650 degrees C on the surface of the V-4Cr-4Ti vanadium alloy that is a potential core component of future nuclear systems. A thermodynamic assessment determined the deposit conditions in terms of master alloy, activator, filler and temperature. The partial pressures of the main gaseous species (SiCl 4 , SiCl 2 and VCl 2 ) in the pack were calculated with the master alloy Si and the mixture VSi 2 + Si. The VSi 2 + Si master alloy was used to limit vanadium loss from the surface. The obtained coating consisted of multi-layered V x Si y silicides with an outer layer of VSi 2 . This silicide developed a protective layer of silica at 650 degrees C in air and was not susceptible to the pest phenomenon, unlike other refractory silicides (MoSi 2 , NbSi 2 ). We suggest that VSi 2 exhibits no risk of rapid degradation in the gas fast reactor (GFR) conditions. (authors)

Study of ammonia synthesis over uranium catalysts

International Nuclear Information System (INIS)

Spitsyn, V.I.; Erofeev, B.V.; Mikhajlenko, I.E.; Gorelkin, I.I.; Ivanov, L.S.

1980-01-01

The effect of induced radiactivity and chemical composition of uranium catalysts on their catalytic activity in the ammonia synthesis reaction has been studied. The catalyst samples comprise pieces of metal uranium and chip irradiated in nuclear reactor by the 4.3x10 16 n/cm 2 integral flux of slow neutrons. Studies of catalytic activity was carried out at 1 atm and 340-510 deg C when stoichiometric nitrogen-hydrogen mixture passed through the following installation. At different temperatures uranium nitrides of different composition are shown to be formed. Uranium nitrides with the composition close to UN 2 are the samples with the highest catalYtic activity. The reduction of catalytic activity of uranium catalysts with the increased temperature of their formation above 400 deg C is explained by low catalytic activity of forming UNsub(1.7) in comparison with UN 2 . Catalytic properties of irradiated and nonirradiated samples do not differ from one another

Determination of uranium in industrial and environmental samples. Vol. 4

Energy Technology Data Exchange (ETDEWEB)

El-Sweify, F H; Shehata, M K; Metwally, E M; El-Shazly, E A.A.; El-Naggar, H A [Nuclear Chemistry Department, Hot Laborities Center, Atomic Energy Authority, Cairo (Egypt)

1996-03-01

The phosphate ores used at `Abu zaabal fertilizer and chemical company` for the production of some chemicals and fertilizers contain detectable amounts of uranium. In this study, the content of uranium in samples of different products of fertilizers, gypsum, and phosphate ore were determined using NAA, and gamma ray spectroscopy of the irradiated samples. Another method based on measuring the natural radioactivity of {sup 238} U series for non-irradiated samples using gamma-ray spectroscopy was also used for determine uranium content in the samples. In the NAA method, the content of U(ppm) in the samples was been computed from the photopeak activity of the lines = 106.1, 228.2, and 277.5 KeV of {sup 239} Np induced in the irradiated samples, and the uranium standard simultaneously irradiated. the gamma-ray spectra, and the decay curves are given. In the second method the gamma-ray spectra of the natural radioactivity of the samples and uranium standard were measured. The gamma-transition of energies 295.1, 251.9 KeV for {sup 214} Pb; 609.3, 768.4, 1120.3, 1238.1 KeV for {sup 214} Bi were determined. The uranium {sup 23U} traces in drainage water was also determined spectrophotometrically using arsenazo-III after preconcentration of uranium from the pretreated drainage water in column packed with chelex-100 resin. The recovery was found to be 90 {+-} 5%. 11 figs., 3 tabs.

The Leakage Current Improvement of a Ni-Silicided SiGe/Si Junction Using a Si Cap Layer and the PAI Technique

International Nuclear Information System (INIS)

Chang Jian-Guang; Wu Chun-Bo; Ji Xiao-Li; Ma Hao-Wen; Yan Feng; Shi Yi; Zhang Rong

2012-01-01

We investigate the leakage current of ultra-shallow Ni-silicided SiGe/Si junctions for 45 nm CMOS technology using a Si cap layer and the pre-amorphization implantation (PAI) process. It is found that with the conventional Ni silicide method, the leakage current of a p + (SiGe)—n(Si) junction is large and attributed to band-to-band tunneling and the generation-recombination process. The two leakage contributors can be suppressed quite effectively when a Si cap layer is added in the Ni silicide method. The leakage reduction is about one order of magnitude and could be associated with the suppression of the agglomeration of the Ni germano-silicide film. In addition, the PAI process after the application of a Si cap layer has little effect on improving the junction leakage but reduces the sheet resistance of the silicide film. As a result, the novel Ni silicide method using a Si cap combined with PAI is a promising choice for SiGe junctions in advanced technology. (cross-disciplinary physics and related areas of science and technology)

A Very High Uranium Density Fission Mo Target Suitable for LEU Using atomization Technology

Energy Technology Data Exchange (ETDEWEB)

Kim, C. K.; Kim, K. H.; Lee, Y. S.; Ryu, H. J.; Woo, Y. M.; Jang, S. J.; Park, J. M.; Choi, S. J. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2011-05-15

Currently HEU minimization efforts in fission Mo production are underway in connection with the global threat reduction policy. In order to convert HEU to LEU for the fission Mo target, higher uranium density material could be applied. The uranium aluminide targets used world widely for commercial {sup 99}Mo production are limited to 3.0 g-U/cc in uranium density of the target meat. A consideration of high uranium density using the uranium metal particles dispersion plate target is taken into account. The irradiation burnup of the fission Mo target are as low as 8 at.% and the irradiation period is shorter than 7 days. Pure uranium material has higher thermal conductivity than uranium compounds or alloys. It is considered that the degradation by irradiation would be almost negligible. In this study, using the computer code of the PLATE developed by ANL the irradiation behavior was estimated. Some considerations were taken into account to improve the irradiation performance further. It has been known that some alloying elements of Si, Cr, Fe, and Mo are beneficial for reducing the swelling by grain refinement. In the RERTR program recently the interaction problem could be solved by adding a small amount of Si to the aluminum matrix phase. The fabrication process and the separation process for the proposed atomized uranium particles dispersion target were reviewed

Development of LEU targets for 99Mo production and their chemical processing status 1989

International Nuclear Information System (INIS)

Vandegrift, G.F.; Kwok, J.D.; Chamberlain, D.B.; Hoh, J.C.; Streets, E.W.; Vogler, S.; Thresh, H.R.; Domagala, R.F.; Wiencek, T.C.; Matos, J.E.

1991-01-01

Most of the world's supply of Tc-99m for medical purposes is currently produced from Mo-99 derived from the fissioning of high enriched uranium (HEU). Substitution of low enriched uranium (LEU) silicide fuel for the HEU alloy and aluminide fuels used in current target designs will allow equivalent Mo-99 yields with no change in target geometries. Substitution of uranium metal will also allow the substitution of LEU for HEU. Efforts performed in 1989 focused on (1) fabrication of a uranium metal target by Hot Isostatic Pressing uranium metal foil to zirconium, (2) experimental investigation of the dissolution step for U 3 Si 2 targets, allowing us to present a conceptual design for the dissolution process and equipment, and (3) investigation of the procedures used to reclaim irradiated uranium from Mo-production targets, allowing us to further analyze the waste and by-product problems associated with the substitution of LEU for HEU. (orig.)

Safeguarding subcriticality during loading and shuffling operations in the higher density of the RSG-GAS's silicide core

International Nuclear Information System (INIS)

Sembiring, T.M.; Kuntoro, I.

2003-01-01

The core conversion program of the RSG-GAS reactor is to convert the all-oxide to all-silicide core. The silicide equilibrium core with fuel meat density of 3.55 gU cm -3 is an optimal core for RSG-GAS reactor and it can significantly increase the operation cycle length from 25 to 32 full power days. Nevertheless, the subcriticality of the shutdown core and the shutdown margin are lower than of the oxide core. Therefore, the deviation of subcriticality condition in the higher silicide core caused by the fuel loading and shuffling error should be reanalysed. The objective of this work is to analyse the sufficiency of the subcriticality condition of the shutdown core to face the worst condition caused by an error during loading and shuffling operations. The calculations were carried out using the 2-dimensional multigroup neutron diffusion code of Batan-FUEL. In the fuel handling error, the calculated results showed that the subcriticality condition of the shutdown higher density silicide equilibrium core of RSG-GAS can be maintained. Therefore, all fuel management steps are fixed in the present reactor operation manual can be applied in the higher silicide equilibrium core of RSG-GAS reactor. (author)

Kinetics of low pressure chemical vapor deposition of tungsten silicide from dichlorocilane reduction of tungsten hexafluoride

International Nuclear Information System (INIS)

Srinivas, D.; Raupp, G.B.; Hillman, J.

1990-01-01

The authors report on experiments to determine the intrinsic surface reaction rate dependences and film properties' dependence on local reactant partial pressures and wafer temperature in low pressure chemical vapor deposition (LPCVD) of tungsten silicide from dichlorosilane reduction of tungsten hexafluoride. Films were deposited in a commercial-scale Spectrum CVD cold wall single wafer reactor under near differential, gradientless conditions. Over the range of process conditions investigated, deposition rate was found to be first order in dichlorosillane and negative second order in tungsten hexafluoride partial pressure. The apparent activation energy in the surface reaction limited regime was found to be 70-120 kcal/mol. The silicon to tungsten ratio of as deposited silicide films ranged from 1.1 to 2.4, and increased with increasing temperature and dichlorosillane partial pressure, and decreased with increasing tungsten hexafluoride pressure. These results suggest that the apparent silicide deposition rate and composition are controlled by the relative rates of at least two competing reactions which deposit stoichiometric tungsten silicides and/or silicon

Measurement of uranium enrichment by 14 MeV neutron irradiation

International Nuclear Information System (INIS)

Rezende, H.R.

1987-01-01

a non-destructive technique for the determination of uranium in UO 2 samples was developed, making use of the change in the fission cross section of a nuclide with the neutron energy. The active interrogation method was used by irradiating the samples with pulsed 14 MeV neutrons and further detection of delayed fission neutrons. In order to discriminate U-238 from U-235 the neutron energy was tailored by means of two concentric cylinders of lead and paraffin/poliethylene, 11 and 4 cm thick. Between neutron pulses, delayed neutrons from fission were detected by a long counter built with five BF 3 proportional counters. Calibration curves for enrichment and total mass versus delayed neutron response were obtained using available UO 2 pellets of known enrichment. Enrichment detection limit, obtained with 95% confidence level by the Student distribution was estimated to be 0.33%. The minimal detectable mass was estimated to be 4.4 g. (author) [pt

Measure of uranium enrichment by 14 MeV neutron irradiation

International Nuclear Information System (INIS)

Rezende, H.R.

1987-01-01

A non-destructive technique for the determination of uranium in UO 2 samples was developed, marking use of the change in the fission cross of a nuclide with the neutron energy. The active interrogation method was used by irradiating the samples with pulsed 14 MeV neutrons and furtherdetection of delayed fission neutrons. In order to descriminated U-238 from U-235 the neutron energy was tailored by means of two concentric cylinders of lead and paraffin/poliethylene, 11 and 4 cm thick. Between neutron pulses, delayed neutrons from fission were detected by a long counter built with five BF 3 proportional counters. Calibration curves for enrichment and total mass versus delayed neutron response were obtained using available UO 2 pellets of Known enrichment. Enrichment detection limit, obtained with 95% confidence level by the the Student distribution was estimated to be 0.33%. The minimal detectable mass was estimated to be 4.4 g. (Author) [pt

Health and environmental problems of using antiarmour munitions containing depleted uranium core

International Nuclear Information System (INIS)

Matousek, J.

2006-01-01

In the 1970s, core of depleted uranium commenced to be introduced into the breakthrough antitank munitions of various calibers and types in order to considerably enhance their effectiveness due to extremely high density in comparison with steel. The health and environmental threats of using this munitions and other weaponry where depleted uranium has been utilised as counterbalance stem from the pyrophoric character of uranium, burnt due to material deformation and friction when penetrating armour targets creating thus highly respirable aerosol of uranium oxides that are deposited in alveoli after being inhaled or in other tissues after being ingested. Composition and main properties of depleted uranium are presented. Chronic effects of deposited particles of uranium oxides are due to internal irradiation of sensitive organs at proceeding radioactive decay accompanied with alpha irradiation. Long-term internal irradiation by radionuclides producing alpha-rays leads to proved risk of increased incidence of carcinoma and leukaemia not to speak on chronic chemical toxicity of uranium, independent of its isotopic composition. Environmental impact of extensive use of munitions with depleted uranium in the recent armed conflicts is assessed. (authors)

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

International Nuclear Information System (INIS)

Muhammad, Farhan; Majid, Asad

2009-01-01

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

Tungsten silicide contacts to polycrystalline silicon and silicon-germanium alloys

International Nuclear Information System (INIS)

Srinivasan, G.; Bain, M.F.; Bhattacharyya, S.; Baine, P.; Armstrong, B.M.; Gamble, H.S.; McNeill, D.W.

2004-01-01

Silicon-germanium alloy layers will be employed in the source-drain engineering of future MOS transistors. The use of this technology offers advantages in reducing series resistance and decreasing junction depth resulting in reduction in punch-through and SCE problems. The contact resistance of metal or metal silicides to the raised source-drain material is a serious issue at sub-micron dimensions and must be minimised. In this work, tungsten silicide produced by chemical vapour deposition has been investigated as a contact metallization scheme to both boron and phosphorus doped polycrystalline Si 1- x Ge x , with 0 ≤x ≤ 0.3. Cross bridge Kelvin resistor (CKBR) structures were fabricated incorporating CVD WSi 2 and polycrystalline SiGe. Tungsten silicide contacts to control polysilicon CKBR structures have been shown to be of high quality with specific contact resistance ρ c values 3 x 10 -7 ohm cm 2 and 6 x 10 -7 ohm cm 2 obtained to boron and phosphorus implanted samples respectively. The SiGe CKBR structures show that the inclusion of Ge yields a reduction in ρ c for both dopant types. The boron doped SiGe exhibits a reduction in ρ c from 3 x 10 -7 to 5 x 10 -8 ohm cm 2 as Ge fraction is increased from 0 to 0.3. The reduction in ρ c has been shown to be due to (i) the lowering of the tungsten silicide Schottky barrier height to p-type SiGe resulting from the energy band gap reduction, and (ii) increased activation of the implanted boron with increased Ge fraction. The phosphorus implanted samples show less sensitivity of ρ c to Ge fraction with a lowest value in this work of 3 x 10 -7 ohm cm 2 for a Ge fraction of 0.3. The reduction in specific contact resistance to the phosphorus implanted samples has been shown to be due to increased dopant activation alone

Neutron activation analysis of high pure uranium using preconcentration

International Nuclear Information System (INIS)

Sadikov, I.I.; Rakhimov, A.V.; Salimov, M.I.; Zinov'ev, V.G.

2006-01-01

Full text: Uranium and its compounds are used as nuclear fuel, and requirements for purity of initial uranium are very high. Therefore highly sensitive and multielemental analysis of uranium is required. One of such methods is neutron activation analysis (NAA). During irradiation of uranium by nuclear reactor neutrons the induced radioactivity of a sample is formed by uranium radionuclide 239 U (T 1/2 = 23,4 min.) and its daughter radionuclide 239 Np (T 1/2 = 2,39 d). Short-lived 239 U almost completely decays in 24 hours after irradiation and the radioactivity of the sample is mainly due to 239 Np and is more than 10 9 Bq for 0.1 g of uranium sample (F = 1*10 14 cm -2 s -1 , t irr . = 5 h). That is why nondestructive determination of the impurities is impossible and they should be separated from 239 Np. When irradiated uranium yields fission products - radionuclides of some elements with mass numbers 91-104 and 131-144. The main problem in NAA of uranium is to take into account correctly the influence of fission products on the analysis results. We have developed a radiochemical separation procedure for RNAA of uranium [1]. Comparing the results of analysis carried out by radiochemical NAA and instrumental NAA with preconcentration of trace elements can be used for evaluating the interference of fission products on uranium analysis results. Preconcentration of trace elements have been carried out by extraction chromatography in 'TBP - 6M HNO 3 ' system [1]. Experiments have shown that if 0.1 g uranium sample is taken for analysis (F = 1*10 14 cm -2 s -1 , t irr . =5 h) the apparent concentration of Y, Zr, Mo, Cs, La, Ce, Pr, Nd exceeds the true concentration by 2500-3000 times and so determination of these elements is not possible by radiochemical NAA. (author)

Development of new ORIGEN2 data library sets for research reactors with light water cooled oxide and silicide LEU (20 w/o) fuels based on JENDL-3.3 nuclear data

International Nuclear Information System (INIS)

Liem, Peng Hong; Sembiring, Tagor Malem

2013-01-01

Highlights: • We developed new ORIGEN2 data library sets for research reactors based on JENDL-3.3. • The sets cover oxide and silicide LEU fuels with meat density up to 4.74 g U/cm 3 . • Two kinds of data library sets are available: fuel region and non-fuel regions. • We verified the new data library sets with other codes. • We validated the new data library against a non-destructive test. -- Abstract: New sets of ORIGEN2 data library dedicated to research/testing reactors with light water cooled oxide and silicide LEU fuel plates based on JENDL-3.3 nuclear data were developed, verified and validated. The new sets are considered to be an extension of the most recent release of ORIGEN2.2UPJ code, i.e. the ORLIBJ33 library sets. The newly generated ORIGEN2 data library sets cover both oxide and silicide LEU fuels with fuel meat density range from 2.96 to 4.74 g U/cm 3 used in the present and future operation of the Indonesian 30 MWth RSG GAS research reactor. The new sets are expected applicable also for other research/testing reactors which utilize similar fuels or have similar neutron spectral indices. In addition to the traditional ORIGEN2 library sets for fuel depletion analyses in fuel regions, in the new data library sets, new ORIGEN2 library sets for irradiation/activation analyses were also prepared which cover all representative non-fuel regions of RSG GAS such as reflector elements, irradiation facilities, etc. whose neutron spectra are significantly softer than fuel regions. Verification with other codes as well as validation with a non-destructive test result showed promising results where a good agreement was confirmed

Determination of UO2F2, UO2 and UF4 in tetrafluoride of uranium samples

International Nuclear Information System (INIS)

Contreras Guzman, Ariel; Arlegui Hormazabal, Oscar

2003-01-01

The combustible elements for investigation reactors that at the present are manufacturing by the Chilean Nuclear Energy Commission (CCHEN) they are based on aluminum and silicide uranium powdered which is obtained from metallic uranium. At the present the Conversion Units, is developing the technology of transformation UF 6 in metallic Uranium, reason for which is necessary that the Chemical Analysis Laboratory have a methodology that allows to quantify the presence of UO 2 F 2 , UO 2 and UF 4 in the samples obtained in this transformation process. For this reason we are implements the methodology of sequential analysis that had been developed previously, for the Institute of Energy and Nuclear Investigations, IPEN Brasil, and to adapt it to the present conditions in the Laboratory of Chemical Analysis of the CCHEN. This method is based on the different solubilities that present those sample in front of solvents as ethanol and solutions of ammonium oxalate, what allows the separation of these compounds for a later analysis by means of the method of Davies and Gray. This method is based on the reduction of the uranium (VI) to uranium (IV) with ferrous ion amid phosphoric acid, quantifying the present uranium in the samples by means of titration with potassium dicromate. With the purpose of checking the efficiency of the method, the sum of all values of uranium coming from each compound and compares it with the total uranium of the sample (author)

Formation of copper silicides by high dose metal vapor vacuum arc ion implantation

International Nuclear Information System (INIS)

Rong Chun; Zhang Jizhong; Li Wenzhi

2003-01-01

Si(1 1 1) was implanted by copper ions with different doses and copper distribution in silicon matrix was obtained. The as-implanted samples were annealed at 300 and 540 deg. C, respectively. Formation of copper silicides in as-implanted and annealed samples were studied. Thermodynamics and kinetics of the reaction were found to be different from reaction at copper-silicon interface that was applied in conventional studies of copper-silicon interaction. The defects in silicon induced by implantation and formation of copper silicides were recognized by Si(2 2 2) X-ray diffraction (XRD)

Evaluation of refractory-metal-clad uranium nitride and uranium dioxide fuel pins after irradiation for times up to 10 450 hours at 990 C

Science.gov (United States)

Bowles, K. J.; Gluyas, R. E.

1975-01-01

The effects of some materials variables on the irradiation performance of fuel pins for a lithium-cooled space power reactor design concept were examined. The variables studied were UN fuel density, fuel composition, and cladding alloy. All pins were irradiated at about 990 C in a thermal neutron environment to the design fuel burnup. An 85-percent dense UN fuel gave the best overall results in meeting the operational goals. The T-111 cladding on all specimens was embrittled, possibly by hydrogen in the case of the UN fuel and by uranium and oxygen in the case of the UO2 fuel. Tests with Cb-1Zr cladding indicate potential use of this cladding material. The UO2 fueled specimens met the operational goals of less than 1 percent cladding strain, but other factors make UO2 less attractive than low-density UN for the contemplated space power reactor use.

Study and simulation of the behaviour under irradiation of helium in uranium dioxide; Etude et modelisation du comportement sous irradiation de l'helium dans le dioxyde d'uranium

Energy Technology Data Exchange (ETDEWEB)

Martin, G

2007-06-15

Large quantities of helium are produced from {alpha}-decay of actinides in nuclear fuels during its in-pile operating and its storage. It is important to understand the behaviour of helium in these matrix in order to well simulate the evolution and the resistance of the fuel element. During this thesis, we have used nuclear reaction analyses (NRA) to follow the evolution of the helium implanted in polycrystalline and monocrystalline uranium dioxide (UO{sub 2}). An experimental rig was developed to follow the on-line helium release in UO{sub 2} and the evolution of {sup 3}He profiles as a function of annealing temperature. An automated procedure taking into account the evolution of the depth resolution was developed. Analyses performed with a nuclear microprobe allowed to characterise the spatial distribution of helium at the grain scale and to study the influence of the sample microstructure on the helium migration. This work put into evidence the particular role of grain boundaries and irradiation defects in the helium release process. The analyse of experimental results with a diffusion model corroborates these interpretations. It allowed to determine quantitatively physical properties that characterise the helium behaviour in uranium dioxide (diffusion coefficient, activation energy..). (author)

Development of very high-density low-enriched uranium fuels

International Nuclear Information System (INIS)

Snelgrove, J.L.; Hofman, G.L.; Trybus, C.L.; Wiencek, T.C.

1997-02-01

The RERTR program has recently begun an aggressive effort to develop dispersion fuels for research and test reactors with uranium densities of 8 to 9 g U/cm 3 , based on the use of γ-stabilized uranium alloys. Fabrication development teams and facilities are being put into place and preparations for the first irradiation test are in progress. The first screening irradiations are expected to begin in late April 1997 and first results should be available by end of 1997. Discussions with potential international partners in fabrication development and irradiation testing have begun

High temperature structural silicides

International Nuclear Information System (INIS)

Petrovic, J.J.

1997-01-01

Structural silicides have important high temperature applications in oxidizing and aggressive environments. Most prominent are MoSi 2 -based materials, which are borderline ceramic-intermetallic compounds. MoSi 2 single crystals exhibit macroscopic compressive ductility at temperatures below room temperature in some orientations. Polycrystalline MoSi 2 possesses elevated temperature creep behavior which is highly sensitive to grain size. MoSi 2 -Si 3 N 4 composites show an important combination of oxidation resistance, creep resistance, and low temperature fracture toughness. Current potential applications of MoSi 2 -based materials include furnace heating elements, molten metal lances, industrial gas burners, aerospace turbine engine components, diesel engine glow plugs, and materials for glass processing

Local solid phase growth of few-layer graphene on silicon carbide from nickel silicide supersaturated with carbon

International Nuclear Information System (INIS)

Escobedo-Cousin, Enrique; Vassilevski, Konstantin; Hopf, Toby; Wright, Nick; O'Neill, Anthony; Horsfall, Alton; Goss, Jonathan; Cumpson, Peter

2013-01-01

Patterned few-layer graphene (FLG) films were obtained by local solid phase growth from nickel silicide supersaturated with carbon, following a fabrication scheme, which allows the formation of self-aligned ohmic contacts on FLG and is compatible with conventional SiC device processing methods. The process was realised by the deposition and patterning of thin Ni films on semi-insulating 6H-SiC wafers followed by annealing and the selective removal of the resulting nickel silicide by wet chemistry. Raman spectroscopy and X-ray photoelectron spectroscopy (XPS) were used to confirm both the formation and subsequent removal of nickel silicide. The impact of process parameters such as the thickness of the initial Ni layer, annealing temperature, and cooling rates on the FLG films was assessed by Raman spectroscopy, XPS, and atomic force microscopy. The thickness of the final FLG film estimated from the Raman spectra varied from 1 to 4 monolayers for initial Ni layers between 3 and 20 nm thick. Self-aligned contacts were formed on these patterned films by contact photolithography and wet etching of nickel silicide, which enabled the fabrication of test structures to measure the carrier concentration and mobility in the FLG films. A simple model of diffusion-driven solid phase chemical reaction was used to explain formation of the FLG film at the interface between nickel silicide and silicon carbide.

Formation of (Nd,Y)-silicides by sequential channeled implantation of Y and Nd ions

International Nuclear Information System (INIS)

Jin, S.; Bender, H.; Wu, M.F.; Vantomme, A.; Langouche, G.

2000-01-01

A buried hexagonal Nd 0.32 Y 0.68 Si 1.7 layer is formed by a sequential implantation of Y and Nd ions into (1 1 1)-oriented silicon wafers. The orientation relationship between the epitaxial Nd 0.32 Y 0.68 Si 1.7 and the silicon is (0 0 0 1) Nd 0.32 Y 0.68 Si 1.7 //(1 1 1) Si with Nd 0.32 Y 0.68 Si 1.7 // Si . High temperature annealing (1000 deg. C) results in a gradual transition into an orthorhombic ternary (Nd,Y)-silicide. Between the orthorhombic (Nd,Y)-silicide and the Si a preferential orientation relationship exists: (1 1 0) orth //(1 1(bar) 0) Si with orth // Si . However, as not all orthorhombic silicide grains follow this epitaxial relationship, the minimum yield in the Rutherford backscattering spectrometry (RBS) spectrum increases compared to the results after a low temperature annealing

Formation of (Nd,Y)-silicides by sequential channeled implantation of Y and Nd ions

Science.gov (United States)

Jin, S.; Bender, H.; Wu, M. F.; Vantomme, A.; Langouche, G.

2000-03-01

A buried hexagonal Nd0.32Y0.68Si1.7 layer is formed by a sequential implantation of Y and Nd ions into (1 1 1)-oriented silicon wafers. The orientation relationship between the epitaxial Nd0.32Y0.68Si1.7 and the silicon is (0 0 0 1)Nd0.32Y0.68Si1.7//(1 1 1)Si with Nd0.32Y0.68Si1.7//Si. High temperature annealing (1000°C) results in a gradual transition into an orthorhombic ternary (Nd,Y)-silicide. Between the orthorhombic (Nd,Y)-silicide and the Si a preferential orientation relationship exists: (1 1 0)orth//(1 1¯ 0)Si with orth//Si. However, as not all orthorhombic silicide grains follow this epitaxial relationship, the minimum yield in the Rutherford backscattering spectrometry (RBS) spectrum increases compared to the results after a low temperature annealing.

Fracture of niobium-base silicide coated alloy

International Nuclear Information System (INIS)

Davydova, A.D.; Zotov, Yu.P.; Ivashchenko, O.V.; Kushnareva, N.P.; Yarosh, I.P.

1990-01-01

Mechanical properties and character of fracture of Nb-W-Mo-Zr-C alloy composition with complex by composition and structure silicide coating under different states of stage-by-stage coating are studied. Structural features, character of fracture from ductile to quasibrittle transcrystalline one and, respectively, the composition plasticity level are defined by interrelation of fracture processes in coating, matrix plastic flow and possibility and way of stress relaxation on their boundary

Development of 99Mo isotope production targets employing uranium metal foils

International Nuclear Information System (INIS)

Hofman, G.L.; Wiencek, T.C.; Wood, E.L.; Snelgrove, J.L.

1997-01-01

The Reduced Enrichment Research and Test Reactor Program has continued its effort in the past 3 yr to develop use of low-enriched uranium (LEU) to produce the fission product 99 Mo. This work comprises both target and chemical processing development and demonstration. Two major target systems are now being used to produce 99 Mo with highly enriched uranium-one employing research reactor fuel technology (either uranium-aluminum alloy or uranium aluminide-aluminum dispersion) and the other using a thin deposit of UO 2 on the inside of a stainless steel (SST) tube. This paper summarizes progress in irradiation testing of targets based on LEU uranium metal foils. Several targets of this type have been irradiated in the Indonesian RSG-GAS reactor operating at 22.5 MW

Evaluation of plutonium, uranium, and thorium use in power reactor fuel cycles

International Nuclear Information System (INIS)

Kasten, P.R.; Homan, F.J.

1977-01-01

The increased cost of uranium and separative work has increased the attractiveness of plutonium use in both uranium and thorium fuel cycles in thermal reactors. A technology, fuel utilization, and economic evaluation is given for uranium and thorium fuel cycles in various reactor types, along with the use of plutonium and 238 U. Reactors considered are LWRs, HWRs, LWBRs, HTGRs, and FBRs. Key technology factors are fuel irradiation performance and associated physical property values. Key economic factors are unit costs for fuel fabrication and reprocessing, and for refabrication of recycle fuels; consistent cost estimates are utilized. In thermal reactors, the irradiation performance of ceramic fuels appears to be satisfactory. At present costs for uranium ore and separative work, recycle of plutonium with thorium rather than uranium is preferable from fuel utilization and economic viewpoints. Further, the unit recovery cost of plutonium is lower from LWR fuels than from natural-uranium HWR fuels; use of LWR product permits plutonium/thorium fueling to compete with uranium cycles. Converting uranium cycles to thorium cycles increases the energy which can be extracted from a given uranium resource. Thus, additional fuel utilization improvement can be obtained by fueling all thermal reactors with thorium, but this requires use of highly enriched uranium; use of 235 U with thorium is most economic in HTGRs followed by HWRs and then LWRs. Marked improvement in long-term fuel utilization can be obtained through high thorium loadings and short fuel cycle irradiations as in the LWBR, but this imposes significant economic penalties. Similar operating modes are possible in HWRs and HTGRs. In fast reactors, use of the plutonium-uranium cycle gives advantageous fuel resource utilization in both LMFBRs and GCFRs; use of the thorium cycle provides more negative core reactivity coefficients and more flexibility relative to use of recycle fuels containing uranium of less than 20

Development of very-high-density low-enriched uranium fuels

International Nuclear Information System (INIS)

Snegrove, J.L.; Hofmann, G.L.; Trybus, C.L.; Wiencek, T.C.

1997-01-01

The RERTR (=Reduced Enrichment for Research and Test Reactors) program has begun an aggressive effort to develop dispersion fuels for research and test reactors with uranium densities of 8 to 9 g U/cm 3 , based on the use of γ-stabilized uranium alloys. Fabrication development teams and facilities are being put into place, and preparations for the first irradiation test are in progress. The first screening irradiations are expected to begin in late April 1997 and the first results should be available by the end of 1997. Discussions with potential international partners in fabrication development and irradiation testing have begun. (author)

Effect of Utilization of Silicide Fuel with the Density 4.8 gU/cc on the Kinetic Parameters of RSG-GAS Reactor

International Nuclear Information System (INIS)

Setiyanto; Sembiring, Tagor M.; Pinem, Surian

2007-01-01

Presently, the RSG-GAS reactor using silicide fuel element of 2.96 gU/cc. For increasing reactor operation time, its planning to change to higher density fuel. The kinetic calculation of silicide core with density 4.8 gU/cc has been carried out, since it has an influence on the reactor operation safety. The calculated kinetic parameters are the effective delayed neutron fraction, the delayed neutron decay constant, prompt neutron lifetime and feedback reactivity coefficient very important for reactor operation safety. the calculation is performed in 2-dimensional neutron diffusion-perturbation method using modified Batan-2DIFF code. The calculation showed that the effective delayed neutron fraction is 7. 03256x10 -03 , total delay neutron time constant is 7.85820x10 -02 s -1 and the prompt neutron lifetime is 55.4900 μs. The result of prompt neutron lifetime smaller 10 % compare with silicide fuel of 4.8 gU/cc. The calculated results showed that all of the feedback reactivity coefficient silicide core 4.8 gU/cc is negative. Totally, the feedback reactivity coefficient of silicide fuel of 4.8 gU/cc is 10% less than that of silicide fuel of 2.96 gU/cc. The results shown that kinetic parameters result decrease compared with the silicide core with density 2.96 gU/cc, but no significant influence in the RSG-GAS reactor operation. (author)

Radiation damage of metal uranium; Radijaciono ostecenje metalnog urana

Energy Technology Data Exchange (ETDEWEB)

Mihajlovic, A [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

1965-11-15

This report is concerned with the role of dispersion second phase in uranium and burnup rate. The role of dispersion phases in radiation stability of metal uranium was studies by three methods: variation of electric conductivity dependent on the neutron flux and temperature of pure uranium for different states of dispersion second phase; influence of dispersion phase on the radiation creep; transmission electron microscopy of fresh and irradiated uranium.

Phase transformations in Higher Manganese Silicides

Energy Technology Data Exchange (ETDEWEB)

Allam, A. [MADIREL, UMR 7246 CNRS - Universite Aix-Marseille, av Normandie-Niemen, 13397 Marseille Cedex 20 (France); IM2NP, UMR 7334 CNRS - Universite Aix-Marseille, av Normandie-Niemen, Case 142, 13397 Marseille Cedex 20 (France); Boulet, P. [MADIREL, UMR 7246 CNRS - Universite Aix-Marseille, av Normandie-Niemen, 13397 Marseille Cedex 20 (France); Nunes, C.A. [Departamento de Engenharia de Materiais (DEMAR), Escola de Engenharia de Lorena (EEL), Universidade de Sao Paulo - USP, Caixa Postal 116, 12600-970 Lorena, Sao Paulo (Brazil); Sopousek, J.; Broz, P. [Masaryk University, Faculty of Science, Department of Chemistry, Kolarska 2, 611 37 Brno (Czech Republic); Masaryk University, Central European Institute of Technology, CEITEC, Kamenice 753/5, 625 00 Brno (Czech Republic); Record, M.-C., E-mail: m-c.record@univ-cezanne.fr [IM2NP, UMR 7334 CNRS - Universite Aix-Marseille, av Normandie-Niemen, Case 142, 13397 Marseille Cedex 20 (France)

2013-02-25

Highlights: Black-Right-Pointing-Pointer The phase transitions of the Higher Manganese Silicides were investigated. Black-Right-Pointing-Pointer The samples were characterised by XRD, DTA and DSC. Black-Right-Pointing-Pointer Mn{sub 27}Si{sub 47} is the stable phase at room temperature and under atmospheric pressure. Black-Right-Pointing-Pointer At around 800 Degree-Sign C, Mn{sub 27}Si{sub 47} is transformed into Mn{sub 15}Si{sub 26}. Black-Right-Pointing-Pointer The phase transition is of a second order. - Abstract: This work is an investigation of the phase transformations of the Higher Manganese Silicides in the temperature range [100-1200 Degree-Sign C]. Several complementary experimental techniques were used, namely in situ X-ray Diffraction (XRD), Differential Thermal Analysis (DTA) and Differential Scanning Calorimetry (DSC). The evolution of both the lattice parameters and the thermal expansion coefficients was determined from in situ XRD measurements. The stability of the samples was investigated by thermal analysis (DTA) and Cp measurements (DSC). This study shows that Mn{sub 27}Si{sub 47} which is the stable phase at room temperature and under atmospheric pressure undergoes a phase transformation at around 800 Degree-Sign C. Mn{sub 27}Si{sub 47} is transformed into Mn{sub 15}Si{sub 26}. This phase transformation seems to be of a second order one. Indeed it was not evidenced by DTA and by contrast it appears on the Cp curve.

Logging technique for assaying for uranium in rocks

International Nuclear Information System (INIS)

Givens, W.W.

1973-01-01

A uranium exploration technique is described for determining the uranium content of a formation traversed by borehole. A delayed fission neutron assay log is obtained by irradiating the formation with repetitive bursts of fast neutrons and detecting delayed neutrons resulting from neutron fission of uranium at time intervals between the fast neutron bursts and after dissipation of the neutrons originating in the bursts. In addition, a response log is obtained by irradiating the formation with a source of fast neutrons whereby the neutrons from this source are moderated in the formation to lower energy levels and are subject to absorption. Secondary radiation attendant to these lower energy neutrons is recorded in order to obtain a log representative of the response of the formation to moderation and absorption of the neutrons. The two logs thus obtained are correlated in order to determine a corrected value of uranium content of the formation. (author)

The electronic structure of 4d and 5d silicides

NARCIS (Netherlands)

Speier, W.; Kumar, L.; Sarma, D.D.; Groot, R.A. de; Fuggle, J.C.

1989-01-01

A systematic experimental and theoretical study of the electronic structure of stoichiometric silicides with Nb, Mo, Ta and W is presented. We have employed x-ray photoemission and bremsstrahlung isochromat spectroscopy as experimental techniques and interpreted the measured data by calculation of

Uranium fluoride and metallic uranium as target materials for heavy-element experiments at SHIP

Energy Technology Data Exchange (ETDEWEB)

Kindler, Birgit [Gesellschaft fuer Schwerionenforschung (GSI), Planckstrasse 1, D-64291 Darmstadt (Germany)], E-mail: b.kindler@gsi.de; Ackermann, Dieter; Hartmann, Willi; Hessberger, Fritz Peter; Hofmann, Sigurd; Huebner, Annett; Lommel, Bettina; Mann, Rido; Steiner, Jutta [Gesellschaft fuer Schwerionenforschung (GSI), Planckstrasse 1, D-64291 Darmstadt (Germany)

2008-06-01

In this contribution we describe the production and application of uranium targets for synthesis of heavy elements. The targets are prepared from uranium fluoride (UF{sub 4}) and from metallic uranium with thin carbon foils as backing. Targets of UF{sub 4} were produced by thermal evaporation in a similar way as the frequently applied targets out of Bi, Bi{sub 2}O{sub 3}, Pb, PbS, SmF{sub 3}, and NdF{sub 3,} prepared mostly from isotopically enriched material [Birgit Kindler, et al., Nucl. Instr. and Meth. A 561 (2006) 107; Bettina Lommel, et al., Nucl. Instr. and Meth. A 561 (2006) 100]. In order to use more intensive beams and to avoid scattering of the reaction products in the target, metallic uranium is favorable. However, evaporation of metallic uranium is not feasible at a sustainable yield. Therefore, we established magnetron sputtering of metallic uranium. We describe production and properties of these targets. First irradiation tests show promising results.

The recovery of 99Mo from solutions of irradiated Uranium using a column with nanoparticles of Titanium Dioxide

International Nuclear Information System (INIS)

Androne, G. E.; Petre, M.; Lazar, C. G.

2016-01-01

Molyibdenum-99 (T½ = 66.02 h) decays by beta emission to 99 Tcm (T½ = 6.02 h). The latter nuclide is used in many nuclear medicine applications. The 99 Mo is produced from irradiated high (HEU) or low (LEU) enriched uranium. In this work a sensitive and selective method for recovering Mo from uranium solution, using a column with titanium dioxide nanoparticles, is developed. The titanium dioxide (TiO 2 ) nanoparticles were synthesized via sol-gel method using titanium tetra-chloride as starting material and urea as a reacting medium. A 40 ml uranium solution containing 450 g/L uranyl nitrate, 1 M HNO 3 , and 4 mg Mo was loaded on a column containing 6 g of TiO 2 sorbent at 75°C. After loading, the column was washed with 1 M HNO 3 and H 2 O. Mo was stripped from the column with 0.1 M NaOH at 25°C. The ICP-MS results indicate that 80-95% of the initial mass of Mo was loaded on the column, and 90-94% of this quantity was recovered in the strip fraction. (authors)

77 FR 73056 - Application for a License To Export High-Enriched Uranium

Science.gov (United States)

2012-12-07

... NUCLEAR REGULATORY COMMISSION Application for a License To Export High-Enriched Uranium Pursuant... Complex. Uranium (93.2%). uranium-235 at CERCA AREVA Romans October 10, 2012 contained in 6.2 in France and to October 12, 2012 kilograms irradiate targets at XSNM3729 uranium. the BR-2 Research 11006053...

77 FR 73055 - Application for a License To Export High-Enriched Uranium

Science.gov (United States)

2012-12-07

... NUCLEAR REGULATORY COMMISSION Application for a License To Export High-Enriched Uranium Pursuant.... Security Complex. Uranium uranium-235 at CERCA AREVA October 10, 2012 (93.35%). contained in Romans in France October 12, 2012 10.1 kilograms and to irradiate XSNM3730 uranium. targets at the HFR 11006054...

Burn-up determination of irradiated uranium oxide by means of direct gama spectrometry and by radiochemical method

International Nuclear Information System (INIS)

Cunha, I.I.L.; Nastasi, M.J.C.; Lima, F.W.

1981-09-01

The burn-up of thermal neutrons irradiated U 3 O 8 (natural uranium) samples has been determined by using both direct gamma spectrometry and radiochemical methods and the results obtained were compared. The fission products 144 Ce, 103 Ru, 106 Ru, 137 Cs and 95 Zr were chosen as burn-up monitors. In order to isolate the radioisotopes chosen as monitors, a radiochemical separation procedure has been established, in which the solvent extraction technique was used to separate cerium, cesium and ruthenium one from the other and all of them from uranium. The separation between zirconium and niobium and of both elements from the other radioisotopes and uranium was accomplished by means of adsorption on a silica-gel column, followed by selective elution of zirconium and of niobium. When use was made of the direct gamma-ray spectrometry method, the radioactivity of each nuclide of interest was measured in presence of all others. For this purpose use was made of gamma-ray spectrometry and of a Ge-Li detector. Comparison of burn-up values obtained by both methods was made by means of Student's 't' test, and this showed that results obtained in each case are statistically equal. (Author) [pt

Silicon-germanium and platinum silicide nanostructures for silicon based photonics

Science.gov (United States)

Storozhevykh, M. S.; Dubkov, V. P.; Arapkina, L. V.; Chizh, K. V.; Mironov, S. A.; Chapnin, V. A.; Yuryev, V. A.

2017-05-01

This paper reports a study of two types of silicon based nanostructures prospective for applications in photonics. The first ones are Ge/Si(001) structures forming at room temperature and reconstructing after annealing at 600°C. Germanium, being deposited from a molecular beam at room temperature on the Si(001) surface, forms a thin granular film composed of Ge particles with sizes of a few nanometers. A characteristic feature of these films is that they demonstrate signs of the 2 x 1 structure in their RHEED patterns. After short-term annealing at 600°C under the closed system conditions, the granular films reconstruct to heterostructures consisting of a Ge wetting layer and oval clusters of Ge. A mixed type c(4x2) + p(2x2) reconstruction typical to the low-temperature MBE (Tgr Ge. The other type of the studied nanostructures is based on Pt silicides. This class of materials is one of the friendliest to silicon technology. But as silicide film thickness reaches a few nanometers, low resistivity becomes of primary importance. Pt3Si has the lowest sheet resistance among the Pt silicides. However, the development of a process of thin Pt3Si films formation is a challenging task. This paper describes formation of a thin Pt3Si/Pt2Si structures at room temperature on poly-Si films. Special attention is paid upon formation of poly-Si and amorphous Si films on Si3N4 substrates at low temperatures.

Mechanochemical synthesis and spark plasma sintering of the cerium silicides

Energy Technology Data Exchange (ETDEWEB)

Alanko, Gordon A.; Jaques, Brian; Bateman, Allyssa [Department of Materials Science and Engineering, College of Engineering, Boise State University, 1910 University Drive, Boise, ID 83725 (United States); Butt, Darryl P., E-mail: darrylbutt@boisestate.edu [Department of Materials Science and Engineering, College of Engineering, Boise State University, 1910 University Drive, Boise, ID 83725 (United States); Center for Advanced Energy Studies, 995 University Boulevard, Idaho Falls, ID 83401 (United States)

2014-12-15

Highlights: • Ce{sub 5}Si{sub 3}, Ce{sub 3}Si{sub 2}, CeSi, CeSi{sub 2−x} and CeSi{sub 2} were mechanochemically synthesized. • Temperature and pressure were monitored to investigate reaction progress. • All syntheses proceeded through a MSR event followed by rapid solid-state diffusion. • Milling time before MSR correlates well with effective heat of formation. • Some synthesized material was densified by spark plasma sintering. - Abstract: The cerium silicides, Ce{sub 5}Si{sub 3}, Ce{sub 3}Si{sub 2}, CeSi, CeSi{sub 2−y}, and CeSi{sub 2−x}, have been prepared from the elements by mechanochemical processing in a planetary ball mill. Preparation of the cerium silicide Ce{sub 5}Si{sub 4} was unsuccessfully attempted and potential reasons for this are discussed. Temperature and pressure of the milling vial were monitored in situ to gain insight into the mechanochemical reaction kinetics, which include a mechanically-induced self-propagating reaction (MSR). Some prepared powders were consolidated by spark plasma sintering to high density. Starting materials, as-milled powders, and consolidated samples were characterized by X-ray diffraction, scanning electron microscopy, and energy dispersive spectroscopy. The results obtained help elucidate key questions in mechanochemical processing of intermetallics, showing first phase formation similar to thin films, MSR ignition times that are composition- and milling speed-dependent, and sensitivity of stable compound formation on the impact pressure. The results demonstrate mechanochemical synthesis as a viable technique for rare earth silicides.

Towards the improvement of the oxidation resistance of Nb-silicides in situ composites: A solid state diffusion approach

International Nuclear Information System (INIS)

Mathieu, S.; Knittel, S.; François, M.; Portebois, L.; Mathieu, S.; Vilasi, M.

2014-01-01

Highlights: •Local equilibrium is attained during oxidation at phase boundaries (steady state conditions). •A solid state diffusion model explains the oxidation mechanism of Nb-silicides composites. •The Nb ss fraction is not the only parameters governing the oxidation rate of Nb-silicides. •Aluminium increases the thermodynamic activity of Si in the Nb-silicides composites. •The results indicate the need to develop a Nb–Ti–Hf–Al–Cr–Si thermodynamic database. -- Abstract: The present study focuses on the oxidation mechanism of Nb-silicide composites and on the effect of the composition on the oxidation rate at 1100 °C. A theoretical approach is proposed based on experimental results and used to optimise the oxidation resistance. The growth model based on multiphase diffusion was experimentally tested and confirmed by manufacturing seven composites with different compositions. It was also found that the effect of the composition has to be evaluated at 1100 °C within a short time duration (50 h), where the oxide scale and the internal oxidation zone both grow according to parabolic kinetics

Quantitative EPMA of Nano-Phase Iron-Silicides in Apollo 16 Lunar Regolith

Science.gov (United States)

Gopon, P.; Fournelle, J.; Valley, J. W.; Pinard, P. T.; Sobol, P.; Horn, W.; Spicuzza, M.; Llovet, X.; Richter, S.

2013-12-01

Until recently, quantitative EPMA of phases under a few microns in size has been extremely difficult. In order to achieve analytical volumes to analyze sub-micron features, accelerating voltages between 5 and 8 keV need to be used. At these voltages the normally used K X-ray transitions (of higher Z elements) are no longer excited, and we must rely of outer shell transitions (L and M). These outer shell transitions are difficult to use for quantitative EPMA because they are strongly affected by different bonding environments, the error associated with their mass attenuation coefficients (MAC), and their proximity to absorption edges. These problems are especially prevalent for the transition metals, because of the unfilled M5 electron shell where the Lα transition originates. Previous studies have tried to overcome these limitations by using standards that almost exactly matched their unknowns. This, however, is cumbersome and requires accurate knowledge of the composition of your sample beforehand, as well as an exorbitant number of well characterized standards. Using a 5 keV electron beam and utilizing non-standard X-ray transitions (Ll) for the transition metals, we are able to conduct accurate quantitative analyses of phases down to ~300nm. The Ll transition in the transition metals behaves more like a core-state transition, and unlike the Lα/β lines, is unaffected by bonding effects and does not lie near an absorption edge. This allows for quantitative analysis using standards do not have to exactly match the unknown. In our case pure metal standards were used for all elements except phosphorus. We present here data on iron-silicides in two Apollo 16 regolith grains. These plagioclase grains (A6-7 and A6-8) were collected between North and South Ray Craters, in the lunar highlands, and thus are associated with one or more large impact events. We report the presence of carbon, nickel, and phosphorus (in order of abundance) in these iron-silicide phases

The discovery of uranium fission

International Nuclear Information System (INIS)

Brix, P.

1990-01-01

Uranium was discovered 200 years ago. Its radioactive character was first demonstrated in 1896 and two years later radium was extracted from uranium minerals. In 1911 studies with alpha rays from radioactive decay led to the unexpected discovery of the atomic nucleus. Exposure of beryllium to alpha rays yielded neutrons, first detected in 1932. Starting in 1934, neutron irradiation of uranium produced radioactive substances erroneously attributed to transuranium elements but with confusing properties. Painstaking experiments by chemists left no doubt on 17 December 1938 that barium was produced by these irradiations: the neutrons had split some uranium nuclei. The physics of the fission process was understood two weeks later; after a few months, neutron multiplication was found to be probable. This review deals with the eminent scientists involved, their successes, errors and disappointments, and the unexpected insights which occurred on the paths and detours of scientific research. It is, therefore, instructive also to discuss how fission was not discovered. The momentous discovery must be considered inevitable; the great tragedy was that Germany started World War II just at the time when the possibility of nuclear chain reactions and bombs became known. The consequences and anxieties that remain after 50 years of nuclear fission demand that mankind act with reason and conscience to maintain peace. (author)

The formation of magnetic silicide Fe3Si clusters during ion implantation

Science.gov (United States)

Balakirev, N.; Zhikharev, V.; Gumarov, G.

2014-05-01

A simple two-dimensional model of the formation of magnetic silicide Fe3Si clusters during high-dose Fe ion implantation into silicon has been proposed and the cluster growth process has been computer simulated. The model takes into account the interaction between the cluster magnetization and magnetic moments of Fe atoms random walking in the implanted layer. If the clusters are formed in the presence of the external magnetic field parallel to the implanted layer, the model predicts the elongation of the growing cluster in the field direction. It has been proposed that the cluster elongation results in the uniaxial magnetic anisotropy in the plane of the implanted layer, which is observed in iron silicide films ion-beam synthesized in the external magnetic field.

The formation of magnetic silicide Fe3Si clusters during ion implantation

International Nuclear Information System (INIS)

Balakirev, N.; Zhikharev, V.; Gumarov, G.

2014-01-01

A simple two-dimensional model of the formation of magnetic silicide Fe 3 Si clusters during high-dose Fe ion implantation into silicon has been proposed and the cluster growth process has been computer simulated. The model takes into account the interaction between the cluster magnetization and magnetic moments of Fe atoms random walking in the implanted layer. If the clusters are formed in the presence of the external magnetic field parallel to the implanted layer, the model predicts the elongation of the growing cluster in the field direction. It has been proposed that the cluster elongation results in the uniaxial magnetic anisotropy in the plane of the implanted layer, which is observed in iron silicide films ion-beam synthesized in the external magnetic field

Experimental studies of thermal and chemical interactions between oxide and silicide nuclear fuels with water

Energy Technology Data Exchange (ETDEWEB)

farahani, A.A.; Corradini, M.L. [Univ. of Wisconsi, Madison, WI (United States)

1995-09-01

Given some transient power/cooling mismatch is a nuclear reactor and its inability to establish the necessary core cooling, energetic fuel-coolant interactions (FCI`s commonly called `vapor explosions`) could occur as a result of the core melting and coolant contact. Although a large number of studies have been done on energetic FCI`s, very few experiments have been performed with the actual fuel materials postulated to be produced in severe accidents. Because of the scarcity of well-characterized FCI data for uranium allows in noncommercial reactors (cermet and silicide fuels), we have conducted a series of experiments to provide a data base for the foregoing materials. An existing 1-D shock-tube facility was modified to handle depleted radioactive materials (U{sub 3}O{sub 8}-Al, and U{sub 3}Si{sub 2}-Al). Our objectives have been to determine the effects of the initial fuel composition and temperature and the driving pressure (triggering) on the explosion work output, dynamic pressures, transient temperatures, and the hydrogen production. Experimental results indicate limited energetics, mainly thermal interactions, for these fuel materials as compared to aluminum where more chemical reactions occur between the molten aluminum and water.

NSRR experiment with un-irradiated uranium-zirconium hydride fuel. Design, fabrication process and inspection data of test fuel rod

Energy Technology Data Exchange (ETDEWEB)

Sasajima, Hideo; Fuketa, Toyoshi; Ishijima, Kiyomi; Kuroha, Hiroshi; Ikeda, Yoshikazu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Aizawa, Keiichi

1998-08-01

An experiment plan is progressing in the Nuclear Safety Research Reactor (NSRR) to perform pulse-irradiation with uranium-zirconium hydride (U-ZrH{sub x}) fuel. This fuel is widely used in the training research and isotope production reactor of GA (TRIGA). The objectives of the experiment are to determine the fuel rod failure threshold and to investigate fuel behavior under simulated reactivity initiated accident (RIA) conditions. This report summarizes design, fabrication process and inspection data of the test fuel rods before pulse-irradiation. The experiment with U-ZrH{sub x} fuel will realize precise safety evaluation, and improve the TRIGA reactor performance. The data to be obtained in this program will also contribute development of next-generation TRIGA reactor and its safety evaluation. (author)

Micrographic examination of irradiated uranium

International Nuclear Information System (INIS)

Bloch, J.; Bourgues, J.

1959-01-01

An installation for the metallographic examination of the EL2 reactor uranium rods has been put into service since September 1955 at the High Activity Lab of CEA Saclay center. This paper describes the apparatuses used for these examinations and presents the results of the micrographic studies of the first three rod assemblies removed from EL2 so far. Reprint of a paper published in Revue de Metallurgie, LV, No. 4, 1958, p. 339-344 [fr

The present state of laser isotope separation of uranium

International Nuclear Information System (INIS)

Tashiro, Hideo; Nemoto, Koshichi.

1994-01-01

As the methods of uranium enrichment, gas diffusion method and centrifugal separation method in which power consumption is less and the cost is low have been carried out. On the other hand, as the future technology, the research and development of laser isotope separation technology have been carried out. There are the atomic laser separation process in which the laser beam of visible light is irradiated to atomic state uranium and the molecular laser separation process in which far infrared laser beam is irradiated to uranium hexafluoride molecules. The atomic process is divided into three steps, that is, the processes of uranium evaporation, the reaction of uranium with laser beam and the recovery of enriched uranium. The principle of the laser separation is explained. The state of development of laser equipment and separation equipment is reported. The principle and the present state of development of the molecular separation process which consists of the cooling of UF 6 gas, the generation of high power 16 μm laser pulses and the collection of the reaction product are explained. The present state of both processes in foreign countries is reported. (K.I.)

Optical anisotropy of quasi-1D rare-earth silicide nanostructures on Si(001)

Energy Technology Data Exchange (ETDEWEB)

Chandola, S., E-mail: sandhya.chandola@isas.de [Leibniz-Institut für Analytische Wissenschaften – ISAS – e.V., Schwarzschildstraße 8, 12489 Berlin (Germany); Speiser, E.; Esser, N. [Leibniz-Institut für Analytische Wissenschaften – ISAS – e.V., Schwarzschildstraße 8, 12489 Berlin (Germany); Appelfeller, S.; Franz, M.; Dähne, M. [Institut für Festkörperphysik, Technische Universität Berlin, Hardenbergstraße 36, 10623 Berlin (Germany)

2017-03-31

Highlights: • Reflectance anisotropy spectroscopy (RAS) is capable of distinguishing optically between the semiconducting wetting layer and the metallic nanowires of rare earth (Tb and Dy) silicide nanostructures grown on vicinal Si(001). • The spectra of the wetting layer show a distinctive line shape with a large peak appearing at 3.8 eV, which is assigned to the formation of 2 × 3 and 2 × 4-like subunits of the 2 × 7 reconstruction. The spectra of the metallic nanowires show peaks at the E{sub 1} and E{sub 2} transitions of bulk Si which is assigned to strong substrate strain induced by the nanowires. • The optical anisotropy of the Tb nanowires is larger than for the Dy nanowires, which is related to the preferential formation of more strained bundles as well as larger areas of clean Si surfaces in the case of Tb. • RAS is shown to be a powerful addition to surface science techniques for studying the formation of rare-earth silicide nanostructures. Its surface sensitivity and rapidity of response make it an ideal complement to the slower but higher resolution of scanning probes of STM and AFM. - Abstract: Rare earth metals are known to interact strongly with Si(001) surfaces to form different types of silicide nanostructures. Using STM to structurally characterize Dy and Tb silicide nanostructures on vicinal Si(001), it will be shown that reflectance anisotropy spectroscopy (RAS) can be used as an optical fingerprint technique to clearly distinguish between the formation of a semiconducting two-dimensional wetting layer and the metallic one-dimensional nanowires. Moreover, the distinctive spectral features can be related to structural units of the nanostructures. RAS spectra of Tb and Dy nanostructures are found to show similar features.

Cross-Bridge Kelvin Resistor (CBKR) structures for silicide-semiconductor junctions characterization

NARCIS (Netherlands)

Stavitski, N.; van Dal, M.J.H.; Klootwijk, J.H.; Wolters, Robertus A.M.; Kovalgin, Alexeij Y.; Schmitz, Jurriaan

2006-01-01

Analyzing the contact geometry factors for the conventional CBKR structures, it appeared that the contact geometries conventionally used for the metal-to-silicide contact resistance measurements were not always satisfactory to reveal the specific contact resistance values. To investigate these

Synthesis of molybdenum borides and molybdenum silicides in molten salts and their oxidation behavior in an air-water mixture

NARCIS (Netherlands)

Kuznetsov, S.A.; Kuznetsova, S.V.; Rebrov, E.V.; Mies, M.J.M.; Croon, de M.H.J.M.; Schouten, J.C.

2005-01-01

The formation of various coatings in molybdenum-boron and molybdenum-silicon systems was investigated. Boronizing and siliciding treatments were conducted in molten salts under inert gas atm. in the 850-1050 DegC temp. range for 7 h. The presence of boride (e.g. Mo2B, MoB, Mo2B5) and silicide

Health and environmental problems of using depleted uranium

International Nuclear Information System (INIS)

Matousek, J.

2006-01-01

In the 1970's, a core of depleted uranium (DU) began to be introduced into the break through anti-tank munitions to enhance their effectiveness. The health and environmental threats of DU stem from the pyrophoric character of the core, burnt when penetrating armour to an aerosol of uranium oxides deposited in tissues after inhalation or ingestion. Their delayed effects are due to internal alpha irradiation by daughter products and toxicity of uranium. (authors)

Influence of impurities on silicide contact formation

International Nuclear Information System (INIS)

Kazdaev, Kh.R.; Meermanov, G.B.; Kazdaev, R.Kh.

2002-01-01

Research objectives of this work are to investigate the influence of light impurities implantation on peculiarities of the silicides formation in molybdenum monocrystal implanted by silicon, and in molybdenum films sputtered on silicon substrate at subsequent annealing. Implantation of the molybdenum samples was performed with silicon ions (90 keV, 5x10 17 cm -2 ). Phase identification was performed by X ray analysis with photographic method of registration. Analysis of the results has shown the formation of the molybdenum silicide Mo 3 Si at 900 deg. C. To find out the influence of impurities present in the atmosphere (C,N,O) on investigated processes we have applied combined implantation. At first, molybdenum was implanted with ions of the basic component (silicon) and then -- with impurities ions. Acceleration energies (40keV for C, 45 keV for N and 50 keV for O) were chosen to obtain the same distribution profiles for basic and impurities ions. Ion doses were 5x10 17 cm -2 for Si-ions and 5x10 16 cm -2 - for impurities. The most important results are reported here. The first, for all three kinds of impurities the decreased formation temperatures of the phase Mo 3 Si were observed; in the case of C and N it was ∼100 deg. and in the case of nitrogen - ∼200 deg. Further, simultaneously with the Mo 3 Si phase, the appearance of the rich-metal phase Mo 5 Si 3 was registered (not observed in the samples without additional implantation). In case of Mo/Si-structure, the implantation of the impurities (N,O) was performed to create the peak concentration (∼4at/%) located in the middle of the molybdenum film (∼ 150nm) deposited on silicon substrate. Investigation carried out on unimplanted samples showed the formation of the silicide molybdenum MoSi 2 , observed after annealing at temperatures 900/1000 deg. C, higher than values 500-600 deg. C reported in other works. It is discovered that electrical conductivity of Mo 5 Si 3 -films synthesized after impurities

78 FR 72123 - Request To Amend a License to Export High-Enriched Uranium

Science.gov (United States)

2013-12-02

... NUCLEAR REGULATORY COMMISSION Request To Amend a License to Export High-Enriched Uranium Pursuant... manufacture HEU targets in Belgium. National Nuclear Security Uranium (HEU) uranium France for irradiation in... 5.8 kg of U- 235 contained in 6.2 kg uranium to a new cumulative total of 12.615 kg of U-235...

Performance and fuel cycle cost analysis of one Janus 30 conceptual design for several fuel element design options

Energy Technology Data Exchange (ETDEWEB)

Nurdin, Martias [Research Centre for Nuclear Techniques, National Atomic Energy Agency (Indonesia); Matos, J E; Freese, K E [RERTR Program, Argonne National Laboratory (United States)

1983-09-01

The performance and fuel cycle costs for a 25 MW, JANUS 30 reactor conceptual design by INTERATOM, Federal Republic of Germany, for BATAN, Republic of Indonesia have been studied using 19.75% enriched uranium in four fuel element design options. All of these fuel element designs have either been proposed by INTERATOM for various reactors or are currently in use with 93% enriched uranium in reactors in the Federal Republic of Germany. Aluminide, oxide, and silicide fuels were studied for selected designs using the range of uranium densities that are either currently qualified or are being developed and demonstrated internationally. These uranium densities include 1.7-2.3 g/cm{sup 3} in aluminide fuel, 1.7-3.2 g/cm{sup 3} in oxide fuel, and 2.9-6.8 g/cm{sup 3} in silicide fuel. As of November 1982) both the aluminide and the oxide fuels with about 1.7 g U/cm{sup 3} are considered to be fully-proven for licensing purposes. Irradiation screening and proof testing of fuels with uranium densities greater than 1.7 g/cm{sup 3} are currently in progress, and these tests need to be completed in order to obtain licensing authorization for routine reactor use. To assess the long-term fuel adaptation strategy as well as the present fuel acceptance, reactor performance and annual fuel cycle costs were computed for seventeen cases based on a representative end-of-cycle excess reactivity and duty factor. In addition, a study was made to provide data for evaluating the trade-off between the increased safety associated with thicker cladding and the economic penalty due to increased fuel consumption. (author)

Reuse of ammonium fluoride generated in the uranium hexafluoride conversion; Reutilizacao do fluoreto de amonio gerado na reconversao do hexafluoreto de uranio

Energy Technology Data Exchange (ETDEWEB)

Silva Neto, J.B.; Carvalho, E.F. Urano de; Durazzo, M., E-mail: jbsneto@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Riella, H.G [Universidade Federal de Santa Catarina (UFSC), Florianopolis, SC (Brazil)

2010-07-01

The Nuclear Fuel Centre of IPEN / CNEN - SP develops and manufactures dispersion fuel with high uranium concentration to meet the demand of the IEA-R1 reactor and future research reactors planned to be constructed in Brazil. The fuel uses uranium silicide (U{sub 3}Si{sub 2}) dispersed in aluminum. For producing the fuel, the processes for uranium hexafluoride (UF{sub 6}) conversion consist in obtaining U{sub 3}Si{sub 2} and / or U{sub 3}O{sub 8} through the preparation of intermediate compounds, among them ammonium uranyl carbonate - AUC, ammonium diuranate - DUA and uranium tetrafluoride - UF{sub 4}. This work describes a procedure for preparing uranium tetrafluoride by a dry route using as raw material the filtrate generated when producing routinely ammonium uranyl carbonate. The filtrate consists primarily of a solution containing high concentrations of ammonium (NH{sub 4}{sup +}), fluoride (F{sup -}), carbonate (CO{sub 3}{sup --}) and low concentrations of uranium. The procedure is basically the recovery of NH{sub 4}F and uranium, as UF{sub 4}, through the crystallization of ammonium bifluoride (NH{sub 4}HF{sub 2}) and, in a later step, the addition of UO{sub 2}, occurring fluoridation and decomposition. The UF{sub 4} obtained is further diluted in the UF{sub 4} produced routinely at IPEN / CNEN-SP by a wet route process. (author)

Development of metal uranium fuel and testing of construction materials (I-VI); Part I

International Nuclear Information System (INIS)

Mihajlovic, A.

1965-11-01

This project includes the following tasks: Study of crystallisation of metal melt and beta-alpha transforms in uranium and uranium alloys; Study of the thermal treatment influence on phase transformations and texture in uranium alloys; Radiation damage of metal uranium; Project related to irradiation of metal uranium in the reactor; Development of fuel element for nuclear reactors

Effect of spermidine in PC12 cells on the cell apoptosis induced by enriched uranium

International Nuclear Information System (INIS)

Gu Guixiong; Zhu Shoupeng; Wang Liuyi; Yang Shuqin; Zhu Lingli

2003-01-01

This is a study on injurious effects of cellular spermidine to PC12 cells irradiated by enriched uranium. PC12 cells were cultured in DMEM/F12 medium with enriched uranium, and the exposure doses were calculated. The contents of free spermidine PC12 cells were examined with Dansyl-chloride reaction and thin-layer chromatography. Viability of the cells treated with enriched uranium reduced rapidly and DNA strand break increased significantly with increasing time of the irradiation. Autoradiographic tracks showed that the radionuclide located in the nucleus predominantly. The content of free spermidine in PC12 cells could markedly decrease as the irradiation time increased. The results suggested that PC12 cells exposured to enriched uranium were apoptotic and the free spermidine in cells might play some role in this process

Summary report on fuel development and miniplate fabrication for the RERTR Program, 1978 to 1990

Energy Technology Data Exchange (ETDEWEB)

Wiencek, T.C. [Argonne National Lab., IL (United States). Energy Technology Div.

1995-08-01

This report summarizes the efforts of the Fabrication Technology Section at Argonne National Laboratory in the program of Reduced Enrichment Research and Test Reactors (RERTR). The main objective of this program was to reduce the amount of high enriched ({approx}93% {sup 235}U) uranium (HEU) used in nonpower reactors. Conversion from low-density (0.8--1.6 g U/cm{sup 3}) HEU fuel elements to highly loaded (up to 7 g U/cm{sup 3}) low-enrichment (<20% {sup 235}U) uranium (LEU) fuel elements allows the same reactor power levels, core designs and sizes to be retained while greatly reducing the possibility of illicit diversion of HEU nuclear fuel. This document is intended as an overview of the period 1978--1990, during which the Section supported this project by fabricating mainly powder metallurgy uranium-silicide dispersion fuel plates. Most of the subjects covered in detail are fabrication-related studies of uranium silicide fuels and fuel plate properties. Some data are included for out-of-pile experiments such as corrosion and compatibility tests. Also briefly covered are most other aspects of the RERTR program such as irradiation tests, full-core demonstrations, and technology transfer. References included are for further information on most aspects of the entire program. A significant portion of the report is devoted to data that were never published in their entirety. The appendices contain a list of previous RERTR reports, ANL fabrication procedures, calculations for phases present in two-phase fuels, chemical analysis of fuels, miniplate characteristics, and a summary of bonding runs made by hot isostatic pressing.

Summary report on fuel development and miniplate fabrication for the RERTR Program, 1978 to 1990

International Nuclear Information System (INIS)

Wiencek, T.C.

1995-08-01

This report summarizes the efforts of the Fabrication Technology Section at Argonne National Laboratory in the program of Reduced Enrichment Research and Test Reactors (RERTR). The main objective of this program was to reduce the amount of high enriched (∼93% 235 U) uranium (HEU) used in nonpower reactors. Conversion from low-density (0.8--1.6 g U/cm 3 ) HEU fuel elements to highly loaded (up to 7 g U/cm 3 ) low-enrichment ( 235 U) uranium (LEU) fuel elements allows the same reactor power levels, core designs and sizes to be retained while greatly reducing the possibility of illicit diversion of HEU nuclear fuel. This document is intended as an overview of the period 1978--1990, during which the Section supported this project by fabricating mainly powder metallurgy uranium-silicide dispersion fuel plates. Most of the subjects covered in detail are fabrication-related studies of uranium silicide fuels and fuel plate properties. Some data are included for out-of-pile experiments such as corrosion and compatibility tests. Also briefly covered are most other aspects of the RERTR program such as irradiation tests, full-core demonstrations, and technology transfer. References included are for further information on most aspects of the entire program. A significant portion of the report is devoted to data that were never published in their entirety. The appendices contain a list of previous RERTR reports, ANL fabrication procedures, calculations for phases present in two-phase fuels, chemical analysis of fuels, miniplate characteristics, and a summary of bonding runs made by hot isostatic pressing

Microstructural characterization of atom clusters in irradiated pressure vessel steels and model alloys

International Nuclear Information System (INIS)

Auger, P.; Pareige, P.; Akamatsu, M.; Van Duysen, J.C.

1993-01-01

In order to characterize the microstructural evolution of iron solid solution under irradiation, two pressure vessel steels irradiated in service conditions, and, for comparison, low copper model alloys irradiated with neutrons and electrons, have been studied through small angle neutron scattering and atom probe experiments. In Fe-Cu model alloys, copper clusters are formed containing uncertain proportions of iron. In the low copper industrial steels, the feature is more complex; solute atoms such as Ni, Mn and Si, sometimes associated with Cu, segregate as ''clouds'' more or less condensed in the iron solid solution. These silicides, or at least Si, Ni, Mn association, may facilitate the copper segregation although the initial iron matrix contains a low copper concentration. (authors). 24 refs., 3 figs., 2 tabs

Microstructural characterization of atom clusters in irradiated pressure vessel steels and model alloys

Energy Technology Data Exchange (ETDEWEB)

Auger, P; Pareige, P [Rouen Univ., 76 - Mont-Saint-Aignan (France); Akamatsu, M; Van Duysen, J C [Electricite de France (EDF), 77 - Ecuelles (France)

1994-12-31

In order to characterize the microstructural evolution of iron solid solution under irradiation, two pressure vessel steels irradiated in service conditions, and, for comparison, low copper model alloys irradiated with neutrons and electrons, have been studied through small angle neutron scattering and atom probe experiments. In Fe-Cu model alloys, copper clusters are formed containing uncertain proportions of iron. In the low copper industrial steels, the feature is more complex; solute atoms such as Ni, Mn and Si, sometimes associated with Cu, segregate as ``clouds`` more or less condensed in the iron solid solution. These silicides, or at least Si, Ni, Mn association, may facilitate the copper segregation although the initial iron matrix contains a low copper concentration. (authors). 24 refs., 3 figs., 2 tabs.

Bioassay method for Uranium in urine by Delay Neutron counting; Metoda Bioassay Uranium dalam urin dengan pencacahan Netron Kasip

Energy Technology Data Exchange (ETDEWEB)

Suratman,; Purwanto,; Sukarman-Aminjoyo, [Yogyakarta Nuclear Research Centre, National Atomic Energy Agency, Yogyakarta (Indonesia)

1996-04-15

A bioassay method for uranium in urine by neutron counting has been studied. The aim of this research is to obtain a bioassay method for uranium in urine which is used for the determination of internal dose of radiation workers. The bioassay was applied to the artificially uranium contaminated urine. The weight of the contaminant was varied. The uranium in the urine was irradiated in the Kartini reactor core, through pneumatic system. The delayed neutron was counted by BF3 neutron counter. Recovery of the bioassay was between 69.8-88.8 %, standard deviation was less than 10 % and the minimum detection was 0.387 {mu}g.

75 FR 6223 - Application For a License To Export High-Enriched Uranium

Science.gov (United States)

2010-02-08

... NUCLEAR REGULATORY COMMISSION Application For a License To Export High-Enriched Uranium Pursuant to 10 CFR 110.70(c) ``Public notice of receipt of an application,'' please take notice that the..., Uranium (93.35%). uranium (16.3 targets for December 28, 2009, XSNM3623, kilograms U-235). irradiation in...

Assessment of current atomic scale modelling methods for the investigation of nuclear fuels under irradiation: Example of uranium dioxide

International Nuclear Information System (INIS)

Bertolus, M.; Freyss, M.; Krack, M.; Devanathan, R.

2015-01-01

We focus here on the assessment of the description of interatomic interactions in uranium dioxide using, on the one hand, electronic structure methods, in particular in the Density Functional Theory (DFT) framework, and on the other hand, empirical potential methods. These two types of methods are complementary, the former enabling results to be obtained from a minimal amount of input data and further insight into the electronic and magnetic properties to be achieved, while the latter are irreplaceable for studies where a large number of atoms need to be considered. We consider basic properties as well as specific ones, which are important for the description of nuclear fuel under irradiation. These are especially energies, which are the main data passed on to higher scale models. For this exercise, we limit ourselves to uranium dioxide (UO 2 ) because of the extensive amount of studies available on this system. (authors)

The formation of magnetic silicide Fe{sub 3}Si clusters during ion implantation

Energy Technology Data Exchange (ETDEWEB)

Balakirev, N. [Kazan National Research Technological University, K.Marx st. 68, Kazan 420015 (Russian Federation); Zhikharev, V., E-mail: valzhik@mail.ru [Kazan National Research Technological University, K.Marx st. 68, Kazan 420015 (Russian Federation); Gumarov, G. [Zavoiskii Physico-Technical Institute of Russian Academy of Sciences, 10/7 Sibirskii trakt st., Kazan 420029 (Russian Federation)

2014-05-01

A simple two-dimensional model of the formation of magnetic silicide Fe{sub 3}Si clusters during high-dose Fe ion implantation into silicon has been proposed and the cluster growth process has been computer simulated. The model takes into account the interaction between the cluster magnetization and magnetic moments of Fe atoms random walking in the implanted layer. If the clusters are formed in the presence of the external magnetic field parallel to the implanted layer, the model predicts the elongation of the growing cluster in the field direction. It has been proposed that the cluster elongation results in the uniaxial magnetic anisotropy in the plane of the implanted layer, which is observed in iron silicide films ion-beam synthesized in the external magnetic field.

Production and analysis of ultradispersed uranium oxide powders

Science.gov (United States)

Zajogin, A. P.; Komyak, A. I.; Umreiko, D. S.; Umreiko, S. D.

2010-05-01

Spectroscopic studies are made of the laser plasma formed near the surface of a porous body containing nanoquantities of uranium compounds which is irradiated by two successive laser pulses. The feasibility of using laser chemical methods for obtaining nanoclusters of uranium oxide particles in the volume of a porous body and the simultaneous possibility of determining the uranium content with good sensitivity are demonstrated. The thermochemical and spectral characteristics of the analogs of their compounds with chlorine are determined and studied. The possibility of producing uranium dioxides under ordinary conditions and their analysis in the reaction products is demonstrated.

Secondary neutral mass spectrometry depth profile analysis of silicides

International Nuclear Information System (INIS)

Beckmann, P.; Kopnarski, M.; Oechsner, H.

1985-01-01

The Direct Bombardment Mode (DBM) of Secondary Neutral Mass Spectrometry (SNMS) has been applied for depth profile analysis of two different multilayer systems containing metal silicides. Due to the extremely high depth resolution obtained with low energy SNMS structural details down to only a few atomic distances are detected. Stoichiometric information on internal oxides and implanted material is supplied by the high quantificability of SNMS. (Author)

Aluminium alloyed iron-silicide/silicon solar cells: A simple approach for low cost environmental-friendly photovoltaic technology.

Science.gov (United States)

Kumar Dalapati, Goutam; Masudy-Panah, Saeid; Kumar, Avishek; Cheh Tan, Cheng; Ru Tan, Hui; Chi, Dongzhi

2015-12-03

This work demonstrates the fabrication of silicide/silicon based solar cell towards the development of low cost and environmental friendly photovoltaic technology. A heterostructure solar cells using metallic alpha phase (α-phase) aluminum alloyed iron silicide (FeSi(Al)) on n-type silicon is fabricated with an efficiency of 0.8%. The fabricated device has an open circuit voltage and fill-factor of 240 mV and 60%, respectively. Performance of the device was improved by about 7 fold to 5.1% through the interface engineering. The α-phase FeSi(Al)/silicon solar cell devices have promising photovoltaic characteristic with an open circuit voltage, short-circuit current and a fill factor (FF) of 425 mV, 18.5 mA/cm(2), and 64%, respectively. The significant improvement of α-phase FeSi(Al)/n-Si solar cells is due to the formation p(+-)n homojunction through the formation of re-grown crystalline silicon layer (~5-10 nm) at the silicide/silicon interface. Thickness of the regrown silicon layer is crucial for the silicide/silicon based photovoltaic devices. Performance of the α-FeSi(Al)/n-Si solar cells significantly depends on the thickness of α-FeSi(Al) layer and process temperature during the device fabrication. This study will open up new opportunities for the Si based photovoltaic technology using a simple, sustainable, and los cost method.

The manufacturing of depleted uranium biological shield components

International Nuclear Information System (INIS)

Metelkin, J.A.

1998-01-01

The unique combination of the physical and mechanical properties of uranium made it possible to manufacture biological shield components of transport package container (TPC) for transportation nuclear power plant irradiated fuel and radionuclides of radiation diagnostic instruments. Protective properties are substantially dependent on the nature radionuclide composition of uranium, that why I recommended depleted uranium after radiation chemical processing. Depleted uranium biological shield (DUBS) has improved specific mass-size characteristics compared to a shield made of lead, steel or tungsten. Technological achievements in uranium casting and machining made it possible to manufacture DUBS components of TPC up to 3 tons of mass and up to 2 metres of the maximum size. (authors)

Influence of Al addition on phase transformation and thermal stability of nickel silicides on Si(0 0 1)

International Nuclear Information System (INIS)

Huang, Shih-Hsien; Twan, Sheng-Chen; Cheng, Shao-Liang; Lee, Tu; Hu, Jung-Chih; Chen, Lien-Tai; Lee, Sheng-Wei

2014-01-01

Highlights: ► The presence of Al slows down the Ni 2 Si–NiSi phase transformation but significantly promotes the NiSi 2−x Al x formation. ► The behavior of phase transformation strongly depends on the Al concentration of the initial Ni 1−x Al x alloys. ► The Ni 0.91 Al 0.09 /Si system exhibits remarkably improved thermal stability, even after high temperature annealing for 1000 s. ► The relationship between microstructures, electrical property, and thermal stability of Ni(Al) silicides is discussed. -- Abstract: The influence of Al addition on the phase transformation and thermal stability of Ni silicides on (0 0 1)Si has been systematically investigated. The presence of Al atoms is found to slow down the Ni 2 Si–NiSi phase transformation but significantly promote the NiSi 2−x Al x formation during annealing. The behavior of phase transformation strongly depends on the Al concentration of the initial Ni 1−x Al x alloys. Compared to the Ni 0.95 Pt 0.05 /Si and Ni 0.95 Al 0.05 /Si system, the Ni 0.91 Al 0.09 /Si sample exhibits remarkably enhanced thermal stability, even after high temperature annealing for 1000 s. The relationship between microstructures, electrical property, and thermal stability of Ni silicides is discussed to elucidate the role of Al during the Ni 1−x Al x alloy silicidation. This work demonstrated that thermally stable Ni 1−x Al x alloy silicides would be a promising candidate as source/drain (S/D) contacts in advanced complementary metal–oxide-semiconductor (CMOS) devices

The conversion of NRU from HEU to LEU fuel

International Nuclear Information System (INIS)

Sears, D.F.; Atfield, M.D.; Kennedy, I.C.

1990-01-01

The program at Chalk River Nuclear Laboratories (CRNL) to develop and test low-enriched uranium fuel (LEU, 3 Si, USiAl, USi Al and U 3 Si 2 (U-3.96 wt% Si; U-3.5 wt% Si-1.5 wt% AL; U-3.2 wt%; Si-3 wt% Al; U-7.3 wt% Si, respectively). Fuel elements were fabricated with uranium loadings suitable for NRU, 3.15 gU/cm 3 , and for NRX, 4.5 gU/cm 3 , and were irradiated under normal fuel-operating conditions. Eight experimental irradiations involving 100 mini-elements and 84 full-length elements (7X12-element rods) were completed to qualify the LEU fuel and the fabrication technology. Post irradiation examinations confirmed that the performance of the LEU fuel, and that of a medium enrichment uranium (MEU, 45% U-235) alloy fuel tested as a back-up, was comparable to the HEU fuel. The uranium silicide dispersion fuel swelling was approximately linear up to burnups exceeding NRU's design terminal burnup (80 at%). NRU was partially converted to LEU fuel when the first 31 prototype fuel rods manufactured with industrial scale production equipment were installed in the reactor. The rods were loaded in NRU at a fuelling rate of about two rods per week over the period 1988 September to December. This partial LEU core (one third of a full NRU core) has allowed the reactor engineers and physicists to evaluate the bulk effects of the LEU conversion on NRU operations. As expected, the irradiation is proceeding without incident

Characterization of atom clusters in irradiated pressure vessel steels and model alloys

International Nuclear Information System (INIS)

Auger, P.; Pareige, P.; Akamatsu, M.; Van Duysen, J.C.

1993-12-01

In order to characterize the microstructural evolution of the iron solid solution under irradiation, two pressure vessel steels irradiated in service conditions and, for comparison, low copper model alloys irradiated with neutrons and electrons have been studied. The characterization has been carried out mainly thanks to small angle neutron scattering and atom probe experiments. Both techniques lead to the conclusion that clusters develop with irradiations. In Fe-Cu model alloys, copper clusters are formed containing uncertain proportions of iron. In the low copper industrial steels, the feature is more complex. Solute atoms like Ni, Mn and Si, sometimes associated with Cu, segregate as ''clouds'' more or less condensed in the iron solid solution. These silicides, or at least Si, Ni, Mn association, may facilitate the copper segregation although the initial iron matrix contains a low copper concentration. (authors). 24 refs., 3 figs., 2 tabs

X-ray photoemission spectromicroscopy of titanium silicide formation in patterned microstructures

Energy Technology Data Exchange (ETDEWEB)

Singh, S.; Solak, H.; Cerrina, F. [Univ. of Wisconsin-Madison, Stoughton, WI (United States)] [and others

1997-04-01

Titanium silicide has the lowest resistivity of all the refractory metal silicides and has good thermal stability as well as excellent compatibility with Al metallization. It is used as an intermediate buffer layer between W vias and the Si substrate to provide good electrical contact in ULSI technology, whose submicron patterned features form the basis of the integrated circuits of today and tomorrow, in the self aligned silicide (salicide) formation process. TiSi{sub 2} exists in two phases: a metastable C49 base-centered orthorhombic phase with specific resistivity of 60-90 {mu}{Omega}-cm that is formed at a lower temperature (formation anneal) and the stable 12-15 {mu}{Omega}-cm resistivity face-centered orthorhombic C54 phase into which C49 is transformed with a higher temperature (conversion anneal) step. C54 is clearly the target for low resistivity VLSI interconnects. However, it has been observed that when dimensions shrink below 1/mic (or when the Ti thickness drops below several hundred angstroms), the transformation of C49 into C54 is inhibited and agglomeration often occurs in fine lines at high temperatures. This results in a rise in resistivity due to incomplete transformation to C54 and because of discontinuities in the interconnect line resulting from agglomeration. Spectromicroscopy is an appropriate tool to study the evolution of the TiSi2 formation process because of its high resolution chemical imaging ability which can detect bonding changes even in the absence of changes in the relative amounts of species and because of the capability of studying thick {open_quotes}as is{close_quotes} industrial samples.

X-ray photoemission spectromicroscopy of titanium silicide formation in patterned microstructures

International Nuclear Information System (INIS)

Singh, S.; Solak, H.; Cerrina, F.

1997-01-01

Titanium silicide has the lowest resistivity of all the refractory metal silicides and has good thermal stability as well as excellent compatibility with Al metallization. It is used as an intermediate buffer layer between W vias and the Si substrate to provide good electrical contact in ULSI technology, whose submicron patterned features form the basis of the integrated circuits of today and tomorrow, in the self aligned silicide (salicide) formation process. TiSi 2 exists in two phases: a metastable C49 base-centered orthorhombic phase with specific resistivity of 60-90 μΩ-cm that is formed at a lower temperature (formation anneal) and the stable 12-15 μΩ-cm resistivity face-centered orthorhombic C54 phase into which C49 is transformed with a higher temperature (conversion anneal) step. C54 is clearly the target for low resistivity VLSI interconnects. However, it has been observed that when dimensions shrink below 1/mic (or when the Ti thickness drops below several hundred angstroms), the transformation of C49 into C54 is inhibited and agglomeration often occurs in fine lines at high temperatures. This results in a rise in resistivity due to incomplete transformation to C54 and because of discontinuities in the interconnect line resulting from agglomeration. Spectromicroscopy is an appropriate tool to study the evolution of the TiSi2 formation process because of its high resolution chemical imaging ability which can detect bonding changes even in the absence of changes in the relative amounts of species and because of the capability of studying thick open-quotes as isclose quotes industrial samples

Irradiation-induced precipitation in a SUS316 stainless steel using three-dimensional atom probe

International Nuclear Information System (INIS)

Hatakeyama, M.; Yamagata, I.

2013-01-01

Precipitation and segregation were investigated in a compositionally modified 316 austenitic stainless steel, neutron-irradiated at 862 K using a three-dimensional atom probe. In the solution-annealed specimen, Mo, Ti, Nb, C and P enrichment were observed in a silicide, with nominal composition Fe 3 Cr 2 Ni 2 Mo 2 Si 2 . In a Ti-rich carbide, nominaling Fe 5 Cr 8 Ni 10 Mo 2 Ti 11 Si 2 C 6 , enrichment of Mo, Si, O, and Nb was observed. Radiation-induced segregation (RIS) at the precipitate–matrix interface was also investigated at an atomic scale. RIS of Ni and P atoms, which are undersized in Fe, was also analyzed around the interface of the Ti-rich carbide and matrix. Results suggest that the carbide–matrix interface is a sink with an interstitial bias. In the cold-worked specimen, complex-precipitates consisting of silicide and carbide were formed

Neutron activation analysis for uranium and associated elements

International Nuclear Information System (INIS)

Bowman, W.W.

1977-01-01

The samples obtained by the Savannah River Laboratory as part of the National Uranium Resource Evaluation program are activated in the intense neutron flux from a Savannah River Plant production reactor. A pilot-scale facility was installed at the reactor site to provide analyses of samples through the initial phase of the program and to develop design data for a full-scale facility. Sediments are analyzed by direct activation of 0.5-g samples. However, to analyze ground or surface water samples, mineral elements from 1-liter samples are concentrated on ion exchange resin and then approximately 5-g samples of resin are activated. Uranium concentration is determined by counting neutrons emitted from specific short-lived products of fission induced in 235 U by the primary neutron flux. Repetitive short cycles of irradiation and counting permit detection and determination of <0.1 μg of uranium. Elements associated with uranium are determined by spectral analysis of the gamma ray activities induced by the cyclic and subsequent longer irradiations. The pilot facility consists of four irradiation positions (plus 2 spare positions), a sample loader and unloader, and counting stations with neutron and gamma ray detectors, all interconnected with a pneumatic sample transport system. A computer controls both the transport system and the data acquisition devices. Gamma ray counting data are stored on magnetic tape for further processing by a large central computer. Facility hardware and software are described. Repetitive analyses of standards have shown an accuracy within +-10% for uranium values and within +-25% for associated elements. A quality assurance program has been developed to maintain these levels of reliability

Interface reactions between Pd thin films and SiC by thermal annealing and SHI irradiation

Energy Technology Data Exchange (ETDEWEB)

Njoroge, E.G., E-mail: eric.njoroge@up.ac.za [Department of Physics, University of Pretoria, Pretoria (South Africa); Theron, C.C. [Department of Physics, University of Pretoria, Pretoria (South Africa); Skuratov, V.A. [Joint Institute for Nuclear Research, Dubna (Russian Federation); Wamwangi, D. [School of Physics, University of Witwatersrand, Johannesburg (South Africa); Hlatshwayo, T.T. [Department of Physics, University of Pretoria, Pretoria (South Africa); Comrie, C.M. [MRD, iThemba LABS, P.O. Box 722, Somerset West 7129 (South Africa); Malherbe, J.B. [Department of Physics, University of Pretoria, Pretoria (South Africa)

2016-03-15

The solid-state reactions between Pd thin films and 6H-SiC substrates induced by thermal annealing, room temperature swift heavy ion (SHI) irradiation and high temperature SHI irradiation have been investigated by in situ and real-time Rutherford backscattering spectrometry (RBS) and Grazing incidence X-ray diffraction (GIXRD). At room temperature, no silicides were detected to have formed in the Pd/SiC samples. Two reaction growth zones were observed in the samples annealed in situ and analysed by real time RBS. The initial reaction growth region led to formation of Pd{sub 3}Si or (Pd{sub 2}Si + Pd{sub 4}Si) as the initial phase(s) to form at a temperature of about 450 °C. Thereafter, the reaction zone did not change until a temperature of 640 °C was attained where Pd{sub 2}Si was observed to form in the reaction zone. Kinetic analysis of the initial reaction indicates very fast reaction rates of about 1.55 × 10{sup 15} at cm{sup −2}/s and the Pd silicide formed grew linear with time. SHI irradiation of the Pd/SiC samples was performed by 167 MeV Xe{sup 26+} ions at room temperature at high fluences of 1.07 × 10{sup 14} and 4 × 10{sup 14} ions/cm{sup 2} and at 400 °C at lower fluences of 5 × 10{sup 13} ions/cm{sup 2}. The Pd/SiC interface was analysed by RBS and no SHI induced diffusion was observed for room temperature irradiations. The sample irradiated at 400 °C, SHI induced diffusion was observed to occur accompanied with the formation of Pd{sub 4}Si, Pd{sub 9}Si{sub 2} and Pd{sub 5}Si phases which were identified by GIXRD analysis.

Laser separation of uranium isotopes

International Nuclear Information System (INIS)

Porter, J.T.

1981-01-01

Method and apparatus for separating uranium isotopes are claimed. The method comprises the steps of irradiating a uranyl source material at a wavelength selective to a desired isotope and at an effective temperature for isotope spectral line splitting below about 77 deg.K., further irradiating the source material within the fluorescent lifetime of the source material to selectively photochemically reduce the excited isotopic species, and chemically separating the reduced isotope species from the remaining uranyl salt compound

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

International Nuclear Information System (INIS)

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

1988-11-01

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

Some physical properties of irradiated and non-irradiated oxide glasses containing uranium

International Nuclear Information System (INIS)

Simon, V.; Ardelean, I.; Simon, S.; Cozar, O.; Milea, I.; Lupsa, I.; Mih, V.

1995-01-01

The x U O 3 (1-x) [2 P 2 O 5 · Na 2 O] non-irradiated and gamma irradiated glasses (0 3+ , U 4+ and U 5+ ions. The gamma irradiation induces paramagnetic defects around the glass network forming sites occupied by phosphorous atoms. The non-irradiated samples are weak paramagnetic up to x = 0.1. For higher U O 3 concentration (0.1 < x ≤ 0.2) the magnetic measurements indicated a larger number of paramagnetic ions which are magnetically isolated and exhibit a Curie type behaviour. (author) 5 figs., 14 refs

Qualification of high-density fuel manufacturing for research reactors at CNEA

Energy Technology Data Exchange (ETDEWEB)

Adelfang, P.; Alvarez, L.; Boero, N.; Calabrese, R.; De La Fuente, M.; Echenique, P.; Markiewicz, M.; Pasqualini, E.; Ruggirello, G.; Taboada, H. [CNEA, Buenos Aires (Argentina)

2001-07-01

CNEA, the National Atomic Energy Commission of Argentina, is at the present a qualified supplier of uranium oxide fuel for research reactors. A new objective in this field is to develop and qualify the manufacturing of LEU high-density fuel for this type of reactors. According with the international trend Silicide fuel and U-xMo fuel are included in our program as the most suitable options. The facilities to complete the qualification of high-density MTR fuels, like the manufacturing plant installations, the reactor, the pool side fuel examination station and the hot cells are fully operational and equipped to perform all the activities required within the program. The programs for both type of fuels include similar activities: development and set up of the fuel material manufacturing technology, set up of fuel plate manufacturing, fabrication and irradiation of miniplates, fabrication and irradiation of full scale fuel elements, post-irradiation examination and feedback for manufacturing improvements. For silicide fuels most of these steps have already been completed. For U-xMo fuel the activities also include the development of alternative ways to obtain U-xMo powder, feasibility studies for large-scale manufacturing and the economical assessment. Set up of U-xMo fuel plate manufacturing is also well advanced and the fabrication of the first full scale prototype is foreseen during this year. (author)

Rare earth silicide nanowires on silicon surfaces

International Nuclear Information System (INIS)

Wanke, Martina

2008-01-01

The growth, structure and electronic properties of rare earth silicide nanowires are investigated on planar and vicinal Si(001) und Si(111) surfaces with scanning tunneling microscopy (STM), low energy electron diffraction (LEED) and angle-resolved photoelectron spectroscopy (ARPES). On all surfaces investigated within this work hexagonal disilicides are grown epitaxially with a lattice mismatch of -2.55% up to +0.83% along the hexagonal a-axis. Along the hexagonal c-axis the lattice mismatch is essentially larger with 6.5%. On the Si(001)2 x 1 surface two types of nanowires are grown epitaxially. The socalled broad wires show a one-dimensional metallic valence band structure with states crossing the Fermi level. Along the nanowires two strongly dispersing states at the anti J point and a strongly dispersing state at the anti Γ point can be observed. Along the thin nanowires dispersing states could not be observed. Merely in the direction perpendicular to the wires an intensity variation could be observed, which corresponds to the observed spacial structure of the thin nanowires. The electronic properties of the broad erbium silicide nanowires are very similar to the broad dysprosium silicide nanowires. The electronic properties of the DySi 2 -monolayer and the Dy 3 Si 5 -multilayer on the Si(111) surface are investigated in comparison to the known ErSi 2 /Si(111) and Er 3 Si 5 /Si(111) system. The positions and the energetic locations of the observed band in the surface Brillouin zone will be confirmed for dysprosium. The shape of the electron pockets in the vector k parallel space is elliptical at the anti M points, while the hole pocket at the anti Γ point is showing a hexagonal symmetry. On the Si(557) surface the structural and electronic properties depend strongly on the different preparation conditions likewise, in particular on the rare earth coverage. At submonolayer coverage the thin nanowires grow in wide areas of the sample surface, which are oriented

Rate Theory Modeling and Simulation of Silicide Fuel at LWR Conditions

Energy Technology Data Exchange (ETDEWEB)

Miao, Yinbin [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Ye, Bei [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Hofman, Gerard [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Yacout, Abdellatif [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Gamble, Kyle [Idaho National Lab. (INL), Idaho Falls, ID (United States). Fuel Modeling and Simulation; Mei, Zhi-Gang [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division

2016-08-29

As a promising candidate for the accident tolerant fuel (ATF) used in light water reactors (LWRs), the fuel performance of uranium silicide (U₃Si₂) at LWR conditions needs to be well understood. In this report, rate theory model was developed based on existing experimental data and density functional theory (DFT) calculations so as to predict the fission gas behavior in U₃Si₂ at LWR conditions. The fission gas behavior of U₃Si₂ can be divided into three temperature regimes. During steady-state operation, the majority of the fission gas stays in intragranular bubbles, whereas the dominance of intergranular bubbles and fission gas release only occurs beyond 1000 K. The steady-state rate theory model was also used as reference to establish a gaseous swelling correlation of U₃Si₂ for the BISON code. Meanwhile, the overpressurized bubble model was also developed so that the fission gas behavior at LOCA can be simulated. LOCA simulation showed that intragranular bubbles are still dominant after a 70 second LOCA, resulting in a controllable gaseous swelling. The fission gas behavior of U₃Si₂ at LWR conditions is benign according to the rate theory prediction at both steady-state and LOCA conditions, which provides important references to the qualification of U₃Si₂ as a LWR fuel material with excellent fuel performance and enhanced accident tolerance.

Oxidation-resistant Ge-doped silicide coating on Cr-Cr2Nb alloys by pack cementation

International Nuclear Information System (INIS)

He Yirong

1997-01-01

The halide-activated pack cementation process was modified to produce a Ge-doped silicide diffusion coating on Cr-Cr 2 Nb alloys in a single processing step. The morphology and composition of the coating depended both on the pack composition and processing schedule and also on the composition and microstructure of the substrate. Higher Ge content in the pack suppressed the formation of CrSi 2 and reduced the growth kinetics of the coating. Ge was not homogeneously distributed in the coatings. Under cyclic and isothermal oxidation conditions, the Ge-doped silicide coating protected the Cr-Nb alloys from significant oxidation and from pesting by the formation of a Ge-doped silica film. (orig.)

Design of high density gamma-phase uranium alloys for LEU dispersion fuel applications

International Nuclear Information System (INIS)

Hofman, Gerard L.; Meyer, Mitchell K.; Ray, Allison E.

1998-01-01

Uranium alloys are candidates for the fuel phase in aluminium matrix dispersion fuels requiring high uranium loading. Certain uranium alloys have been shown to have good irradiation performance at intermediate burnup. previous studies have shown that acceptable fission gas swelling behavior and fuel-aluminium interaction is possible only if the fuel alloy can be maintained in the high temperature body-centered-cubic γ-phase during fabrication and irradiation, at temperatures at which αU is the equilibrium phase. transition metals in Groups V through VIII are known to allow metastable retention of the gamma phase below the equilibrium isotherm. These metals have varying degrees of effectiveness in stabilizing the gamma phase. Certain alloys are metastable for very long times at the relatively low fuel temperatures seen in research operation. In this paper, the existing data on the gamma stability of binary and ternary uranium alloys is analysed. The mechanism and kinetics of decomposition of the gamma phase are assessed with the help of metal alloy theory. Alloys with the highest possible uranium content, good gamma-phase stability, and good neutronic performance are identified for further metallurgical studies and irradiation tests. Results from theory will be compared with experimentally generated data. (author)

Irradiation-induced precipitation in a SUS316 stainless steel using three-dimensional atom probe

Energy Technology Data Exchange (ETDEWEB)

Hatakeyama, M., E-mail: hatake@imr.tohoku.ac.jp [International Research Center for Nuclear Materials Science, IMR/Tohoku University, Narita, Oarai, Ibaraki 311-1313 (Japan); Yamagata, I. [Japan Atom Energy Agency, Narita, Oarai, Ibaraki 311-1393 (Japan)

2013-11-15

Precipitation and segregation were investigated in a compositionally modified 316 austenitic stainless steel, neutron-irradiated at 862 K using a three-dimensional atom probe. In the solution-annealed specimen, Mo, Ti, Nb, C and P enrichment were observed in a silicide, with nominal composition Fe{sub 3}Cr{sub 2}Ni{sub 2}Mo{sub 2}Si{sub 2}. In a Ti-rich carbide, nominaling Fe{sub 5}Cr{sub 8}Ni{sub 10}Mo{sub 2}Ti{sub 11}Si{sub 2}C{sub 6}, enrichment of Mo, Si, O, and Nb was observed. Radiation-induced segregation (RIS) at the precipitate–matrix interface was also investigated at an atomic scale. RIS of Ni and P atoms, which are undersized in Fe, was also analyzed around the interface of the Ti-rich carbide and matrix. Results suggest that the carbide–matrix interface is a sink with an interstitial bias. In the cold-worked specimen, complex-precipitates consisting of silicide and carbide were formed.

Dislocations and radiation damage in {alpha}-uranium; Dislocations et effets des radiations dans l'uranium {alpha}

Energy Technology Data Exchange (ETDEWEB)

Leteurtre, J [Commissariat a l' Energie Atomique, 92 - Fontenay-Aux-Roses (France). Centre d' Etudes Nucleaires

1968-07-01

Dislocations in {alpha}-uranium were studied by electron microscopy. Electropolishing of thin foils was performed at low temperature (-110 deg. C) to prevent oxidation. Burgers vectors of twins dislocations are defined. Interactions between slip and twinning are studied from both experimental and theoretical point of view. Samples irradiated at several burn-up were examined. In order to explain our micrographic results, and also all information gathered in literature about radiation damage in {alpha}-uranium, a coherent model is propound for the fission particles effects. We analyse the influences of parameters: temperature, dislocation density, impurity content. The number of point defects created by one initial fission is determined for pure and annealed metal. The importance of the self-anneal which occurs immediately in each displacement spike, and the anneal due to a new fission on the damage resulting from a previous fission, are estimated. The focussing distance in [100] direction is found to be about 1000 Angstrom, at 4 deg. K. (author) [French] Ce travail est une etude par microscopie electronique des dislocations induites dans l'uranium {alpha}, soit par deformation plastique, soit par irradiation. Une methode de preparation des lames minces a basse temperature (-110 deg. C) a ete mise au point. Les vecteurs de Burgers des diverses dislocations de macles de ce metal ont ete definis. Les interactions glissements- maclages sont etudiees experimentalement et theoriquement. Des echantillons irradies a divers taux de combustion ont ete examines. Pour expliquer nos resultats micrographiques, et aussi l'ensemble des informations recueillies dans la litterature concernant l'endommagement par irradiation de l'uranium-{alpha}, nous proposons un modele coherent de l'effet des fragments de fission dans ce metal. L'influence des parametres: temperature, densite de dislocations, impuretes est analysee. Le nombre de defauts ponctuels crees par une fission dans du metal

Energetics and dynamics of atomic uranium levels

International Nuclear Information System (INIS)

Miron, E.

1978-01-01

New methods for discovering and identifying new electronic levels of atomic uranium and measuring parameters, such as radiative lifetimes and absorption cross-sections, are described. The uranium atoms are produced within an especially designed induction-heated oven. The uranium vapor is irradiated by nitrogen laser pumped, pulsed dye lasers. The various measurements are accomplished by detection of laser induced fluorescence from selectively excited levels. 138 atomic-uranium odd levels in the region 32260-34900 cm -1 and 16 even levels in the region 49500-49900 cm -1 are reported. Unique J values are presented for 64 levels and partial assignment (two possibilities) for 42 levels. Radiative lifetimes are presented for 134 levels. Absorption cross sections were measured for 12 transitions. Isotope shifts of 17 levels are given. Cross-sections for internal excitation transfer in uranium which are induced by collisions with argon atoms, are presented for 11 levels. (author)

X-ray-emission studies of chemical bonding in transition-metal silicides

NARCIS (Netherlands)

Weijs, P.J.W.; Leuken, H. van; Groot, R.A. de; Fuggle, J.C.; Reiter, S.; Wiech, G.; Buschow, K.H.J.

1991-01-01

We present Si L2,3 emission-band spectra of a series of 3d and 4d transition-metal (TM) silicides, together wtih Si K emission-band spectra of four 3d TM disilicides. The data are compared with augmented-spherical-wave density-of-states (DOS) calculations, and good agreement is found. The trends we

Micromechanical simulation of Uranium dioxide polycrystalline aggregate behaviour under irradiation

International Nuclear Information System (INIS)

Pacull, J.

2011-02-01

In pressurized water nuclear power reactor (PWR), the fuel rod is made of dioxide of uranium (UO 2 ) pellet stacked in a metallic cladding. A multi scale and multi-physic approaches are needed for the simulation of fuel behavior under irradiation. The main phenomena to take into account are thermomechanical behavior of the fuel rod and chemical-physic behavior of the fission products. These last years one of the scientific issue to improve the simulation is to take into account the multi-physic coupling problem at the microscopic scale. The objective of this ph-D study is to contribute to this multi-scale approach. The present work concerns the micro-mechanical behavior of a polycrystalline aggregate of UO 2 . Mean field and full field approaches are considered. For the former and the later a self consistent homogenization technique and a periodic Finite Element model base on the 3D Voronoi pattern are respectively used. Fuel visco-plasticity is introduced in the model at the scale of a single grain by taking into account specific dislocation slip systems of UO 2 . A cohesive zone model has also been developed and implemented to simulate grain boundary sliding and intergranular crack opening. The effective homogenous behaviour of a Representative Volume Element (RVE) is fitted with experimental data coming from mechanical tests on a single pellet. Local behavior is also analyzed in order to evaluate the model capacity to assess micro-mechanical state. In particular, intra and inter granular stress gradient are discussed. A first validation of the local behavior assessment is proposed through the simulation of intergranular crack opening measured in a compressive creep test of a single fuel pellet. Concerning the impact of the microstructure on the fuel behavior under irradiation, a RVE simulation with a representative transient loading of a fuel rod during a power ramp test is achieved. The impact of local stress and strain heterogeneities on the multi

Report of Post Irradiation Examination for Dry Process Fuel

International Nuclear Information System (INIS)

Par, Jang Jin; Jung, I. H.; Kang, K. H.; Moon, J. S.; Lee, C. R.; Ryu, H. J.; Song, K. C.; Yang, M. S.; Yoo, B. O.; Jung, Y. H.; Choo, Y. S.

2006-08-01

The spent PWR fuel typically contains 0.9 wt.% of fissile uranium and 0.6 wt.% of fissile plutonium, which exceeds the natural uranium fissile content of 0.711 wt.%. The neutron economy of a CANDU reactor is sufficient to utilize the DUPIC fuel, even though the neutron-absorbing fission products contained in the spent PWR fuel were remained in the DUPIC fuel. The DUPIC fuel cycle offers advantages to the countries operating both the PWR and CANDU reactors, such as saving the natural uranium, reducing the spent fuel in both PWR and CANDU, and acquiring the extra energy by reuse of the PWR spent fuel. This report contains the results of post-irradiation examination of the DUPIC fuel irradiated four times at HANARO from May 2000 to August 2006 present except the first irradiation test of simulated DUPIC fuel at HANARO on August 1999

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

International Nuclear Information System (INIS)

Muhammad, Farhan; Majid, Asad

2009-01-01

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

Ceramography of Irradiated tristructural isotropic (TRISO) Fuel from the AGR-2 Experiment

Energy Technology Data Exchange (ETDEWEB)

Rice, Francine Joyce [Idaho National Lab. (INL), Idaho Falls, ID (United States); Stempien, John Dennis [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-09-01

Ceramography was performed on cross sections from four tristructural isotropic (TRISO) coated particle fuel compacts taken from the AGR-2 experiment, which was irradiated between June 2010 and October 2013 in the Advanced Test Reactor (ATR). The fuel compacts examined in this study contained TRISO-coated particles with either uranium oxide (UO2) kernels or uranium oxide/uranium carbide (UCO) kernels that were irradiated to final burnup values between 9.0 and 11.1% FIMA. These examinations are intended to explore kernel and coating morphology evolution during irradiation. This includes kernel porosity, swelling, and migration, and irradiation-induced coating fracture and separation. Variations in behavior within a specific cross section, which could be related to temperature or burnup gradients within the fuel compact, are also explored. The criteria for categorizing post-irradiation particle morphologies developed for AGR-1 ceramographic exams, was applied to the particles in the AGR-2 compacts particles examined. Results are compared with similar investigations performed as part of the earlier AGR-1 irradiation experiment. This paper presents the results of the AGR-2 examinations and discusses the key implications for fuel irradiation performance.

A solvent proceed for the extraction of the irradiate uranium and plutonium in the reactor core; Un procede par solvant pour l'extraction du plutonium de l'uranium irradie dans les piles

Energy Technology Data Exchange (ETDEWEB)

Goldschmidt, B; Regnaut, P; Prevot, I [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1955-07-01

Description of the conditions of plutonium, fission products and of uranium separation by selective extraction of the nitrates by organic solvent, containing a simultaneous extraction of plutonium and uranium, followed by a plutonium re-extraction after reduction, and an uranium re-extraction. The rates of decontamination being insufficient in this first stage, we also describes the processes of decontamination permitting separately to get the rates wanted for uranium and plutonium. Finally, we describes the beginning of the operation that consists in a nitric dissolution of the active uranium while capturing the products of gaseous fission, as well as the final concentration of the products of fission in a concentrated solution. (authors) [French] Description des conditions de separation du plutonium, des produits de fission et de l'uranium au moyen d'une extraction selective des nitrates par solvant organique, comprenant une extraction simultanee du plutonium et de l'uranium, suivie d'une reextraction du plutonium apres reduction, et d'une reextraction de l'uranium. Les taux de decontamination etant insuffisants dans ce premier stade, on decrit egalement les processus de decontamination permettant separement d'obtenir les taux desires pour l'uranium et le plutonium. Enfin, on decrit aussi le debut de l'operation qui consiste en une dissolution nitrique de l'uranium actif en captant les produits de fission gazeux, ainsi que la concentration finale des produits de fission sous forme de solution concentree. (auteurs)

Possibility of using metal uranium fuel in heavy water reactors

International Nuclear Information System (INIS)

Djuric, B.; Mihajlovic, A.; Drobnjak, Dj.

1965-01-01

The review of metal uranium properties including irradiation in the reactor core lead to the following conclusions. Using metal uranium in the heavy water reactors would be favourable from economic point of view for ita high density, i.e. high conversion factor and low cost of fuel elements fabrication. Most important constraint is swelling during burnup and corrosion

Interaction between uranium oxide alloyed with Al2O3·SiO2 and pyrocarbon coating during irradiation of micro fuel elements

International Nuclear Information System (INIS)

Chernikov, A.S.; Khromov, Y.F.; Svistunov, D.E.; Chuiko, E.E.

1989-01-01

The thermodynamics of the interaction between uranium oxide and carbon was previously studied in the presence of Al 2 O 3 ·SiO 2 , SiC, and UC 1.86 ; in this case, the quantity of the reacting substances does not have any effect on the attainment of the equilibrium state. Based on the obtained results, it is interesting to study the characteristic features of the interaction between the alloyed UO x cores (kernels) with the PyC-coating under the conditions involving irradiation of the micro fuel elements with thermal neutrons and the formation of solid fission products. The data concerning the characteristics of a micro fuel element (the weight of the core, its composition, etc.) are useful for carrying out a quantitative evaluation of the additives required for fixing the alkali-earth fission products by obtaining stable compounds of aluminosilicates with Ba, Sr, Rb, and Cs at different levels of depletion (burnup) of the oxide fuel. An analysis of the interaction processes in such a complex system as the irradiated alloyed uranium oxide fuel located in a micro fuel element is carried out by comparing the chemical potential of oxygen (RT ln P O 2 ) for the competing constituents of the system

Micro-SHINE Uranyl Sulfate Irradiations at the Linac

Energy Technology Data Exchange (ETDEWEB)

Youker, Amanda J. [Argonne National Lab. (ANL), Argonne, IL (United States); Kalensky, Michael [Argonne National Lab. (ANL), Argonne, IL (United States); Chemerisov, Sergey [Argonne National Lab. (ANL), Argonne, IL (United States); Schneider, John [Argonne National Lab. (ANL), Argonne, IL (United States); Byrnes, James [Argonne National Lab. (ANL), Argonne, IL (United States); Vandegrift, George F. [Argonne National Lab. (ANL), Argonne, IL (United States)

2016-08-01

Peroxide formation due to water radiolysis in a uranyl sulfate solution is a concern for the SHINE Medical Technologies process in which Mo-99 is generated from the fission of dissolved low enriched uranium. To investigate the effects of power density and fission on peroxide formation and uranyl-peroxide precipitation, uranyl sulfate solutions were irradiated using a 50-MeV electron linac as part of the micro-SHINE experimental setup. Results are given for uranyl sulfate solutions with both high and low enriched uranium irradiated at different linac powers.

Dislocations and radiation damage in {alpha}-uranium; Dislocations et effets des radiations dans l'uranium {alpha}

Energy Technology Data Exchange (ETDEWEB)

Leteurtre, J. [Commissariat a l' Energie Atomique, 92 - Fontenay-Aux-Roses (France). Centre d' Etudes Nucleaires

1968-07-01

Dislocations in {alpha}-uranium were studied by electron microscopy. Electropolishing of thin foils was performed at low temperature (-110 deg. C) to prevent oxidation. Burgers vectors of twins dislocations are defined. Interactions between slip and twinning are studied from both experimental and theoretical point of view. Samples irradiated at several burn-up were examined. In order to explain our micrographic results, and also all information gathered in literature about radiation damage in {alpha}-uranium, a coherent model is propound for the fission particles effects. We analyse the influences of parameters: temperature, dislocation density, impurity content. The number of point defects created by one initial fission is determined for pure and annealed metal. The importance of the self-anneal which occurs immediately in each displacement spike, and the anneal due to a new fission on the damage resulting from a previous fission, are estimated. The focussing distance in [100] direction is found to be about 1000 Angstrom, at 4 deg. K. (author) [French] Ce travail est une etude par microscopie electronique des dislocations induites dans l'uranium {alpha}, soit par deformation plastique, soit par irradiation. Une methode de preparation des lames minces a basse temperature (-110 deg. C) a ete mise au point. Les vecteurs de Burgers des diverses dislocations de macles de ce metal ont ete definis. Les interactions glissements- maclages sont etudiees experimentalement et theoriquement. Des echantillons irradies a divers taux de combustion ont ete examines. Pour expliquer nos resultats micrographiques, et aussi l'ensemble des informations recueillies dans la litterature concernant l'endommagement par irradiation de l'uranium-{alpha}, nous proposons un modele coherent de l'effet des fragments de fission dans ce metal. L'influence des parametres: temperature, densite de dislocations, impuretes est analysee. Le nombre de defauts ponctuels crees

Progress in the development of uranium silicide (U3Si2) fuel at BATAN

International Nuclear Information System (INIS)

Suripto, A.; Soentono, S.

1995-01-01

After successful fabrication of two full-size prototype fuel elements containing ∼3.0 gU/cm 3 in the form of U 3 Si 2 -Al dispersion now undergoing irradiation in the Reaktor Serba Guna G.A. Siwabessy (RSG-GAS) core since 1990, further development in U 3 Si 2 -A2 dispersion fuel element manufacturing has been pursued, whose progress in discussed in this paper, with a special attention on the use of much higher-loading aimed at obtaining a better understanding on the influence of higher-loading on fuel core and plate manufacturing and quality. At present, high-loading U 3 Si 2 -AI dispersion miniplates are being manufactured for preparing some mini-fuel elements to be test-irradiated in the new MTR in-pile loop of the RSG-GAS. (author)

Preparation of High-Density Uranium-Silicide U3Sl2-Uss: Effects of Preirradiation Heat Treatment on As-Cast Ingot Fuel Plates

International Nuclear Information System (INIS)

Suripto, A; Yuwono

1998-01-01

Heat treatment experiments upon U 3 Si 2 - U ss ingot have been cam e d out to obtain free uranium particle size improvement which is required to enhance the U-Al inter-diffusion reaction in the fuel plate meat. . Heat treatment experiments upon fuel plates containing dispersion of U 3 Si 2 - U ss in Al matrix have also been carried out to study the effect of temperature and treatment duration on the extent of inter-diffusion reaction between free uranium particle and aluminium matrix in the fuel plate meat. Both the experiments indicate that a drastic size improvement has occurred with the U 3 Si 2 as well as free uranium particles upon heat treatment at controlled temperature between the U 3 Si 2 peritectic and peritectoid temperatures and that the inter-diffusion reaction between free uranium and Al matrix occurs quite significantly at temperatures higher than that ordinarily used in the fabrication procedure

Rare earth silicide nanowires on silicon surfaces

Energy Technology Data Exchange (ETDEWEB)

Wanke, Martina

2008-11-10

The growth, structure and electronic properties of rare earth silicide nanowires are investigated on planar and vicinal Si(001) und Si(111) surfaces with scanning tunneling microscopy (STM), low energy electron diffraction (LEED) and angle-resolved photoelectron spectroscopy (ARPES). On all surfaces investigated within this work hexagonal disilicides are grown epitaxially with a lattice mismatch of -2.55% up to +0.83% along the hexagonal a-axis. Along the hexagonal c-axis the lattice mismatch is essentially larger with 6.5%. On the Si(001)2 x 1 surface two types of nanowires are grown epitaxially. The socalled broad wires show a one-dimensional metallic valence band structure with states crossing the Fermi level. Along the nanowires two strongly dispersing states at the anti J point and a strongly dispersing state at the anti {gamma} point can be observed. Along the thin nanowires dispersing states could not be observed. Merely in the direction perpendicular to the wires an intensity variation could be observed, which corresponds to the observed spacial structure of the thin nanowires. The electronic properties of the broad erbium silicide nanowires are very similar to the broad dysprosium silicide nanowires. The electronic properties of the DySi{sub 2}-monolayer and the Dy{sub 3}Si{sub 5}-multilayer on the Si(111) surface are investigated in comparison to the known ErSi{sub 2}/Si(111) and Er{sub 3}Si{sub 5}/Si(111) system. The positions and the energetic locations of the observed band in the surface Brillouin zone will be confirmed for dysprosium. The shape of the electron pockets in the (vector)k {sub parallel} space is elliptical at the anti M points, while the hole pocket at the anti {gamma} point is showing a hexagonal symmetry. On the Si(557) surface the structural and electronic properties depend strongly on the different preparation conditions likewise, in particular on the rare earth coverage. At submonolayer coverage the thin nanowires grow in wide areas

The Comparison Study of Neutron Activation Analysis and Fission Track Technique for Uranium Determination

International Nuclear Information System (INIS)

Sirinuntavid, Alice; Rodthongkom, Chouvana

2007-08-01

Full text: Comparison between Neutron Activation Analysis (NAA) and fission track technique for uranium determination in solid samples was studied by use of standard reference materials, i.e., ore, coal fly ash, soil. For NAA, the epithermal neutron was applied for activated irradiation. Then, the 74.5 keV gamma from U-239 or 277.7 keV gamma from Np-239 was measured. For high Uranium content samples, NAA method with 74.5 keV gamma measurement, gave higher precision result than the 277.7 keV gamma measurement method. NAA method with 277.7 keV gamma measurement, gave higher sensitivity and precision result for low Uranium content samples and the uranium contained less than 10 ppm samples. Nevertheless, the latter procedure needed longer time for neutron irradiation and analysis procedure. In comparison the results of Uranium analysis between NAA and fission track, it was found that no significant difference within 95 % of confidence level

Calculation and measurement of the uranium temperature during irradiation in the experimental channel in the reflector of the RA reactor - Annex 15; Prilog 15 - Proracun i merenje temperature urana pri ozracivanju u eksperimentalnom kanalu reflektora na reaktoru RA

Energy Technology Data Exchange (ETDEWEB)

Nikolic, M; Strugar, P; Mitrovic, S [Institute of Nuclear Sciences Boris Kidric, Reaktor RA, Vinca, Beograd (Serbia and Montenegro)

1964-12-15

Upon demand of the Laboratory for fuel reprocessing, six domestic metal uranium pellets were exposed to neutron flux ( 4 - 5 10{sup 12} n cm{sup -2} sec {sup -1}) in the RA reactor. Irradiation of fuel demanded special analyses for safety reasons. Weight of the fuel pellets was 13 - 20 g, having diameter 20 mm. pellets were placed in leak tight aluminium capsules with helium. The irradiation was dome in the aluminium experimental channel in the graphite reflector. Theoretical study has shown that the expected fuel temperature in the core could be up to 300 deg C at nominal power. For that reason temperature of the capsule with the uranium sample was measured during irradiation by using thermocouples. Results showed the discrepancy between measure and calculated values to be about 30%.

The consequences and hazards of depleted uranium weapons used by US army since gulf war

International Nuclear Information System (INIS)

Mao Yongze

1999-01-01

Military equipment and development of depleted uranium weapon in USA, the depleted uranium weapon used in gulf war by USA army, personnel irradiation in the gulf war, and the protection in the gulf war are introduced. The radioactivity, radioactive characteristics, chemical toxicity and hazard of the depleted uranium are also introduced

Fission product phases in irradiated carbide fuels

International Nuclear Information System (INIS)

Ewart, F.T.; Sharpe, B.M.; Taylor, R.G.

1975-09-01

Oxide fuels have been widely adopted as 'first charge' fuels for demonstration fast reactors. However, because of the improved breeding characteristics, carbides are being investigated in a number of laboratories as possible advanced fuels. Irradiation experiments on uranium and mixed uranium-plutonium carbides have been widely reported but the instances where segregate phases have been found and subjected to electron probe analysis are relatively few. Several observations of such segregate phases have now been made over a period of time and these are collected together in this document. Some seven fuel pins have been examined. Two of the irradiations were in thermal materials testing reactors (MTR); the remainder were experimental assemblies of carbide gas bonded oxycarbide and sodium bonded oxycarbide in the Dounreay Fast Reactor (DFR). All fuel pins completed their irradiation without failure. (author)

Reactivity And Neutron Flux At Silicide Fuel Element In The Core Of RSG-GAS

International Nuclear Information System (INIS)

Hamzah, Amir

2000-01-01

In order to 4.8 and 5.2 gr U/cm exp 3 loading of U 3 Si 2 --Al fuel plates characterization, he core reactivity change and neutron flux depression had been done. Control rod calibration method was used to reactivity change measurement and neutron flux distribution was measured using foil activation method. Measurement of insertion of A-type of testing fuel element with U-loading above cannot be done due to technical reason, so the measurement using full type silicide fuel element of 2.96 gr U/cm exp 3 loading. The reactivity change measurement result of insertion in A-9 and C-3 is + 2.67 cent. The flux depression at silicide fuel in A-9 is 1.69 times bigger than oxide and in C-3 is 0.68 times lower than oxide

A long-term ultrahigh temperature application of layered silicide coated Nb alloy in air

Science.gov (United States)

Sun, Jia; Fu, Qian-Gang; Li, Tao; Wang, Chen; Huo, Cai-Xia; Zhou, Hong; Yang, Guan-Jun; Sun, Le

2018-05-01

Nb-based alloy possessed limited application service life at ultrahigh temperature (>1400 °C) in air even taking the effective protective coating strategy into consideration for last decades. In this work a long duration of above 128 h at 1500 °C in air was successfully achieved on Nb-based alloy thanked to multi-layered silicide coating. Through optimizing interfaces, the MoSi2/NbSi2 silicide coating with Al2O3-adsorbed-particles layer exhibited three-times higher of oxidation resistance capacity than the one without it. In MoSi2-Al2O3-NbSi2 multilayer coating, the Al2O3-adsorbed-particles layer playing as an element-diffusion barrier role, as well as the formed porous Nb5Si3 layer as a stress transition zone, contributed to the significant improvement.

Uranium in blood

International Nuclear Information System (INIS)

Koul, S.L.; Chadderton, L.T.

1979-01-01

The authors conduct a feasibility study of a determination of the concentration of uranium in human blood. Smples of blood from six individuals were taken, predivided into two groups of three. One group was of healthy males and used as a norm. The other group was of patients with certified leukemia. The samples were irradiated by thermal neutrons and fission fragments detected using Lexan polycarbonate discs. (G.T.H.)

Uranium logging in earth formations

International Nuclear Information System (INIS)

Givens, W.W.

1979-01-01

A technique is provided for assaying the formations surrounding a borehole for uranium. A borehole logging tool cyclically irradiates the formations with neutrons and responds to neutron fluxes produced during the period of time that prompt neutrons are being produced by the neutron fission of uranium in the formations. A borehole calibration tool employs a steady-state (continuous output) neutron source, firstly, to produce a response to neutron fluxes in models having known concentrations of uranium and, secondly, to to produce a response to neutron fluxes in the formations surrounding the borehole. The neutron flux responses of the borehole calibration tool in both the model and the formations surrounding the borehole are utilized to correct the neutron flux response of the borehole logging tool for the effects of epithermal/thermal neutron moderation, scattering, and absorption within the borehole itself

Development of very-high-density low-enriched-uranium fuels

International Nuclear Information System (INIS)

Snelgrove, J.L.; Hofman, G.L.; Meyer, M.K.; Trybus, C.L.; Wiencek, T.C.

1997-01-01

Following a hiatus of several years and following its successful development and qualification of 4.8 g U cm -3 U 3 Si 2 -Al dispersion fuel for application with low-enriched uranium in research and test reactors, the US Reduced Enrichment for Research and Test Reactors program has embarked on the development of even-higher-density fuels. Our goal is to achieve uranium densities of 8-9 g cm -3 in aluminum-based dispersion fuels. Achieving this goal will require the use of high-density, γ-stabilized uranium alloy powders in conjunction with the most-advanced fuel fabrication techniques. Key issues being addressed are the reaction of the fuel alloys with aluminum and the irradiation behavior of the fuel alloys and any reaction products. Test irradiations of candidate fuels in very-small (micro) plates are scheduled to begin in the Advanced Test Reactor during June, 1997. Initial results are expected to be available in early 1998. We are performing out-of-reactor studies on the phase structure of the candidate alloys on diffusion of the matrix material into the aluminum. In addition, we are modifying our current dispersion fuel irradiation behavior model to accommodate the new fuels. Several international partners are participating in various phases of this work. (orig.)

Electrical and optical properties of sub-10 nm nickel silicide films for silicon solar cells

International Nuclear Information System (INIS)

Brahmi, Hatem; Ravipati, Srikanth; Yarali, Milad; Wang, Weijie; Ryou, Jae-Hyun; Mavrokefalos, Anastassios; Shervin, Shahab

2017-01-01

Highly conductive and transparent films of ultra-thin p-type nickel silicide films have been prepared by RF magnetron sputtering of nickel on silicon substrates followed by rapid thermal annealing in an inert environment in the temperature range 400–600 °C. The films are uniform throughout the wafer with thicknesses in the range of 3–6 nm. The electrical and optical properties are presented for nickel silicide films with varying thickness. The Drude–Lorentz model and Fresnel equations were used to calculate the dielectric properties, sheet resistance, absorption and transmission of the films. These ultrathin nickel silicide films have excellent optoelectronic properties for p-type contacts with optical transparencies up to 80% and sheet resistance as low as ∼0.15 µΩ cm. Furthermore, it was shown that the use of a simple anti-reflection (AR) coating can recover most of the reflected light approaching the values of a standard Si solar cell with the same AR coating. Overall, the combination of ultra-low thickness, high transmittance, low sheet resistance and ability to recover the reflected light by utilizing standard AR coating makes them ideal for utilization in silicon based photovoltaic technologies as a p-type transparent conductor. (paper)

Electrical and optical properties of sub-10 nm nickel silicide films for silicon solar cells

Science.gov (United States)

Brahmi, Hatem; Ravipati, Srikanth; Yarali, Milad; Shervin, Shahab; Wang, Weijie; Ryou, Jae-Hyun; Mavrokefalos, Anastassios

2017-01-01

Highly conductive and transparent films of ultra-thin p-type nickel silicide films have been prepared by RF magnetron sputtering of nickel on silicon substrates followed by rapid thermal annealing in an inert environment in the temperature range 400-600 °C. The films are uniform throughout the wafer with thicknesses in the range of 3-6 nm. The electrical and optical properties are presented for nickel silicide films with varying thickness. The Drude-Lorentz model and Fresnel equations were used to calculate the dielectric properties, sheet resistance, absorption and transmission of the films. These ultrathin nickel silicide films have excellent optoelectronic properties for p-type contacts with optical transparencies up to 80% and sheet resistance as low as ~0.15 µΩ cm. Furthermore, it was shown that the use of a simple anti-reflection (AR) coating can recover most of the reflected light approaching the values of a standard Si solar cell with the same AR coating. Overall, the combination of ultra-low thickness, high transmittance, low sheet resistance and ability to recover the reflected light by utilizing standard AR coating makes them ideal for utilization in silicon based photovoltaic technologies as a p-type transparent conductor.

The text of the Agreement of 1 July 1983 concerning the Agency's assistance to Romania for the transfer of enriched uranium for irradiation studies in a research reactor

International Nuclear Information System (INIS)

1983-12-01

The full text of the Supply and Project Agreement of 1 July 1983 between the Agency and the Government of the Socialist Republic of Romania concerning the Agency's assistance for the transfer of moderately enriched uranium from the USSR for irradiation studies in a research reactor in Romania is presented

Bioassay method for Uranium in urine by Delay Neutron counting

International Nuclear Information System (INIS)

Suratman; Purwanto; Sukarman-Aminjoyo

1996-01-01

A bioassay method for uranium in urine by neutron counting has been studied. The aim of this research is to obtain a bioassay method for uranium in urine which is used for the determination of internal dose of radiation workers. The bioassay was applied to the artificially uranium contaminated urine. The weight of the contaminant was varied. The uranium in the urine was irradiated in the Kartini reactor core, through pneumatic system. The delayed neutron was counted by BF3 neutron counter. Recovery of the bioassay was between 69.8-88.8 %, standard deviation was less than 10 % and the minimum detection was 0.387 μg

Purification in the interaction between yttria mould and Nb-silicide-based alloy during directional solidification: A novel effect of yttrium

International Nuclear Information System (INIS)

Ma, Limin; Tang, Xiaoxia; Wang, Bin; Jia, Lina; Yuan, Sainan; Zhang, Hu

2012-01-01

Nb-silicide-based alloys were directionally solidified in yttria moulds. As a result of thermal dissociation of yttria, the alloys were slightly contaminated with oxygen, which caused a competitive oxidation between yttrium and hafnium. The addition of 0.15 at.% yttrium reduced the oxygen increment by 42%, because the buoyant inclusions concentrated around the top surface. The yttrium addition caused a significant purification of the interaction between the yttria mould and the Nb-silicide-based alloys during the directional solidification.

KEY RESULTS FROM IRRADIATION AND POST-IRRADIATION EXAMINATION OF AGR-1 UCO TRISO FUEL

Energy Technology Data Exchange (ETDEWEB)

Demkowicz, Paul A.; Hunn, John D.; Petti, David A.; Morris, Robert N.

2016-11-01

The AGR-1 irradiation experiment was performed as the first test of tristructural isotropic (TRISO) fuel in the US Advanced Gas Reactor Fuel Development and Qualification Program. The experiment consisted of 72 right cylinder fuel compacts containing approximately 3×105 coated fuel particles with uranium oxide/uranium carbide (UCO) fuel kernels. The fuel was irradiated in the Advanced Test Reactor for a total of 620 effective full power days. Fuel burnup ranged from 11.3 to 19.6% fissions per initial metal atom and time average, volume average irradiation temperatures of the individual compacts ranged from 955 to 1136°C. This paper focuses on key results from the irradiation and post-irradiation examination, which revealed a robust fuel with excellent performance characteristics under the conditions tested and have significantly improved the understanding of UCO coated particle fuel irradiation behavior within the US program. The fuel exhibited a very low incidence of TRISO coating failure during irradiation and post-irradiation safety testing at temperatures up to 1800°C. Advanced PIE methods have allowed particles with SiC coating failure to be isolated and meticulously examined, which has elucidated the specific causes of SiC failure in these specimens. The level of fission product release from the fuel during irradiation and post-irradiation safety testing has been studied in detail. Results indicated very low release of krypton and cesium through intact SiC and modest release of europium and strontium, while also confirming the potential for significant silver release through the coatings depending on irradiation conditions. Focused study of fission products within the coating layers of irradiated particles down to nanometer length scales has provided new insights into fission product transport through the coating layers and the role various fission products may have on coating integrity. The broader implications of these results and the application of

Fabrication of chamfered uranium-plutonium mixed carbide pellets

International Nuclear Information System (INIS)

Arai, Yasuo; Iwai, Takashi; Shiozawa, Kenichi; Handa, Muneo

1985-10-01

Chamfered uranium-plutonium mixed carbide pellets for high burnup irradiation test in JMTR were fabricated in glove boxes with purified argon gas. The size of die and punch in a press was decided from pellet densities and dimensions including the angle of chamfered parts. No chip or crack caused by adopting chamfered pellets was found in both pressing and sintering stages. In addition to mixed carbide pellets, uranium carbide pellets used as insulators were also successfully fabricated. (author)

Experimental study of Aspergillus flavus fungus from uranium mines

Energy Technology Data Exchange (ETDEWEB)

Kusak, V. (Ceskoslovenska Akademie Ved, Prague. Ustav Experimentalni Mediciny)

1982-06-01

Cultivation is discussed of fungus strain Aspergillus flavus obtained from materials from uranium mines. It was found that an addition of 0.6 g of uranium in form of uranyl acetate or of 0.6 g of thorium in form on thorium nitrate in 1000 ml of the standard medium had stimulating effects on the growth and sporulation of Aspergillus flavus. Irradiating the cultivated fungus through a polyethylene foil did not show a stimulating effect. It is stated that uranium and its daughters must be directly present in the culture medium for their stimulating effect on growth and sporulation to manifest itself.

RERTR programme. French situation in 1996

International Nuclear Information System (INIS)

Guidez, J.; Ballagny, A.

1996-01-01

This paper summarizes the status of the RERTR Programme in France in 1996. The reactors which are affected by the RERTR Programme are those that consume a considerable amount of fuel. These are the neutron beam reactors Orphee (at Saclay) and RHF (at Grenoble) and the irradiation reactors Osiris (at Saclay) and Siloe (at Grenoble). Plans to construct a new 100 MW, multi-purpose MTR reactor using LEU fuel at the Atomic Research Center at Cadarache are described. Two main topics of fuel research and development are summarized: (1) improving knowledge on the reference silicide fuel, including optimization of the manufacturing process, thermal properties, behavior with cladding failure, etc., and (2) research into low-enriched uranium fuel with a higher U-235 content in order to limit fuel assembly consumption and to improve performance levels. Tests planned in the research and development programme for the silicide fuel are tabulated. The fuel cycle option adopted by the CEA to get rid of spent fuel elements is reprocessing. The 'Caramel' fuel elements consumed in the Osiris reactor until 1995 are currently being reprocessed in CEA facilities at Marcoule. The UAl and UAlx fuel elements irradiated in Siloe and Orphee are currently being reprocessed in the Cogema facilities at Marcoule (Plant UP1). However, the U3Si2 fuel elements irradiated in the Osiris reactor since 1995 will be progressively sent to interim storage dry facilities located at Cadarache. (author)

Thermal expansion and elastic moduli of the silicide based intermetallic alloys Ti5Si3(X) and Nb5Si3

International Nuclear Information System (INIS)

Zhang, L.; Wu, J.

1997-01-01

Silicides are among those potential candidates for high temperature application because of their high melting temperature, low density and good oxidation resistance. Recent interest is focused on molybdenum silicides and titanium silicides. Extensive investigation has been carried out on MoSi 2 , yet comparatively less work was performed on titanium silicides such as Ti 5 Si 3 and Ti 3 and TiSi 2 which are of lower density than MoSi 2 . Fundamental understanding of the titanium silicides' properties for further evaluation their potential for practical application are thus needed. The thermal expansion coefficients and elastic moduli of intermetallic compounds are two properties important for evaluation as a first step. The thermal expansion determines the possible stress that might arise during cooling for these high melting point compounds, which is crucial to the preparation of defect free specimens; and the elastic moduli are usually reflections of the cohesion in crystal. In Frommeyer's work and some works afterwards, the coefficients of thermal expansion were measured on both polycrystalline and single crystal Ti 5 Si 3 . The elastic modulus of polycrystalline Ti 5 Si 3 was measured by Frommeyer and Rosenkranz. However, in the above works, the referred Ti 5 Si 3 was the binary one, no alloying effect has been reported on this matter. Moreover, the above parameters (coefficient of thermal expansion and elastic modulus) of Nb 5 Si 3 remain unreported so far. In this paper, the authors try to extend the knowledge of alloyed Ti 5 Si 3 compounds with Nb and Cr additions. Results on the coefficients of thermal expansion and elastic moduli of Ti 5 Si 3 compounds and Nb 5 Si 3 are presented and the discussion is focused on the alloying effect

Core-hole effects in the x-ray-absorption spectra of transition-metal silicides

NARCIS (Netherlands)

WEIJS, PJW; CZYZYK, MT; VANACKER, JF; SPEIER, W; GOEDKOOP, JB; VANLEUKEN, H; HENDRIX, HJM; DEGROOT, RA; VANDERLAAN, G; BUSCHOW, KHJ; WIECH, G; FUGGLE, JC

1990-01-01

We report systematic differences between the shape of the Si K x-ray-absorption spectra of transition-metal silicides and broadened partial densities of Si p states. We use a variety of calculations to show that the origin of these discrepancies is the core-hole potential appropriate to the final

A Study on Characterization of Light-Induced Electroless Plated Ni Seed Layer and Silicide Formation for Solar Cell Application

Science.gov (United States)

Takaloo, Ashkan Vakilipour; Joo, Seung Ki; Es, Firat; Turan, Rasit; Lee, Doo Won

2018-03-01

Light-induced electroless plating (LIEP) is an easy and inexpensive method that has been widely used for seed layer deposition of Nickel/Copper (Ni/Cu)-based metallization in the solar cell. In this study, material characterization aspects of the Ni seed layer and Ni silicide formation at different bath conditions and annealing temperatures on the n-side of a silicon diode structure have been examined to achieve the optimum cell contacts. The effects of morphology and chemical composition of Ni film on its electrical conductivity were evaluated and described by a quantum mechanical model. It has been found that correlation exists between the theoretical and experimental conductivity of Ni film. Residual stress and phase transformation of Ni silicide as a function of annealing temperature were evaluated using Raman and XRD techniques. Finally, transmission line measurement (TLM) technique was employed to determine the contact resistance of Ni/Si stack after thermal treatment and to understand its correlation with the chemical-structural properties. Results indicated that low electrical resistive mono-silicide (NiSi) phase as low as 5 mΩ.cm2 was obtained.

Contribution to the micrographic study of uranium and its alloys; Contribution a l'etude micrographique de l'uranium et de ses alliages

Energy Technology Data Exchange (ETDEWEB)

Monti, H [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1956-06-15

The present report is the result of research carried out by the radio metallurgy section, to perfect micrographic techniques applicable to the study of samples of irradiated uranium. In the first part of this work, two polishing baths are developed, having the qualities with a minimum of disadvantages inherent in their respective compositions: they are, on the one hand perchloric acid-ethanol mixtures, and on the other hand a phospho-chromic-ethanol bath. In the chapter following, the micrographic attack of uranium is studied. The only satisfactory process is oxidation by cathode bombardment forming epitaxic layers. In the third chapter, an attempt is made to characterise the different surface states of the uranium by dissolution potential measurements and electronic diffraction. In the fourth chapter are given some examples of the application of these techniques to the micrographic study of various uranium alloys. In an appendix, it is shown how the chemical oxidation after phospho-chromic-alcohol polishing allows the different inclusions present in the molten uranium to be distinguished. By X-ray diffraction, uranium monocarbide and mononitride inclusions in particular are characterised. (author) [French] Le present rapport est le resultat de recherches effectuees au service de radiometallurgie pour la mise au point de techniques micrographiques applicables a l'etude d'echantillons d'uranium irradie. Dans la premiere partie de ce travail, nous mettons au point deux bains de polissage qui presentent les qualites inherentes a leur composition respective, avec le minimum d'inconvenients: ce sont d'une part des melanges acide perchlorique-ethanol, et d'autre part un bain phospho-chromique-ethanol. Dans le chapitre suivant, nous etudions l'attaque micrographique de l'uranium. Seul le procede d'oxydation par bombardement cathodique formant des couches epitaxiques, est satisfaisant. Dans le troisieme chapitre, nous essayons de caracteriser les differents etats de

Heavy-ion irradiation induced diamond formation in carbonaceous materials

International Nuclear Information System (INIS)

Daulton, T. L.

1999-01-01

The basic mechanisms of metastable phase formation produced under highly non-equilibrium thermodynamic conditions within high-energy particle tracks are investigated. In particular, the possible formation of diamond by heavy-ion irradiation of graphite at ambient temperature is examined. This work was motivated, in part, by earlier studies which discovered nanometer-grain polycrystalline diamond aggregates of submicron-size in uranium-rich carbonaceous mineral assemblages of Precambrian age. It was proposed that the radioactive decay of uranium formed diamond in the fission particle tracks produced in the carbonaceous minerals. To test the hypothesis that nanodiamonds can form by ion irradiation, fine-grain polycrystalline graphite sheets were irradiated with 400 MeV Kr ions. The ion irradiated graphite (and unirradiated graphite control) were then subjected to acid dissolution treatments to remove the graphite and isolate any diamonds that were produced. The acid residues were then characterized by analytical and high-resolution transmission electron microscopy. The acid residues of the ion-irradiated graphite were found to contain ppm concentrations of nanodiamonds, suggesting that ion irradiation of bulk graphite at ambient temperature can produce diamond

Irradiation testing of high density uranium alloy dispersion fuels

International Nuclear Information System (INIS)

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

1997-10-01

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

Contribution to the study of defects created by α particles in uranium at 4.2 K

International Nuclear Information System (INIS)

Raharinaivo, A.L.

1969-01-01

A device is described for the irradiation, in liquid helium, of metallic strips with α particles produced by radioactive sources. It has thereby been possible to measure changes in resistivity of variously treated uranium samples (cold- worked, annealed, previously exposed to neutrons, etc. ) as a function of the irradiation flux. The annealings carried out after irradiation compare favorably to those effected after a quenching from 100 to 4 K (JOUSSET experiments). The results are discussed; it is concluded that a defect, very probably of the interstitial type, is mobile in uranium at temperatures below 5 K. (author) [fr

Controlling the formation and stability of ultra-thin nickel silicides - An alloying strategy for preventing agglomeration

Science.gov (United States)

Geenen, F. A.; van Stiphout, K.; Nanakoudis, A.; Bals, S.; Vantomme, A.; Jordan-Sweet, J.; Lavoie, C.; Detavernier, C.

2018-02-01

The electrical contact of the source and drain regions in state-of-the-art CMOS transistors is nowadays facilitated through NiSi, which is often alloyed with Pt in order to avoid morphological agglomeration of the silicide film. However, the solid-state reaction between as-deposited Ni and the Si substrate exhibits a peculiar change for as-deposited Ni films thinner than a critical thickness of tc = 5 nm. Whereas thicker films form polycrystalline NiSi upon annealing above 450 ° C , thinner films form epitaxial NiSi2 films that exhibit a high resistance toward agglomeration. For industrial applications, it is therefore of utmost importance to assess the critical thickness with high certainty and find novel methodologies to either increase or decrease its value, depending on the aimed silicide formation. This paper investigates Ni films between 0 and 15 nm initial thickness by use of "thickness gradients," which provide semi-continuous information on silicide formation and stability as a function of as-deposited layer thickness. The alloying of these Ni layers with 10% Al, Co, Ge, Pd, or Pt renders a significant change in the phase sequence as a function of thickness and dependent on the alloying element. The addition of these ternary impurities therefore changes the critical thickness tc. The results are discussed in the framework of classical nucleation theory.

Semi-continuous on-line uranium ore analyzer

International Nuclear Information System (INIS)

Campbell, P.; Gurdy, E.M.; Hatchowski, L.

1984-01-01

The efficient process control of a uranium mill and its associated mining operation requires a nearly continuous knowledge of the uranium concentration in the ore. Generally, the approach is to use laboratory assays of grab samples from the mill feed belt. In some cases, elaborate and expensive systems have been used to ensure random sampling, but even with these systems, mass-balance discrepancies still exist. There is a requirement for an on-line instrument that can give a prompt, accurate analysis of a large portion of the feed stock. The authors have recently evaluated a laboratory system that achieves these goals using neutron activation and delayed neutron counting. The development of the on-line uranium ore analyzer is a consequence of previous work done at the Whiteshell Nuclear Research Establishment (WNRE) (Campbell et al., 1978 and 1981), and is based on the emission of delayed neutrons after the neutron irradiation of fissionable materials. The mechanism of delayed neutron emission has been described (Keepin, 1965), but briefly is as follows. The fission fragments resulting from irradiation are in an excited state. Certain of these delayed fission products, precursors, decay to a more stable state by the emission of a delayed neutron. The delayed neutrons can be divided into six groups with effective half-lives ranging from 0.2-55 sec. It is important to note that more than 50% of the delayed neutrons are emitted within the first 6 sec after irradiation; this has an influence on the design of the analysis instrument

Study of optical and luminescence properties of silicon — semiconducting silicide — silicon multilayer nanostructures

International Nuclear Information System (INIS)

Galkin, N.G.; Galkin, K.N.; Dotsenko, S.A.; Goroshko, D.L.; Shevlyagin, A.V.; Chusovitin, E.A.; Chernev, I.M.

2017-01-01

By method of in situ differential spectroscopy it was established that at the formation of monolayer Fe, Cr, Ca, Mg silicide and Mg stannide islands on the atomically clean silicon surface an appearance of loss peaks characteristic for these materials in the energy range of 1.1-2.6 eV is observed. An optimization of growth processes permit to grow monolithic double nanoheterostructures (DNHS) with embedded Fe, Cr and Ca nanocrystals, and also polycrystalline DNHS with NC of Mg silicide and Mg stannide and Ca disilicide. By methods of optical spectroscopy and Raman spectroscopy it was shown that embedded NC form intensive peaks in the reflectance spectra at energies up to 2.5 eV and Raman peaks. In DNS with β-FeSi2 NC a photoluminescence and electroluminescence at room temperature were firstly observed.

Neutronic analysis of the JMTR with LEU fuel and burnable poison

Energy Technology Data Exchange (ETDEWEB)

Nagaoka, Yoshiharu; Oyamada, Rokuro [Japan Atomic Energy Research Institute, Oarai-machi Ibaraki-ken (Japan); Matos, J E; Woodruff, W L [Argonne National Laboratory, Argonne, IL (United States)

1985-07-01

The results of neutronics calculations are presented for the JMTR equilibrium core with LEU silicide fuel, boron and cadmium burnable poisons in the sideplates, and a cycle length of 24 days instead of 11 days with the current HEU fuel. The data indicate that several options are feasible provided that silicide fuels with high uranium densities are successfully demonstrated and licensed (author)

Neutronic analysis of the JMTR with LEU fuel and burnable poison

International Nuclear Information System (INIS)

Nagaoka, Yoshiharu; Oyamada, Rokuro; Matos, J.E.; Woodruff, W.L.

1985-01-01

The results of neutronics calculations are presented for the JMTR equilibrium core with LEU silicide fuel, boron and cadmium burnable poisons in the sideplates, and a cycle length of 24 days instead of 11 days with the current HEU fuel. The data indicate that several options are feasible provided that silicide fuels with high uranium densities are successfully demonstrated and licensed (author)

Construction of an apparatus for measuring the low-temperature thermal conductivity before and after neutron irradiation. Application to uranium dioxide (1963); Realisation d'un appareil pour la mesure de la conductibilite thermique a basse temperature avant et apres irradiation neutronique. Application au dioxyde d'uranium (1963)

Energy Technology Data Exchange (ETDEWEB)

Bethoux, O [Commisariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

1963-09-15

An apparatus has been studied and built which makes it possible to alternatively irradiate a sample at room temperature in the reactor 'Melusine' at the Grenoble Nuclear Research Centre, and to measure its thermal conductivity between 20 and 100 deg. K in perfect safety. The results obtained on UO{sub 2} have made it possible on the one hand to check experimentally that the spin-phonon diffusion leads to a thermal resistance independent of temperature above 30 deg. K, and on the other hand to propose a simple theory which takes into count the role played by the damage due to U-235 fission products in the decrease of thermal conductivity after irradiation. (author) [French] Un appareil permettant alternativement d'irradier un echantillon a temperature ambiante dans le reacteur ''Melusine'' du C.E.N.G., et de mesurer sa conductibilite thermique entre 20 et 100 deg. K en toute securite, a ete etudie et construit Les resultats obtenus sur UO{sub 2} ont permis, d'une part, de verifier experimentalement que la diffusion spin-phonon conduit a une resistance thermique independante de la temperature au-dessus de 30 deg. K, et, d'autre part, de proposer une theorie simple tenant compte du role joue par les degats dus aux produits de fission de l'uranium 235, dans la deterioration de la conductibilite thermique apres irradiation. (auteur)

Irradiation testing of high-density uranium alloy dispersion fuels

International Nuclear Information System (INIS)

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

1997-01-01

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

Role of Ti 3 Al/silicides on tensile properties of Timetal 834 at ...

Indian Academy of Sciences (India)

Extremely fine coherent precipitates of ordered Ti3Al and relatively coarse incoherent precipitates of 2 silicide exist together in the near -titanium alloy, Timetal 834, in the dual phase matrix of primary and transformed . In order to assess the role of these precipitates, three heat treatments viz. WQ, WQ–A and WQ–OA, ...

Experimental study of Aspergillus flavus fungus from uranium mines

International Nuclear Information System (INIS)

Kusak, V.

1982-01-01

Cultivation is discussed of fungus strain Aspergillus flavus obtained from materials from uranium mines. It was found that an addition of 0.6 g of uranium in form of uranyl acetate or of 0.6 g of thorium in form on thorium nitrate in 1000 ml of the standard medium had stimulating effects on the growth and sporulation of Aspergillus flavus. Irradiating the cultivated fungus through a polyethylene foil did not show a stimulating effect. It is stated that uranium and its daughters must be directly present in the culture medium for their stimulating effect on growth and sporulation to manifest itself. (H.S.)

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

International Nuclear Information System (INIS)

Iwai, Takashi; Shimizu, Michio; Nakagawa, Tetsuya

1990-02-01

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

Processing of irradiated, enriched uranium fuels at the Savannah River Plant

Energy Technology Data Exchange (ETDEWEB)

Hyder, M L; Perkins, W C; Thompson, M C; Burney, G A; Russell, E R; Holcomb, H P; Landon, L F

1979-04-01

Uranium fuels containing /sup 235/U at enrichments from 1.1% to 94% are processed and recovered, along with neptunium and plutonium byproducts. The fuels to be processed are dissolved in nitric acid. Aluminum-clad fuels are disssolved using a mercury catalyst to give a solution rich in aluminum. Fuels clad in more resistant materials are dissolved in an electrolytic dissolver. The resulting solutions are subjected to head-end treatment, including clarification and adjustment of acid and uranium concentration before being fed to solvent extraction. Uranium, neptunium, and plutonium are separated from fission products and from one another by multistage countercurrent solvent extraction with dilute tri-n-butyl phosphate in kerosene. Nitric acid is used as the salting agent in addition to aluminum or other metal nitrates present in the feed solution. Nuclear safety is maintained through conservative process design and the use of monitoring devices as secondary controls. The enriched uranium is recovered as a dilute solution and shipped off-site for further processing. Neptunium is concentrated and sent to HB-Line for recovery from solution. The relatively small quantities of plutonium present are normally discarded in aqueous waste, unless the content of /sup 238/Pu is high enough to make its recovery desirable. Most of the /sup 238/Pu can be recovered by batch extraction of the waste solution, purified by counter-current solvent extraction, and converted to oxide in HB-Line. By modifying the flowsheet, /sup 239/Pu can be recovered from low-enriched uranium in the extraction cycle; neptunium is then not recovered. The solvent is subjected to an alkaline wash before reuse to remove degraded solvent and fission products. The aqueous waste is concentrated and partially deacidified by evaporation before being neutralized and sent to the waste tanks; nitric acid from the overheads is recovered for reuse.

Processing of irradiated, enriched uranium fuels at the Savannah River Plant

International Nuclear Information System (INIS)

Hyder, M.L.; Perkins, W.C.; Thompson, M.C.; Burney, G.A.; Russell, E.R.; Holcomb, H.P.; Landon, L.F.

1979-04-01

Uranium fuels containing 235 U at enrichments from 1.1% to 94% are processed and recovered, along with neptunium and plutonium byproducts. The fuels to be processed are dissolved in nitric acid. Aluminum-clad fuels are disssolved using a mercury catalyst to give a solution rich in aluminum. Fuels clad in more resistant materials are dissolved in an electrolytic dissolver. The resulting solutions are subjected to head-end treatment, including clarification and adjustment of acid and uranium concentration before being fed to solvent extraction. Uranium, neptunium, and plutonium are separated from fission products and from one another by multistage countercurrent solvent extraction with dilute tri-n-butyl phosphate in kerosene. Nitric acid is used as the salting agent in addition to aluminum or other metal nitrates present in the feed solution. Nuclear safety is maintained through conservative process design and the use of monitoring devices as secondary controls. The enriched uranium is recovered as a dilute solution and shipped off-site for further processing. Neptunium is concentrated and sent to HB-Line for recovery from solution. The relatively small quantities of plutonium present are normally discarded in aqueous waste, unless the content of 238 Pu is high enough to make its recovery desirable. Most of the 238 Pu can be recovered by batch extraction of the waste solution, purified by counter-current solvent extraction, and converted to oxide in HB-Line. By modifying the flowsheet, 239 Pu can be recovered from low-enriched uranium in the extraction cycle; neptunium is then not recovered. The solvent is subjected to an alkaline wash before reuse to remove degraded solvent and fission products. The aqueous waste is concentrated and partially deacidified by evaporation before being neutralized and sent to the waste tanks; nitric acid from the overheads is recovered for reuse

Study on the electrolytic reduction of Uranium-VI to Uranium-IV in a nitrate system

International Nuclear Information System (INIS)

Araujo, B.F. de; Almeida, S.G. de; Forbicini, S.; Matsuda, H.T.; Araujo, J.A. de.

1981-05-01

The determination of the best conditions to prepare hydrazine stabilized uranium (IV) nitrate solutions for utilization in Purex flowsheets is dealt with. Electrolytic reduction of U(VI) has been selected as the basic method, using an open electrolytic cell with titanum and platinum electrodes. The hydrazine concentration, the current density, acidity, U(VI) concentration and reduction time were the parameters studied and U(IV)/U(VI) ratio was used to evaluate the degree of reduction. From the results it could be concluded that the technique is reliable. The U(IV) solutions remains constant for at least two weeks and can be used in the chemical processing of irradiated uranium fuels. (Author) [pt

Production of Fission Product 99Mo using High-Enriched Uranium Plates in Polish Nuclear Research Reactor MARIA: Technology and Neutronic Analysis

Directory of Open Access Journals (Sweden)

Jaroszewicz Janusz

2014-07-01

Full Text Available The main objective of 235U irradiation is to obtain the 99mTc isotope, which is widely used in the domain of medical diagnostics. The decisive factor determining its availability, despite its short lifetime, is a reaction of radioactive decay of 99Mo into 99mTc. One of the possible sources of molybdenum can be achieved in course of the 235U fission reaction. The paper presents activities and the calculation results obtained upon the feasibility study on irradiation of 235U targets for production of 99Mo in the MARIA research reactor. Neutronic calculations and analyses were performed to estimate the fission products activity for uranium plates irradiated in the reactor. Results of dummy targets irradiation as well as irradiation uranium plates have been presented. The new technology obtaining 99Mo is based on irradiation of high-enriched uranium plates in standard reactor fuel channel and calculation of the current fission power generation. Measurements of temperatures and the coolant flow in the molybdenum installation carried out in reactor SAREMA system give online information about the current fission power generated in uranium targets. The corrective factors were taken into account as the heat generation from gamma radiation from neighbouring fuel elements as well as heat exchange between channels and the reactor pool. The factors were determined by calibration measurements conducted with aluminium mock-up of uranium plates. Calculations of fuel channel by means of REBUS code with fine mesh structure and libraries calculated by means of WIMS-ANL code were performed.

Progress in developing very-high-density low-enriched-uranium fuels

International Nuclear Information System (INIS)

Snelgrove, J.L.; Hofman, G.L.; Meyer, M.K.; Hayes, S.L.; Wiencek, T.C.; Strain, R.V.

1999-01-01

Preliminary results from the postirradiation examinations of microplates irradiated in the RERTR-1 and -2 experiments in the ATR have shown several binary and ternary U-Mo alloys to be promising candidates for use in aluminum-based dispersion fuels with uranium densities up to 8 to 9 g/cm 3 . Ternary alloys of uranium, niobium, and zirconium performed poorly, however, both in terms of fuel/matrix reaction and fission-gas-bubble behavior, and have been dropped from further study. Since irradiation temperatures achieved in the present experiments (approximately 70 deg. C) are considerably lower than might be experienced in a high-performance reactor, a new experiment is being planned with beginning-of-cycle temperatures greater than 200 deg. C in 8-g U/cm 3 fuel. (author)

Depleted uranium - influence on the health and environment; Ochudobneny uran - vplyva na zdravie a zivotne prostredie

Energy Technology Data Exchange (ETDEWEB)

Rosskopfova, O [Katedra jadrovej chemie, Prirodovedecka fakulta, Univerzita Komenskeho, Bratislava (Slovakia)

2002-07-01

The uranium as radioactive element occurs in low concentrations in all components of environment. In the sample of natural uranium the isotope U-235 has the highest share (99.27 weight per cent). Chemical toxicity of uranium is comparable with toxicity of the elements like As and Pb. Depleted uranium is adjoining product in the production of enriched uranium, which is required in the production of the nuclear fuel and in the production of material used in nuclear arms. It mainly includes isotope U-238, and the content of isotopes U-235 and U-234 is sharply lowered. According to NRC depleted uranium is defined like uranium, in which percentile share of isotope U-235 is less than 0.711 weight percent. The activity of depleted uranium from viewpoint of external irradiation does not represent higher risk. Much higher risk for man represent the neurotoxic effects of uranium, which can get into human body by inhaling of dispersed particles, of contaminated dust and aerosols from atmosphere or by consumption of contaminated foodstuffs and water. Basic dangerous of irradiation by depleted uranium are mainly aerosols, which increase the probability of occurrence of lung cancer. The next dangerous is the damage of another organs like kidneys, liver and bones, where these aerosols are transported by blood like oxides from the lungs. In the environmental parts because of presence of natural uranium the depleted uranium is difficultly identifiably by standard detective methods. Thus it is necessary to use suitable radiochemical separative methods in combination with suitable detective method. (author)

Techno-economic study on conversion of SAFARI-1 to LEU silicide fuel

International Nuclear Information System (INIS)

Ball, G.; Malherbe, F.J.

2004-01-01

This paper marks the conclusion of the techno-economic study into the conversion of SAFARI-1 reactor in South Africa to LEU silicide fuel. Several different fuel types were studied and their characteristics compared to the current HEU fuel. The technical feasibility of operating SAFARI-1 with the different fuels as well as the overall economic impact of the fuels is discussed and conclusions drawn.(author)

Determination by neutron activation of the uranium-235 concentration in uranium oxides; Determination par activation neutronique de la concentration d'uranium-235 dans des oxydes d'urane

Energy Technology Data Exchange (ETDEWEB)

May, S; Leveque, P [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

1959-07-01

Classical methods of measuring isotopic abundance have the disadvantage of being long and of requiring chemical separation. A non-destructive method of measuring the uranium-235 content is described. It is based on an overall measurement of the short lived fission product activity formed during a 15 s neutron irradiation. The precision is of the order {+-} 1.5 per cent for 20 per cent enriched samples. The error due to the contribution from fast fission is discussed in detail. (author) [French] Les methodes classiques de mesure de l'abondance isotopique presentent le gros inconvenient d'etre longues et de necessiter des separations chimiques. Nous exposons une methode non destructive de mesure de la concentration d'uranium-235. Elle est basee sur la mesure globale de l'activite des produits de fission de courte periode formes par une irradiation neutronique de 15 s de l'echantillon. La precision est de l'ordre de {+-} 1,5 pour cent pour des echantillons enrichis jusqu'a 20 pour cent. L'erreur a la contribution de la fission rapide est discutee en detail. (auteur)

Moissanite (SiC) with metal-silicide and silicon inclusions from tuff of Israel: Raman spectroscopy and electron microscope studies

Science.gov (United States)

Dobrzhinetskaya, Larissa; Mukhin, Pavel; Wang, Qin; Wirth, Richard; O'Bannon, Earl; Zhao, Wenxia; Eppelbaum, Lev; Sokhonchuk, Tatiana

2018-06-01

Here, we present studies of natural SiC that occurs in situ in tuff related to the Miocene alkaline basalt formation deposited in northern part of Israel. Raman spectroscopy, SEM and FIB-assisted TEM studies revealed that SiC is primarily hexagonal polytypes 4H-SiC and 6H-SiC, and that the 4H-SiC polytype is the predominant phase. Both SiC polytypes contain crystalline inclusions of silicon (Sio) and inclusions of metal-silicide with varying compositions (e.g. Si58V25Ti12Cr3Fe2, Si41Fe24Ti20Ni7V5Zr3, and Si43Fe40Ni17). The silicides crystal structure parameters match Si2TiV5 (Pm-3m space group, cubic), FeSi2Ti (Pbam space group, orthorhombic), and FeSi2 (Cmca space group, orthorhombic) respectively. We hypothesize that SiC was formed in a local ultra-reduced environment at respectively shallow depths (60-100 km), through a reaction of SiO2 with highly reducing fluids (H2O-CH4-H2-C2H6) arisen from the mantle "hot spot" and passing through alkaline basalt magma reservoir. SiO2 interacting with the fluids may originate from the walls of the crustal rocks surrounding this magmatic reservoir. This process led to the formation of SiC and accompanied by the reducing of metal-oxides to native metals, alloys, and silicides. The latter were trapped by SiC during its growth. Hence, interplate "hot spot" alkali basalt volcanism can now be included as a geological environment where SiC, silicon, and silicides can be found.

Molybdenum-UO2 cermet irradiation at 1145 K.

Science.gov (United States)

Mcdonald, G.

1971-01-01

Two molybdenum-uranium dioxide cermet fuel pins with molybdenum clad were fission-heated in a forced-convection helium coolant for sufficient time to achieve 5.3% burnup. The cermet core contained 20 wt % of 93.2% enriched uranium dioxide. The results were as follows: there was no visible change in the appearance of the molybdenum clad during irradiation; the maximum increase in diameter of the fuel pins was 0.8%; there was no migration of uranium dioxide along grain boundaries and no evident interaction between molybdenum and uranium dioxide; and, finally, approximately 12% of the fission gas formed was released from the cermet core into the gas plenum.

Radiation damage of uranium-thorium oxide, irradiated in water

International Nuclear Information System (INIS)

Bloem, P.J.C.; Nagel, W.; Plas, T. van der; Kema, N.V.

1977-01-01

A suspension in water of spherical particles of UO 2 -ThO 2 with diameter 5μm has been considered as the working fluid in an aqueous, homogeneous, thermal nuclear reactor. Irradiation experiments have shown that these particles suffer a gradual breakdown when irradiated in water. This behaviour is markedly different from that shown on irradiation in absence of water. As damage was defined the amount of solid dissolved by an etching liquid. Electron microscopic pictures showed that at higher irradiation temperatures in water the actual damage was larger than the etching values indicated. (orig.) [de

Contribution to the micrographic study of uranium and its alloys; Contribution a l'etude micrographique de l'uranium et de ses alliages

Energy Technology Data Exchange (ETDEWEB)

Monti, H. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1956-06-15

The present report is the result of research carried out by the radio metallurgy section, to perfect micrographic techniques applicable to the study of samples of irradiated uranium. In the first part of this work, two polishing baths are developed, having the qualities with a minimum of disadvantages inherent in their respective compositions: they are, on the one hand perchloric acid-ethanol mixtures, and on the other hand a phospho-chromic-ethanol bath. In the chapter following, the micrographic attack of uranium is studied. The only satisfactory process is oxidation by cathode bombardment forming epitaxic layers. In the third chapter, an attempt is made to characterise the different surface states of the uranium by dissolution potential measurements and electronic diffraction. In the fourth chapter are given some examples of the application of these techniques to the micrographic study of various uranium alloys. In an appendix, it is shown how the chemical oxidation after phospho-chromic-alcohol polishing allows the different inclusions present in the molten uranium to be distinguished. By X-ray diffraction, uranium monocarbide and mononitride inclusions in particular are characterised. (author) [French] Le present rapport est le resultat de recherches effectuees au service de radiometallurgie pour la mise au point de techniques micrographiques applicables a l'etude d'echantillons d'uranium irradie. Dans la premiere partie de ce travail, nous mettons au point deux bains de polissage qui presentent les qualites inherentes a leur composition respective, avec le minimum d'inconvenients: ce sont d'une part des melanges acide perchlorique-ethanol, et d'autre part un bain phospho-chromique-ethanol. Dans le chapitre suivant, nous etudions l'attaque micrographique de l'uranium. Seul le procede d'oxydation par bombardement cathodique formant des couches epitaxiques, est satisfaisant. Dans le troisieme chapitre, nous essayons

Neutronic analysis of the JMTR with LEU fuel and burnable poison

International Nuclear Information System (INIS)

Nagaoka, Yoshiharu; Oyamada, Rokuro; Matos, J.E.; Woodruff, W.L.

1984-01-01

The results of neutronics calculations are presented for the JMTR equilibrium core with LEU silicide fuel, boron and cadmium burnable poisons in the sideplates, and a cycle length of 24 days instead of 11 days with the current HEU fuel. The data indicate that several options are feasible provided that silicide fuels with high uranium densities are successfully demonstrated and licensed. 2 refs., 10 figs., 5 tabs

Impact of Nickel silicide Rear Metallization on Series Resistance of Crystalline Silicon Solar Cells

KAUST Repository

Bahabry, Rabab R; Hanna, Amir N; Kutbee, Arwa T; Gumus, Abdurrahman; Hussain, Muhammad Mustafa

2018-01-01

the electrical characteristics of nickel mono-silicide (NiSi)/Cu-Al ohmic contact on the rear side of c-Si solar cells. We observe a significant enhancement in the fill factor of around 6.5% for NiSi/Cu-Al rear contacts leading to increasing the efficiency by 1.2

A study on improving the performance of a research reactor's equilibrium core

Directory of Open Access Journals (Sweden)

Muhammad Atta

2013-01-01

Full Text Available Utilizing low enriched uranium silicide fuel (U3Si2-Al of existing uranium density (3.285 g/cm3, different core configurations have been studied in search of an equilibrium core with an improved performance for the Pakistan Research Reactor-1. Furthermore, we have extended our analysis to the performance of higher density silicide fuels with a uranium density of 4.0 and 4.8 U g/cm3. The criterion used in selecting the best performing core was that of “unit flux time cycle length per 235U mass per cycle”. In order to analyze core performance by improving neutron moderation, utilizing higher-density fuel, the effect of the coolant channel width was also studied by reducing the number of plates in the standard/control fuel element. Calculations employing computer codes WIMSD/4 and CITATION were performed. A ten energy group structure for fission neutrons was used for the generation of microscopic cross-sections through WIMSD/4. To search the equilibrium core, two-dimensional core modelling was performed in CITATION. Performance indicators have shown that the higher-density uranium silicide-fuelled core (U density 4.8 g/cm3 without any changes in standard/control fuel elements, comprising of 15 standard and 4 control fuel elements, is the best performing of all analyzed cores.

Distinction between natural and depleted uranium using instrumental neutron activation analysis

International Nuclear Information System (INIS)

Haddad, Kh.

2008-01-01

A convenient method to discriminate between natural and depleted uranium samples was developed in this work. Traces of natural and depleted uranium were irradiated separately and the ratios of 95 Zr/ 103 Ru, 239 Np/ 95 Zr, 239 Np/ 103 Ru were measured. The results show that these ratios can be used as indicators of the uranium isotopic composition of the sample. These ratios are independent of the secular equilibrium of the 238 U with its daughters in the sample and indicate the isotopic composition for trace amounts. Date and truffle samples has been analysed also using this method. Results show that the uranium content in this product was less than the detection limit.(author)

Pulsed CO laser for isotope separation of uranium

Energy Technology Data Exchange (ETDEWEB)

Baranov, Igor Y.; Koptev, Andrey V. [Rocket-Space Technics Department, Baltic State Technical University, 1, 1st Krasnoarmeyskaya st.,St. Petersburg, 190005 (Russian Federation)

2012-07-30

This article proposes a technical solution for using a CO laser facility for the industrial separation of uranium used in the production of fuel for nuclear power plants, employing a method of laser isotope separation of uranium with condensation repression in a free jet. The laser operation with nanosecond pulse irradiation can provide an acceptable efficiency in the separating unit and a high efficiency of the laser with the wavelength of 5.3 {mu}m. In the present work we also introduce a calculation model and define the parameters of a mode-locked CO laser with a RF discharge in the supersonic stream. The average pulsed CO laser power of 3 kW is sufficient for efficient industrial isotope separation of uranium in one stage.

Thermal behavior analysis of U-Mo/Al dispersion fuel

Energy Technology Data Exchange (ETDEWEB)

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

2004-07-01

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

Thermal behavior analysis of U-Mo/Al dispersion fuel

International Nuclear Information System (INIS)

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

2004-01-01

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

CEA fuel pencil qualification under irradiation: from component conception to fuel assembly irradiation in a power reactor

International Nuclear Information System (INIS)

Marin, J.-F.; Pillet, Claude; Francois, Bernard; Morize, Pierre; Petitgrand, Sylvie; Atabek, R.-M.; Houdaille, Brigitte.

1981-06-01

Fabrication of fuel pins made of uranium oxide pellets and of a zircaloy 4 cladding is described. Irradiation experiment results are given. Thermomechanical behavior of the fuel pin in a power reactor is examined [fr

Waste Photovoltaic Panels for Ultrapure Silicon and Hydrogen through the Low-Temperature Magnesium Silicide.

Czech Academy of Sciences Publication Activity Database

Dytrych, Pavel; Bumba, Jakub; Kaštánek, František; Fajgar, Radek; Koštejn, Martin; Šolcová, Olga

Roč. 56, č. 45 ( 2017 ), s. 12863-12869 ISSN 0888-5885 R&D Projects: GA ČR GA15-14228S Institutional support: RVO:67985858 Keywords : magnesium silicide * waste photovoltaic panels * ultrapure silicon Subject RIV: CI - Industrial Chemistry, Chemical Engineering OBOR OECD: Chemical process engineering Impact factor: 2.843, year: 2016

Progress in doping of ruthenium silicide (Ru2Si3)

International Nuclear Information System (INIS)

Vining, C.B.; Allevato, C.E.

1992-01-01

This paper reports that ruthenium silicide (Ru 2 Si 3 ) is currently under development as a promising thermoelectric material suitable for space power applications. Key to realizing the potentially high figure of merit values of this material is the development of appropriate doping techniques. In this study, manganese and iridium have been identified as useful p- and n-type dopants, respectively. Resistivity values have been reduced by more than 3 orders of magnitude. Anomalous Hall effect results, however, complicate interpretation of some of the results and further effort is required to achieve optimum doping levels

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

International Nuclear Information System (INIS)

Cronenberg, A.W.; Rest, J.

1990-01-01

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

The contribution of radioisotopes in secular equilibrium in the transport index of fissile uranium compounds in different enrichments

International Nuclear Information System (INIS)

Silva, Teresinha de Moraes da; Sordi, Gian M.A.A.

2008-01-01

Full text: This work shows the contribution of radioisotopes in secular equilibrium in the transport index (TI) of some fissile uranium compounds: uranium oxides UO 2 , U 3 O 8 and uranium silicide U 3 Si 2 , taking into account the different enrichment grades.The range of enrichment (E%) studied was 3,4,5,7,10,20,30,40,50,93 and 100. Initially, the cell of optimum moderation ratio was built, since it represents the most reactive of the system (consisting of uranium), with maximum infinitive multiplication factor k∞, in certain concentration of uranium for each enrichment. This was made using the computer program Gamtec II. The critical radius of a sphere was calculated for a cell of optimum moderation ratio, in order to calculate the critical mass of the uranium compound or of the uranium element for each specific enrichment. For this the program Citation was used. In this study it was calculated the smallest critical mass of the uranium compound or the smallest critical mass of the uranium element. The objective was to match the largest mass of the uranium with each specific enrichment. The largest safety mass corresponds to 45% the critical mass the compound uranium or uranium element. Then, we calculated the uranium element safety mass, which it related to a fifth of this mass to the value 50, which corresponds to criticality safety index (CSI). That is, 20% of the safety mass is the value where the transport is carried out with subcritical mass, going in favor of the security. From the uranium element safety mass (USM) was determined for each enrichment , and it was calculated the mass of 235 U, activity 235 U and dose rate of 235 U, the same items were calculated for the isotope 238 U. The total dose rate was calculated for two isotopes, and applying the transport index definition as the gamma dose rate for the distance of 1 m from the packed, it was determined the TI for 20% of the safety mass for each enrichment of the compound studied. The study of

Thermal restoration of the lattice parameter of weakly irradiated UO{sub 2} (1962); Restauration thermique de parametre du l'UO{sub 2} faiblement irradie (1962)

Energy Technology Data Exchange (ETDEWEB)

Bloch, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1962-07-01

Weak irradiations of uranium dioxide lead to an expansion of the cristal lattice. We have studied the restoration of the lattice parameter after successive isochronous annealings. The process occurs in two main stages, one starting at 200 deg. C and the other at 500 deg. C. (author) [French] De faibles irradiations du bioxyde d'uranium produisent une dilatation de la maille cristalline. Nous avons etudie la restauration du parametre apres des recuits isochrones successifs. Elle se fait en deux stades principaux: l'un a partir de 200 deg. C, l'autre a partir de 500 deg. C. (auteur)

Progress in developing processes for converting 99Mo production from high- to low-enriched uranium--1998

International Nuclear Information System (INIS)

Conner, C.

1998-01-01

During 1998, the emphasis of our activities was focused mainly on target fabrication. Successful conversion requires a reliable irradiation target; the target being developed uses thin foils of uranium metal, which can be removed from the target hardware for dissolution and processing. This paper describes successes in (1) improving our method for heat-treating the uranium foil to produce a random-small grain structure, (2) improving electrodeposition of zinc and nickel fission-fragment barriers onto the foil, and (3) showing that these fission fragment barriers should be stable during transport of the targets following irradiation. A method was also developed for quantitatively electrodepositing uranium and plutonium contaminants in the 99 Mo. Progress was also made in broadening international cooperation in our development activities

On the separation of so-called non-volatile uranium fission products of uranium using the conversion of neutron-irradiated uranium dioxide and graphite

International Nuclear Information System (INIS)

Elhardt, W.

1979-01-01

The investigations are continued in the following work which arose from the concept of separating uranium fission products from uranium. This is achieved in that due to the lattice conversions occurring during the course of solid chemical reactions, fission products can easily pass from the uranium-contained solid to a second solid. The investigations carried out primarily concern the release behaviour of cerium and neodymium in the temperature region of 1200 to 1700 0 C. UO 2 + graphite, both in powder form, are selected as suitable reaction system having the preconditions needed for the lattice conversion for the release effect. The target aimed at from the practical aspect for the improved release of lanthanoids is achieved by an isobar test course - changing temperature from 1200 to 1500 0 C at constant pressure, with a cerium release of 75-80% and a neodynium release of 80-90% (maximum at 1400 0 C). The concepts on the mechanism of the fission product release are related to transport processes in crystal lattices, as well as chemical solid reactions and evaporation processes on the surface of UC 2 grains. (orig./RB) [de

Optical metrology of Ni and NiSi thin films used in the self-aligned silicidation process

International Nuclear Information System (INIS)

Kamineni, V. K.; Bersch, E. J.; Diebold, A. C.; Raymond, M.; Doris, B. B.

2010-01-01

The thickness-dependent optical properties of nickel metal and nickel monosilicide (NiSi) thin films, used for self-aligned silicidation process, were characterized using spectroscopic ellipsometry. The thickness-dependent complex dielectric function of nickel metal films is shown to be correlated with the change in Drude free electron relaxation time. The change in relaxation time can be traced to the change in grain boundary (GB) reflection coefficient and grain size. A resistivity based model was used as the complementary method to the thickness-dependent optical model to trace the change in GB reflection coefficient and grain size. After silicidation, the complex dielectric function of NiSi films exhibit non-Drude behavior due to superimposition of interband absorptions arising at lower frequencies. The Optical models of the complete film stack were refined using x-ray photoelectron spectroscopy, Rutherford backscattered spectroscopy, and x-ray reflectivity (XRR).

Analysis of fuel cycles with natural uranium, Phase I; Analiza gorivnih ciklusa sa prirodnim uranom, I faza

Energy Technology Data Exchange (ETDEWEB)

Stojadinovic, A; Zivkovic, Z; Raisic, N [Institute of Nuclear Sciences Boris Kidric, Laboratorija za fiziku i dinamiku reaktora, Vinca, Beograd (Serbia and Montenegro)

1964-12-15

This paper contains analyses of fuel cycles with natural uranium for the following cases: plutonium recycling is not done; recycling of plutonium and irradiated uranium with the condition of equal multiplication factor at the beginning of each cycle; and recycling of plutonium only.

Determination of burnup in irradiated nuclear fuels by the method of stable 148Nd fission products

International Nuclear Information System (INIS)

Souza Sarkis, J.E. de.

1982-01-01

A method is described for the isotopic analysis and determination of burnup in irradiated nuclear fuel by mass spectrometric technique. The burnup is calculed from the determination of the concentration of uranium, plutonium and the fission product 148 Nd in the samples of UO 2 irradiated fuel from a P.W.R. type reactor. The method involves the separation of fractions of uranium, plutonium and neodimium from the dissolved irradiated fuel by ion exchange technique. The determination of uranium, plutonium and the fission product 148 Nd is carried out by isotope diluition mass spectrometry technique using as isotope tracers the nuclides 233 U, 242 Pu and 150 Nd. For the chemical processing and handling of the irradiated sample a Hot Chemistry Laboratory was mounted. It consists of glove boxes and equipments for radiation monitoring and protection. The results obtained indicate an atom percent burnup of 2.181 + - 0.035% wich corresponds an 20.937 + - 0.739 Gwd/ton of thermal energy produced. (Author) [pt

Separation of heavier rare earths from neutron irradiated uranium targets

International Nuclear Information System (INIS)

Bhargava, V.K.; Rao, V.K.; Marathe, S.G.; Sahakundu, S.M.; Iyer, R.H.

1978-01-01

A radiochemical method is described for the separation of heavier rare earths from the fission of uranium. The method is particularly suitable for the separation of low yield (10sup(-5)%-10sup(-7)%), highly asymmetric rare earth fission products viz. sup(179,177)Lu, sup(175)Yb, sup(173)Tm, sup(172,171)Er, sup(167)Ho and sup(161,160)Tb in the neutron induced fission of natural and depleted uranium targets. Additional separation steps have been incorporated for decontamination from sup(239)Np (an activation product) and sup(93-90)Y (a high fission-yield product) which show similar chemical behaviour to rare earths. Separation of individual rare earths is achieved by a cation exchange method performed at 80 deg C by elution with α-hydroxyisobutyric acid (α-HIBA). (author)

The reprocessing of irradiated fuels improvement and extension of the solvent extraction process; Le traitement des combustibles irradies amelioration et extension du procede utilisant les solvants

Energy Technology Data Exchange (ETDEWEB)

Faugeras, P; Chesne, A [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

1964-07-01

Improvements made in the conventional tri-butylphosphate process are described, in particular. the concentration and the purification of plutonium by one extraction cycle using tri-butyl-phosphate with reflux; and the use of an apparatus working continuously for precipitating plutonium oxalate, for calcining the oxalate, and for fluorinating the oxide. The modifications proposed for the treatment of irradiated uranium - molybdenum alloys are described, in particular, the dissolution of the fuel, and the concentration of the fission product solutions. The solvent extraction treatment is used also for the plutonium fuels utilized for the fast breeder reactor (Rapsodie) An outline of the process is presented and discussed, as well as the first experimental results and the plans for a pilot plant having a capacity of 1 kg/day. The possible use of tn-lauryl-amine in the plutonium purification cycle is now under consideration for the processing plant at La Hague. The flowsheet for this process and its performance are presented. The possibility of vitrification is considered for the final treatment of the concentrated radioactive wastes from the Marcoule (irradiated uranium) and La Hague (irradiated uranium-molybdenum) Centers. Three possible processes are described and discussed, as well as the results obtained from the operation of the corresponding experimental units using tracers. (authors) [French] On decrit les ameliorations apportees au procede classique utilisant le phosphate tributylique, et notamment la concentration et la purification du plutonium par un cycle d'extraction au tributylphosphate avec reflux, l'utilisation d'un appareillage continu de precipitation d'oxalate de plutonium, de calcination de l'oxalate, et de fluoration de l'oxyde. On presente les modifications envisagees pour le traitement des alliages uranium-molybdene irradies, principalement en ce qui concerne la dissolution du combustible et la concentration des solutions de produits de fission

Irradiation behavior of uranium oxide - Aluminum dispersion fuel

International Nuclear Information System (INIS)

Hofman, Gerard L.; Rest, Jeffrey; Snelgrove, James L.

1996-01-01

An oxide version of the DART code has been generated in order to assess the irradiation behavior of UO 2 -Al dispersion fuel. The aluminum-fuel interaction models were developed based on U 3 O 8 -Al irradiation data. Deformation of the fuel element occurs due to fuel particle swelling driven by both solid and gaseous fission products and as a consequence of the interaction between the fuel particles and the aluminum matrix. The calculations show that, with the assumption that the correlations derived from U 3 O 8 are valid for UO 2 , the LEU UO 2 -Al with a 42% fuel volume loading (4 g U/cm 3 ) irradiated at fuel temperatures greater than 413 K should undergo breakaway swelling at core burnups greater than about 1.12 x 10 27 fissions m -3 (∼63% 235 U burnup). (author)

Irradiation behavior of uranium oxide-aluminum dispersion fuel

International Nuclear Information System (INIS)

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

1996-01-01

An oxide version of the DART code has been generated in order to assess the irradiation behavior of UO 2 -Al dispersion fuel. The aluminum-fuel interaction models were developed based on U 3 O 8 -Al irradiation data. Deformation of the fuel element occurs due to fuel particle swelling driven by both solid and gaseous fission products, as well as a consequence of the interaction between the fuel particles and the aluminum matrix. The calculations show, that with the assumption that the correlations derived from U 3 O 8 are valid for UO 2 , the LEU UO 2 -Al with a 42% fuel volume loading (4 gm/cc) irradiated at fuel temperatures greater than 413 K should undergo breakaway swelling at core burnups greater than about 1.12 x 10 27 fissions m -3 (∼ 63% 235 U burnup)

Reduced Enrichment for Research and Test Reactors. Proceedings of the XIV international meeting

Energy Technology Data Exchange (ETDEWEB)

Suripto, Asmedi; Hastowo, Hudi; Hersubeno, J B [eds.

1995-07-01

Apart from the progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program the national programs of Indonesia, Japan and China were presented. The major events, findings, and activities of 1991 are reviewed with a brief summary of the results which the RERTR Program had achieved by the end of 1990 in collaboration with its many international partners. The RERTR program, has concentrated its efforts on technology transfer and implementation activities consistent with the guidance received from the Department of Energy at the end of 1990. A number of presentations were devoted to development of new fuel uranium silicide fuel elements, fuel irradiation testing and reactor core conversions from highly enriched (HEU) to slightly enriched uranium (LEU). Calculations and measurements of converted reactor core parameters were shown related to safety test and analysis. Fuel cycle issue were discussed as well. One should note that a significant number of papers were devoted to Indonesian GA SIWABESSY reactor core conversion and related topics.

Reduced Enrichment for Research and Test Reactors. Proceedings of the XIV international meeting

International Nuclear Information System (INIS)

Suripto, Asmedi; Hastowo, Hudi; Hersubeno, J.B.

1995-01-01

Apart from the progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program the national programs of Indonesia, Japan and China were presented. The major events, findings, and activities of 1991 are reviewed with a brief summary of the results which the RERTR Program had achieved by the end of 1990 in collaboration with its many international partners. The RERTR program, has concentrated its efforts on technology transfer and implementation activities consistent with the guidance received from the Department of Energy at the end of 1990. A number of presentations were devoted to development of new fuel uranium silicide fuel elements, fuel irradiation testing and reactor core conversions from highly enriched (HEU) to slightly enriched uranium (LEU). Calculations and measurements of converted reactor core parameters were shown related to safety test and analysis. Fuel cycle issue were discussed as well. One should note that a significant number of papers were devoted to Indonesian GA SIWABESSY reactor core conversion and related topics

Analysis of burnable poison in Ford Nuclear Reactor fuel to extend fuel lifetime. Final report, August 1, 1994--September 29, 1996

Energy Technology Data Exchange (ETDEWEB)

Burn, R.R.; Lee, J.C.

1996-12-01

The objective of the project was to establish the feasibility of extending the lifetime of fuel elements for the Ford Nuclear Reactor (FNR) by replacing current aluminide fuel with silicide fuel comprising a heavier uranium loading but with the same fissile enrichment of 19.5 wt% {sup 235}U. The project has focused on fuel designs where burnable absorbers, in the form of B{sub 4}C, are admixed with uranium silicide in fuel plates so that increases in the control reactivity requirements and peak power density, due to the heavier fuel loading, may be minimized. The authors have developed equilibrium cycle models simulating current full-size aluminide core configurations with 43 {approximately} 45 fuel elements. Adequacy of the overall equilibrium cycle approach has been verified through comparison with recent FNR experience in spent fuel discharge rates and simulation of reactor physics characteristics for two representative cycles. Fuel cycle studies have been performed to compare equilibrium cycle characteristics of silicide fuel designs, including burnable absorbers, with current aluminide fuel. These equilibrium cycle studies have established the feasibility of doubling the fuel element lifetime, with minimal perturbations to the control reactivity requirements and peak power density, by judicious additions of burnable absorbers to silicide fuel. Further study will be required to investigate a more practical silicide fuel design, which incorporates burnable absorbers in side plates of each fuel element rather than uniformly mixes them in fuel plates.

Uranium content of petroleum by Fission track technique

International Nuclear Information System (INIS)

Paschaa, A.S.; Mafra, O.Y.; Oliveira, C.A.N.; Pinto, L.R.

1982-01-01

This paper examines the feasibility of the fission track registration technique to investigate the natural uranium concentration in petroleum. The application is briefly described and the results obtained indicate the presence of uranium concentrations in samples of Brazilian petroleum which are over the detection limit of the fission track technique. The irradiations were performed by using fluxes with predominance of thermal neutrons, which have a fission cross-section for U 235 equal to 579 barns. Since the neutron fluxes were not comp sed exclusively of thermal neutrons, fissions from fast neutrons would also be taken into account for U 238 and Th 232

Irradiation testing of miniature fuel plates for the RERTR program

Energy Technology Data Exchange (ETDEWEB)

Senn, R L; Martin, M M [Oak Ridge National Laboratory, Oak Ridge, TN 37830 (United States)

1983-08-01

An irradiation test facility, which provides a test bed for irradiating a variety of miniature fuel plates miniplates) for the Reduced Enrichment Research and Test Reactors (RERTR) program, has been placed into operation. The objective of these tests is to screen various candidate fuel materials as to their suitability for replacing the highly enriched uranium fuel materials currently used by the world's test and research reactors with a lower enrichment fuel material, without significantly degrading reactor operating characteristics and power levels. The use of low uranium enrichment of about 20% {sup 235}U in place of highly enriched fuel for these reactors would reduce the potential for {sup 235}U diversion. Fuel materials currently being evaluated in this first phase of these screening tests include aluminum-base dispersion-type fuel plates with fuel cores of 1) high uranium content U{sup 3}){sup 8}-Al being developed by ORNL, 2) high uranium content UAI{sub x}-Al being developed by EG and G Idaho, Inc., and 3) very high uranium content U{sub 3}Si-Al- being developed by ANL. The miniplates are 115-mm long by 50-mm wide with overall plate thicknesses of 1.27 or 1.52 mm. The fuel core dimensions vary according to overall plate thicknesses with a minimal clad thickness requirement of 0.20 mm. Sixty such miniplates (thirty of each thickness) can be irradiated in one test facility. The irradiation test facility, designated as HFED-1 is operating in core position E-7 in the Oak Ridge Research Reactor (ORR), a 30-MW water-moderated reactor. The peak neutron flux measured for this experiment is 1.96 x 10{sup 18} neutrons m{sub -2} s{sub -1}. The various types of miniplates will achieve burnups of up to approximately 2.2x10{sup 27} fissions/m{sup 3} of fuel, which will require approximately eight full power months of irradiation. During reactor shutdown periods, the experiment is removed from the reactor, moved to a special poolside station, disassembled, and inspected

Development of U6Fe-Al dispersions for the use of LEU in research and test reactors

International Nuclear Information System (INIS)

Nazare, S.

1983-01-01

For some time now, efforts are being made to develop fuel dispersions that would permit the use of low (approx. 20% 235-U) enriched uranium (LEU) instead of the currently used highly (approx. 93% 235-U) enriched uranium (HEU) in research and test reactors. Since penalties in the performance of the reactor have to be avoided, the 235-U content in the dispersion has at least to be retained at current levels. On account of their high U-densities, the major development effort has been focussed on the uranium silicides (U 3 Si, U 3 Si(Al), and U 3 Si 2 -based dispersions). With silicides as dispersants, it is possible to fabricate fuel element plates with U-densities in the dispersion of about 6.0 gU/cm 3 . In comparison to the silicides, the U 6 Fe-phase offers several advantages namely: higher U-density (approx. 17.0 gU/cm 3 ); relative ease of formation compared to U 3 Si; possible advantages with regard to reprocessing of the spent fuel due to the absence of silicon. The studies outlined here were performed with a view to investigating the preparation, reaction behavior and dimensional stability after heat treatment of U 6 Fe-Al dispersions

Development of U6Fe-Al dispersions for the use of LEU in research and test reactors

International Nuclear Information System (INIS)

Nazare, S.

1983-01-01

For some time now, efforts are being made to develop fuel dispersions that would permit the use of low (∼ 20% 235-U) enriched uranium (LEU) instead of the currently used highly (∼ 93% 235-U) enriched uranium (HEU) in research and test reactors. Since penalties in the performance of the reactor have to be avoided, the 235-U content in the dispersion has at least to be retained at current levels. On account of their high U-densities, the major development effort has been focussed on the uranium silicides [U 3 Si, U 3 Si(Al), and U 3 Si 2 - based dispersions. With silicides as dispersants, it is possible to fabricate fuel element plates with U-densities in the dispersion of about 6.0 g U/cm 3 . In comparison to the silicides, the U 6 Fe-phase offers several advantages namely: - higher U-density (∼ 17.0 g U/cm 3 ); - relative ease of formation compared to U 3 Si; - possible advantages with regard to reprocessing of the spent fuel due to the absence of silicon. The studies outlined here were therefore performed with a view to investigating the preparation, reaction behaviour and dimensional stability after heat treatment of U 6 Fe-Al dispersions

Analysis of fuel cycles with natural uranium; Analiza gorivnih ciklusa sa prirodnim uranom

Energy Technology Data Exchange (ETDEWEB)

Stojanovic, A [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

1965-05-15

A method was developed and a computer code was written for analysis of fuel cycles and it was applied for heavy water and graphite moderated power reactors. Among a variety of possibilities, three methods which enable best utilization of natural uranium and plutonium production were analyzed. Analysis has shown that reprocessing of irradiated uranium and plutonium utilization in the same or similar type of reactor could increase significantly utilization of natural uranium. Increase of burnup is limited exclusively by costs of reprocessing, plutonium extraction and fabrication of new fuel elements.

Fuel element burnup measurements for the equilibrium LEU silicide RSG GAS (MPR-30) core under a new fuel management strategy

International Nuclear Information System (INIS)

Pinem, Surian; Liem, Peng Hong; Sembiring, Tagor Malem; Surbakti, Tukiran

2016-01-01

Highlights: • Burnup measurement of fuel elements comprising the new equilibrium LEU silicide core of RSG GAS. • The burnup measurement method is based on a linear relationship between reactivity and burnup. • Burnup verification was conducted using an in-house, in-core fuel management code BATAN-FUEL. • A good agreement between the measured and calculated burnup was confirmed. • The new fuel management strategy was confirmed and validated. - Abstract: After the equilibrium LEU silicide core of RSG GAS was achieved, there was a strong need to validate the new fuel management strategy by measuring burnup of fuel elements comprising the core. Since the regulatory body had a great concern on the safety limit of the silicide fuel element burnup, amongst the 35 burnt fuel elements we selected 22 fuel elements with high burnup classes i.e. from 20 to 53% loss of U-235 (declared values) for the present measurements. The burnup measurement method was based on a linear relationship between reactivity and burnup where the measurements were conducted under subcritical conditions using two fission counters of the reactor startup channel. The measurement results were compared with the declared burnup evaluated by an in-house in-core fuel management code, BATAN-FUEL. A good agreement between the measured burnup values and the calculated ones was found within 8% uncertainties. Possible major sources of differences were identified, i.e. large statistical errors (i.e. low fission counters’ count rates), variation of initial U-235 loading per fuel element and accuracy of control rod indicators. The measured burnup of the 22 fuel elements provided the confirmation of the core burnup distribution planned for the equilibrium LEU silicide core under the new fuel management strategy.

Understanding and Improving High-Temperature Structural Properties of Metal-Silicide Intermetallics

Energy Technology Data Exchange (ETDEWEB)

Bruce S. Kang

2005-10-10

The objective of this project was to understand and improve high-temperature structural properties of metal-silicide intermetallic alloys. Through research collaboration between the research team at West Virginia University (WVU) and Dr. J.H. Schneibel at Oak Ridge National Laboratory (ORNL), molybdenum silicide alloys were developed at ORNL and evaluated at WVU through atomistic modeling analyses, thermo-mechanical tests, and metallurgical studies. In this study, molybdenum-based alloys were ductilized by dispersing MgAl2O4 or MgO spinel particles. The addition of spinel particles is hypothesized to getter impurities such as oxygen and nitrogen from the alloy matrix with the result of ductility improvement. The introduction of fine dispersions has also been postulated to improve ductility by acting as a dislocation source or reducing dislocation pile-ups at grain boundaries. The spinel particles, on the other hand, can also act as local notches or crack initiation sites, which is detrimental to the alloy mechanical properties. Optimization of material processing condition is important to develop the desirable molybdenum alloys with sufficient room-temperature ductility. Atomistic analyses were conducted to further understand the mechanism of ductility improvement of the molybdenum alloys and the results showed that trace amount of residual oxygen may be responsible for the brittle behavior of the as-cast Mo alloys. For the alloys studied, uniaxial tensile tests were conducted at different loading rates, and at room and elevated temperatures. Thermal cycling effect on the mechanical properties was also studied. Tensile tests for specimens subjected to either ten or twenty thermal cycles were conducted. For each test, a follow-up detailed fractography and microstructural analysis were carried out. The test results were correlated to the size, density, distribution of the spinel particles and processing time. Thermal expansion tests were carried out using thermo

Alpha Radiolysis of Sorbed Water on Uranium Oxides and Uranium Oxyfluorides

Energy Technology Data Exchange (ETDEWEB)

Icenhour, A.S.

2003-09-10

The radiolysis of sorbed water and other impurities contained in actinide oxides has been the focus of a number of studies related to the establishment of criteria for the safe storage and transport of these materials. Gamma radiolysis studies have previously been performed on uranium oxides and oxyfluorides (UO{sub 3}, U{sub 3}O{sub 8}, and UO{sub 2}F{sub 2}) to evaluate the long-term storage characteristics of {sup 233}U. This report describes a similar study for alpha radiolysis. Uranium oxides and oxyfluorides (with {sup 238}U as the surrogate for {sup 233}U) were subjected to relatively high alpha radiation doses (235 to 634 MGy) by doping with {sup 244}Cm. The typical irradiation time for these samples was about 1.5 years, which would be equivalent to more than 50 years irradiation by a {sup 233}U sample. Both dry and wet (up to 10 wt % water) samples were examined in an effort to identify the gas pressure and composition changes that occurred as a result of radiolysis. This study shows that several competing reactions occur during radiolysis, with the net effect that only very low pressures of hydrogen, nitrogen, and carbon dioxide are generated from the water, nitrate, and carbon impurities, respectively, associated with the oxides. In the absence of nitrate impurities, no pressures greater than 1000 torr are generated. Usually, however, the oxygen in the air atmosphere over the oxides is consumed with the corresponding oxidation of the uranium oxide. In the presence of up to 10 wt % water, the oxides first show a small pressure rise followed by a net decrease due to the oxygen consumption and the attainment of a steady-state pressure where the rate of generation of gaseous components is balanced by their recombination and/or consumption in the oxide phase. These results clearly demonstrate that alpha radiolysis of either wet or dry {sup 233}U oxides will not produce deleterious pressures or gaseous components that could compromise the long-term storage of

Status of development and irradiation performance of advanced proliferation resistant MTR fuel at NUKEM

International Nuclear Information System (INIS)

Hrovat, M.; Hassel, H.-W.; Wehner, E.

1985-01-01

This paper describes the current status of development and irradiation performance of fuel elements for Material Test and Research (MTR) Reactors with Medium Enriched Uranium (MEU, ≤ 45 % 235-U) and Low Enriched Uranium (LEU, ≤ 20 % 235-U). (author)

RERTR-12 Insertion 1 Irradiation Summary Report

International Nuclear Information System (INIS)

Perez, D.M.; Lillo, M.A.; Chang, G.S.; Woolstenhulme, N.E.; Roth, G.A.; Wachs, D.M.

2012-01-01

The Reduced Enrichment for Research and Test Reactor (RERTR) experiment RERTR-12 was designed to provide comprehensive information on the performance of uranium-molybdenum (U-Mo) based monolithic fuels for research reactor applications. RERTR-12 insertion 1 includes the capsules irradiated during the first two irradiation cycles. These capsules include Z, X1, X2 and X3 capsules. The following report summarizes the life of the RERTR-12 insertion 1 experiment through end of irradiation, including as-run neutronic analysis results, thermal analysis results and hydraulic testing results.

RERTR-12 Insertion 2 Irradiation Summary Report

International Nuclear Information System (INIS)

Perez, D.M.; Chang, G.S.; Wachs, D.M.; Roth, G.A.; Woolstenhulme, N.E.

2012-01-01

The Reduced Enrichment for Research and Test Reactor (RERTR) experiment RERTR-12 was designed to provide comprehensive information on the performance of uranium-molybdenum (U-Mo) based monolithic fuels for research reactor applications.1 RERTR-12 insertion 2 includes the capsules irradiated during the last three irradiation cycles. These capsules include Z, Y1, Y2 and Y3 type capsules. The following report summarizes the life of the RERTR-12 insertion 2 experiment through end of irradiation, including as-run neutronic analysis results, thermal analysis results and hydraulic testing results.