The experimental program of neutronphysics for advanced water reactors

International Nuclear Information System (INIS)

Martin-Deider, L.; Cathalu, S.; Santamarina, A.; Gomit, M.

1985-11-01

The C.E.A. and E.D.F. has jointly undertaken a program of experimental studies on under-moderated water lattices, with mixed oxide fuel UO 2 -PuO 2 . Undermoderated lattices offer high conversion ratios. This type of lattice could limit in the future the natural uranium consumption of pressurized water reactors. This experimental program is aimed at qualifying neutron transport calculations in a large range of moderating ratio (between 0.5 and 1.5). It includes three experiments: ERASME, a critical experiment of large size in the EOLE reactor at Cadarache; ICARE, an irradiation experiment in the MELUSINE reactor at Grenoble; and an experiment to measure the reactivity effects by oscillations in the MINERVE reactor at Cadarache [fr

Measurements of kinetic parameters by noise techniques on the MINERVE reactor

International Nuclear Information System (INIS)

Carre, J.C.; Da Costa Oliveira, J.

1975-01-01

Noise measurements were determined on ERMINE a fast thermal coupled reactor built in MINERVE. A reactor without feedback, and a reactor with an automatic control rod were both considered. The first case concerned the measurements of auto and cross power spectral density obtained with one or two neutron detectors, and the determination of: neutron lifetime; efficiency for one ion chamber; power level of the reactor; maximal speed and acceleration of the control rod for the design of an automatic reactor control actuator. The second case was concerned with measurements of the auto power spectral density in reactivity for the control rod, and the estimation of: the transfer function of the automatic pilot; the neutron lifetime; and the standard error affecting the results obtained by the oscillation method. The results proved that the pile noise theory with a point kinetic model is sufficient for application on zero power reactors. (U.K.)

Fission product release from UO2 during irradiation. Diffusion data and their application to reactor fuel pins

International Nuclear Information System (INIS)

Findlay, J.R.; Johnson, F.A.; Turnbull, J.A.; Friskney, C.A.

1980-01-01

Release of fission product species from UO 2 , and to a limited extent from (U, Pu)0 2 was studied using small scale in-reactor experiments in which these interacting variables may be separated, as far as is possible, and their influences assessed. Experiments were at fuel ratings appropriate to water reactor fuel elements and both single crystal and poly-crystalline specimens were used. They employed highly enriched uranium such that the relative number of fissions occurring in plutonium formed by neutron capture was small. The surface to volume ratio (S/V) of the specimens was well defined thus reducing the uncertainties in the derivation of diffusion coefficients. These experiments demonstrate many of the important characteristics of fission product behaviour in UO 2 during irradiation. The samples used for these experiments were small being always less than 1g with a fissile content usually between 2 and 5mg. Polycrystalline materials were taken from batches of production fuel prepared by conventional pressing and sintering techniques. The enriched single crystals were grown from a melt of sodium and potassium chloride doped with UO 2 powder 20% 235 U content. The irradiations were performed in the DIDO reactor at Harwell. The neutron flux at the specimen was 4x10 16 neutrons m -2 s -1 providing a heat rating within the samples of 34.5 MW/teU

UO{sub 2} and PuO{sub 2} utilization in high temperature engineering test reactor with helium coolant

Energy Technology Data Exchange (ETDEWEB)

Waris, Abdul, E-mail: awaris@fi.itb.ac.id; Novitrian,; Pramuditya, Syeilendra; Su’ud, Zaki [Nuclear Physics and Biophysics Research Division, Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung (Indonesia); Aji, Indarta K. [Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung (Indonesia)

2016-03-11

High temperature engineering test reactor (HTTR) is one of high temperature gas cooled reactor (HTGR) types which has been developed by Japanese Atomic Energy Research Institute (JAERI). The HTTR is a graphite moderator, helium gas coolant, 30 MW thermal output and 950 °C outlet coolant temperature for high temperature test operation. Original HTTR uses UO{sub 2} fuel. In this study, we have evaluated the use of UO{sub 2} and PuO{sub 2} in form of mixed oxide (MOX) fuel in HTTR. The reactor cell calculation was performed by using SRAC 2002 code, with nuclear data library was derived from JENDL3.2. The result shows that HTTR can obtain its criticality condition if the enrichment of {sup 235}U in loaded fuel is 18.0% or above.

Versatility of {l_brace}M(30-crown-10){r_brace} (M = K{sup +}, Ba{sup 2+}) as a guest in UO{sub 2}{sup 2+} complexes of 3.1.3.1 - and 3.3.3 homo-oxa-calixarenes

Energy Technology Data Exchange (ETDEWEB)

Masci, B. [Univ Roma La Sapienza, Dipartimento Chim, I-00185 Rome, (Italy); Thuery, P. [CEA Saclay, DSM/DRECAM/SCM, CNRS-URA 331, F-91191 Gif Sur Yvette, (France)

2007-07-01

The reaction between p-R-[3.1.3.1]- or [3.3.3] homo-oxa-calixarenes and uranyl salts in the presence of 30-crown-10 and the alkali or alkaline-earth metal cations K{sup +} or Ba{sup 2+} gives various supramolecular assemblages characterized by 'complex-within-complex' architectures. These can be of the simple nesting or sandwich types, as in [{l_brace}Ba(30-crown-10){r_brace}{l_brace}UO{sub 2}(L{sup 1}){r_brace}]. 2H{sub 2}O.3CHCl{sub 3} (L{sup 1}H{sub 4} p-tert-butyl[3.1.3.1] homo-oxa-calixarene) and [{l_brace}Ba(30-crown-10){r_brace}{l_brace}UO{sub 2}(L{sup 4}){r_brace}{sub 2}].2CHCl{sub 3} (L{sup 4}H{sub 3} p-bromo[3.3.3]homo-oxa-calixarene), respectively, with the cation held in the cavity of the homo-oxa-calixarene complexes in cone conformation by weak interactions, but more original structures arise when uranyl-cation bonds are present. In [{l_brace}Ba(30-crown-10){r_brace}{l_brace}UO{sub 2}(L{sup 2}){r_brace}] (L{sup 2}H{sub 4} p-phenyl[3.1.3.1] homo-oxa-calixarene), the barium ion included in the crown ether is bound to the uranyl oxo group located out of the calixarene cavity, resulting in the formation of a neutral species which self-organizes to form a columnar assembly by auto-inclusion. In [{l_brace}K(30-crown-10){r_brace}{l_brace}UO{sub 2}K(L{sup 1})(H{sub 2}O){sub 3}{r_brace}]{sub 2}.6H{sub 2}O, the nesting-type subunit dimerizes around two oxo-bound potassium ions. Finally, the use of the coordinating solvent dimethylsulfoxide leads to the neutral complex [UO{sub 2}Ba(L{sup 3})(dmso){sub 2}(MeOH)]{sub 2} (L{sup 3}H{sub 4} = p-methyl[3.1.3.1] homo-oxa-calixarene), in which the crown ether is absent and two oxo-, phenoxo- and ether-bound barium atoms ensure the dimerization of the uranyl complex. (authors)

Optimization of UO{sub 2} Granule Characteristics for UO{sub 2}-Mo Pellet Fabrication

Energy Technology Data Exchange (ETDEWEB)

Kim, Dongjoo; Rhee, Young Woo; Kim, Jong Hun; Kim, Keon Sik; Oh, Jang Soo; Yang, Jae Ho; Koo, Yanghyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-05-15

The in-reactor performance, integrity, safety and accident tolerance of the nuclear fuel can be significantly affected by the thermal conductivity of the UO{sub 2} fuel pellet. The improvement in the thermal conductivity of the UO{sub 2} fuel pellet can enhance the fuel performance in various ways. Typically, the FGR (Fission Gas Release) can be reduced by the application of a large-grain fuel pellet because the moving path of the fission gas to the grain boundary is much longer. In addition, the mobility of the fission gases is reduced by the lower temperature gradient in the UO{sub 2} fuel pellet. That is to say, the capacity of the fission gas retention of the fuel pellet can increase. In addition, the lower centerline temperature of the fuel pellet affects the accident tolerance for nuclear fuel as well as the enhancement of fuel safety and fuel pellet integrity under normal operation conditions. In addition, the nuclear reactor power can be uprated owing to the higher safety margin. Thus, many researches on enhancing the thermal conductivity of a nuclear fuel pellet for LWRs have been performed in various ways. From the viewpoint of the development of fuel pellet fabrication technology, an enhancement of the thermal conductivity of a pellet can be obtained by the addition of a higher thermal conductive material in the UO{sub 2} pellet. It is known that a UO{sub 2}-metal composite pellet is one of the most effective concepts. However, to maximize the effect of the metallic phase for thermal conductivity enhancement, a continuous channel of the metallic phase in the UO{sub 2} matrix must be formed. Additionally, if the fabrication process of a UO{sub 2}-metal composite pellet is compatible with a conventional sintering process, the developed technology will be favorable. To enhance the thermal conductivity of a UO{sub 2} pellet, there are the various methods for an appropriate arrangement of the high thermal conductive material in a UO{sub 2} matrix. In this

Effects of UO2 fuel microstructure and density on fuel in-reactor performance

International Nuclear Information System (INIS)

Hansson, L.

1988-02-01

The volume changes of UO 2 fuel pellets, produced by neutron irradiation, can be characterized by two processes: fission spike induced densification through pore skrinkage and later fission produced induced swelling of UO 2 matrix. In-pile densification is controlled by the initial density and microstructure of the fuel, particularly by the pore size distribution. The extent of swelling depends mainly on the amount of fission products produced, but the fission gas release as well as the swelling may be reduced by increasing the grain size of UO 2 . Fabrication of fuel pellets having certain in-reactor properties requires detailed knowledge of the effects of individual fabrication parameters. The irradiation experience of fuels fabricated by using different conversion and pelletizing methods is extensive. Based on this experience, some general characteristics of stable/well-performing fuel microstructures have been summarized

Contribution of the study of a nuclear reactor accident: residual power aspects and thermodynamic of U-UO_2 and UO_2-ZrO_2 systems

International Nuclear Information System (INIS)

Baichi, Mehdi

2001-01-01

This work is a contribution to the study of early delocalization and fission product releases during the formation of corium coming from a nuclear reactor accident. The first part deals with an analysis of corium cooling. The contribution to the power of each corium element has been calculated with time. The main elements are represented but the addition of Pu, Mo and Nb has been proposed. The last release experimental data taken into account result in a loss of residual power of 25% exclusive of corium between the emergency stop and ten days. The second part deals with the early delocalization observed during Vercors experiments. A critical selection on the U-UO_2 and UO_2-ZrO_2 systems has been carried out. In order to complete the small and inconsistent data, thermodynamic activity measurements have been performed by mass spectrometry. The UO_2 activity on UO_2-ZrO_2 presents a positive deviation from ideality at 2200 K and approximates ideality at 2400 K. All the data have been used for optimizing the systems with Thermo-Calc. This work has allowed to calculate the ternary systems and to define the required approach to analyze the metallic phase and corium oxides densities. (author) [fr

Fully coupled multiphysics modeling of enhanced thermal conductivity UO{sub 2}–BeO fuel performance in a light water reactor

Energy Technology Data Exchange (ETDEWEB)

Liu, R. [Department of Mechanical and Biomedical Engineering, City University of Hong Kong, Hong Kong (China); Zhou, W., E-mail: wenzzhou@cityu.edu.hk [Department of Mechanical and Biomedical Engineering, City University of Hong Kong, Hong Kong (China); Shen, P. [Department of Mechanical and Biomedical Engineering, City University of Hong Kong, Hong Kong (China); Prudil, A. [Fuel and Fuel Channel Safety Branch, Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Chan, P.K. [Department of Chemistry and Chemical Engineering, Royal Military College of Canada, Kingston, Ontario (Canada)

2015-12-15

Highlights: • LWR fuel performance modeling capability developed. • Fully coupled multiphysics studies for enhanced thermal conductivity UO{sub 2}–BeO fuel. • UO{sub 2}–BeO fuel decreases fuel temperature and lessens thermal stresses. • UO{sub 2}–BeO fuel facilitates a reduction in PCMI. • Reactor safety can be improved for UO{sub 2}–BeO fuel. - Abstract: Commercial light water reactor fuel UO{sub 2} has a low thermal conductivity that leads to the development of a large temperature gradient across the fuel pellet, limiting the reactor operational performance due to the effects that include thermal stresses causing pellet cladding interaction and the release of fission product gases. This study presents the development of a modeling and simulation for enhanced thermal conductivity UO{sub 2}–BeO fuel behavior in a light water reactor, using self-defined multiple physics models fully coupled based on the framework of COMSOL Multiphysics. Almost all the related physical models are considered, including heat generation and conduction, species diffusion, thermomechanics (thermal expansion, elastic strain, densification, and fission product swelling strain), grain growth, fission gas production and release, gap heat transfer, mechanical contact, gap/plenum pressure with plenum volume, cladding thermal and irradiation creep and oxidation. All the phenomenal models and materials properties are implemented into COMSOL Multiphysics finite-element platform with a 2D axisymmetric geometry of a fuel pellet and cladding. UO{sub 2}–BeO enhanced thermal conductivity nuclear fuel would decrease fuel temperatures and facilitate a reduction in pellet cladding interaction from our simulation results through lessening thermal stresses that result in fuel cracking, relocation, and swelling, so that the safety of the reactor would be improved.

Spectral shift controlled reactor, UO2 once-through cycle optimized

International Nuclear Information System (INIS)

1978-05-01

This paper presents technical and economic data on the SSCR which may be of use in the International Fuel Cycle Evaluation Program to intercompare alternative nuclear systems. Included in this data is information on the optimized UO 2 once-through fuel cycle. The ''optimized'' cycle refers to a UO 2 once-through cycle which has better fuel resource utilization than the conventional UO 2 cycle employed in current design PWRs. This fuel cycle uses more in-core batches and a higher discharge exposure than current PWR fuel management schemes. The proposed cycle is not optimal in a mathematical sense, however, since additional resource savings can be obtained if the discharge exposure is extended to even higher values and the number of in-core fuel batches is increased further. The present cycle was selected as ''optimal'' based on the assumption that it can be achieved with only an extension of fuel design technology and can therefore be deployed in a relatively short time frame. In the longer term, modification to reactor geometry as well as further extensions of discharge burnup might be considered to realize additional reduction in uranium resource requirements. The data contained in this paper has been developed by an ongoing program which at the present time is only 50% complete. The data presented here should therefore be considered preliminary and will be updated in the future as required

Microspheres of UO2, ThO2 and PuO2 for the high temperature reactor

International Nuclear Information System (INIS)

Brandau, E.

2002-01-01

The production of high temperature reactor fuel, so called pebble fuel, was done in the eighties by a special vibrational dropping process to obtain as sintered UO 2 - or ThO 2 -microspheres, so called 'Kernels', with a diameter size of about 300 μm. These microspheres have been coated and embedded in carbon balls to get the pebble fuel. Since the early nineties BRACE is developing the processings of microspheres starting with sols and suspensions to produce Al 2 O 3 , ZrO 2 , HfO 2 and Actinide oxide microspheres. Two main developments have been made: 1) the preparation of the feed solution (sol, suspension) and the solidification processing, and 2) the equipment, design, and electronic control have been completely changed. A newly developed suspension process for actinide oxides and for metal oxides e.g. Al 2 O 3 , TiO 2 , SiO 2 , ZrO 2 , HfO 2 , CeO 2 , ThO 2 , UO 2 , PuO 2 leads to cheaper production of as sintered microspheres. The processing and the installations will be described and the experience of production will be shown. (author)

A detailed study of the dehydration process in synthetic strelkinite, Na[(UO2)(VO4)] . nH2O (n = 0, 1, 2)

International Nuclear Information System (INIS)

Suleimanov, Evgeny V.; Somov, Nikolay V.; Chuprunov, Evgeny V.; Mayatskikh, Ekaterina F.; Depmeier, Wulf

2012-01-01

Synthetic strelkinite Na[(UO 2 )(VO 4 )] . nH 2 O (n = 0, 1, 2) was systematically investigated by single crystal X-ray diffraction and thermoanalytical methods. The anhydrous form and two hydrates were isolated as single crystals and the structures of these phases solved: Na[(UO 2 )(VO 4 )], monoclinic, P2 1 /c, a = 6.0205(1) Aa, b = 8.3365(1) Aa, c = 10.4164(2) Aa, β = 100.466(2) , V = 514.10(1) Aa 3 , R 1 = 0.0337; Na[(UO 2 )(VO 4 )] . H 2 O, monoclinic, P2 1 /c, a = 7.722(2) Aa, b = 8.512(1) Aa, c = 10.480(4) Aa, β = 113.18(3) , V = 633.3(3) Aa 3 , R 1 = 0.1658; Na[(UO 2 )(VO 4 )] . 2 H 2 O, monoclinic, P2 1 /n, a = 16.2399(5) Aa, b = 8.2844(2) Aa, c = 10.5011(2) Aa, β = 97.644(2) , V = 1400.24(6) Aa 3 , R 1 = 0.0776. A possible mechanism of the structural transformation processes during dehydration is proposed based on the structures of the anhydrous phase and the hydrates. (orig.)

Testing of HTR UO{sub 2} TRISO fuels in AVR and in material test reactors

Energy Technology Data Exchange (ETDEWEB)

Kania, Michael J., E-mail: MichaelJKania@googlemail.com [Retired from Lockheed Martin Corp, 20 Beach Road, Averill Park, NY 12018 (United States); Nabielek, Heinz, E-mail: heinznabielek@me.com [Retired from Research Center Jülich, Monschauerstrasse 61, 52355 Düren (Germany); Verfondern, Karl [Research Center Juelich,Research Center Jülich, Institute of Energy and Climate Research, 52425 Jülich (Germany); Allelein, Hans-Josef [Research Center Juelich,Research Center Jülich, Institute of Energy and Climate Research, 52425 Jülich (Germany); RWTH Aachen, 52072 Aachen (Germany)

2013-10-15

The German High Temperature Reactor Fuel Development Program successfully developed, licensed and manufactured many thousands of spherical fuel elements that were used to power the experimental AVR reactor and the commercial THTR reactor. In the 1970s, this program extended the performance envelope of HTR fuels by developing and qualifying the TRISO-coated particle system. Irradiation testing in real-time AVR tests and accelerated MTR tests demonstrated the superior manufacturing process of this fuel and its irradiation performance. In the 1980s, another program direction change was made to a low enriched UO{sub 2} TRISO-coated particle system coupled with high-quality manufacturing specifications designed to meet new HTR plant design needs. These needs included requirements for inherent safety under normal operation and accident conditions. Again, the German fuel development program met and exceeded these challenges by manufacturing and qualifying the low-enriched UO{sub 2} TRISO-fuel system for HTR systems with steam generation, gas-turbine systems and very high temperature process heat applications. Fuel elements were manufactured in production scale facilities that contained near defect free UO{sub 2} TRISO coated particles, homogeneously distributed within a graphite matrix with very low levels of uranium contamination. Good irradiation performance for these elements was demonstrated under normal operating conditions to 12% FIMA and under accident conditions not exceeding 1600 °C.

Thermal expansion of ThO2-2 wt% UO2 by HT-XRD

International Nuclear Information System (INIS)

Tyagi, A.K.; Mathews, M.D.

2000-01-01

The linear thermal expansion of polycrystalline ThO 2 -2 wt% UO 2 has been investigated from room temperature to 1473 K in flowing helium atmosphere using high temperature X-ray diffractometry. ThO 2 -2 wt% UO 2 shows a marginally higher linear thermal expansion as compared to pure ThO 2 . The average linear and volume thermal expansion coefficients of ThO 2 -2 wt% UO 2 are found to be α-bar a =9.74x10 -6 K -1 and α-bar v =29.52x10 -6 K -1 (298-1473 K). This study will be useful in designing the nuclear reactor fuel assembly based on ThO 2

BURNY-SQUID, 2-D Burnup of UO2 and Mix UO2 PuO2 Fuel in X-Y or R-Z Geometry

International Nuclear Information System (INIS)

Rosa, I.; Zara, G.; Guidotti, R.

1974-01-01

1 - Nature of physical problem solved: - Multigroup neutron diffusion and burnup equations for two- to five- energy groups over a rectangular region of the x-y or r-z plane. - For a given geometry and initial enrichment, it calculates the two- to five- group flux distributions, the nuclides burnt in a time step t, and then the flux distribution again. This process is repeated until the maximum burn-up is reached. - Criticality search by uniform variation of a control isotope. - Solution of problems with fuel having different geometrical parameters, by means of super-compositions. - Recycle and restart options are available. - UO 2 and PUO 2 -UO 2 fuel can be handled. 2 - Method of solution: The zero-dimension burn-up program RIBOT-5 is coupled with the two-dimension program SQUID and alternately executed. The differential equations are solved by the difference method. 3 - Restrictions on the complexity of the problem: 200 maximum number of compositions 10,000 maximum number of mesh points 5 maximum Number of groups. 4 maximum number of super-compositions. Diagonal symmetry allowed

Measurement of the friction coefficient between UO2 and cladding tube

International Nuclear Information System (INIS)

Tachibana, Toshimichi; Narita, Daisuke; Kaneko, Hiromitsu; Honda, Yutaka

1978-01-01

Most of fuel rods used for light water reactors or fast reactors consist of the cladding tubes filled with UO 2 -PuO 2 pellets. The measurement was made on the coefficient of static friction and the coefficient of dynamic friction in helium under high contact load on UO 2 /Zry-2 and UO 2 /SUS 316 combined samples at the temperature ranging from room temperature to 400 deg. C and from room temperature to 600 deg. C, respectively. The coefficient of static friction for Zry-2 tube and UO 2 pellets was 0.32 +- 0.08 at room temperature and 0.47 +- 0.07 at 400 deg. C, and increased with temperature rise in this temperature range. The coefficient of static friction between 316 stainless steel tube and UO 2 pellets was 0.29 +- 0.04 at room temperature and 1.2 +- 0.2 at 600 deg. C, and increased with temperature rise in this temperature range. The coefficient of dynamic friction for both UO 2 /Zry-2 and UO 2 /SUS 316 combinations seems to be equal to or about 10% excess of the coefficient of static friction. The coefficient of static friction for UO 2 /SUS 316 combination decreased with the increasing number of repetition, when repeating slip several times on the same contact surfaces. (Kobatake, H.)

Radiation damage of UO{sub 2} fuel; Radijaciono ostecenje UO{sub 2} goriva

Energy Technology Data Exchange (ETDEWEB)

Stevanovic, M; Sigulinski, F [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

1966-11-15

Radiation damage study of fuel and fuel elements covers: study of radiation damage methods in Sweden; analysis of testing the fuel and fuel elements at the RA reactor; feasibility study of irradiation in the Institute compared to irradiation abroad in respect to the reactor possibilities. Tasks included in this study are relater to testing of irradiated UO{sub 2} and ceramic fuel elements.

New UO2 fuel studies

International Nuclear Information System (INIS)

Dehaudt, P.; Lemaignan, C.; Caillot, L.; Mocellin, A.; Eminet, G.

1998-01-01

With improved UO 2 fuels, compared with the current PWR, one would enable to: retain the fission products, rise higher burn-ups and deliver the designed power in reactor for longer times, limit the pellet cladding interaction effects by easier deformation at high temperatures. Specific studies are made in each field to understand the basic mechanisms responsible for these improvements. Four programs on new UO 2 fuels are underway in the laboratory: advanced microstructure fuels (doped fuels), fuels containing Er 2 O 3 a burnable absorber, fuels with improved caesium retention, composite fuels. The advanced microstructure UO 2 fuels have special features such as: high grain sizes to lengthen the fission gas diffusion paths, intragranular precipitates as fission gas atoms pinning sites, intergranular silica based viscoplastic phases to improve the creep properties. The grain size growth can be obtained with a long time annealing or with corundum type oxide additives partly soluble in the UO 2 lattice. The amount of doping element compared with its solubility limit and the sintering conditions allows to obtain oxide or metallic precipitates. The fuels containing Er 2 O 3 as a burnable absorber are under irradiation in the TANOX device at the present time. Specific sintering conditions are required to improve the erbium solubility in UO 2 and to reach standard or large grain sizes. The improved caesium retention fuels are doped with SiO 2 +A1 2 O 3 or SiO 2 +ZrO 2 additives which may form stable compounds with the Cs element in accidental conditions. The composite fuels are made of UO 2 particles of about 100 μm in size dispersed in a molybdenum metallic (CERMET) or MgA1 2 O 4 ceramic (CERCER) matrix. The CERMET has a considerably higher thermal conductivity and remains ''cold'' during irradiation. The concept of double barrier (matrix+fuel) against fission products is verified for the CERMET fuel. A thermal analysis of all the irradiated rods shows that the thermal

Synthesis and sintering of UN-UO{sub 2} fuel composites

Energy Technology Data Exchange (ETDEWEB)

Jaques, Brian J., E-mail: BrianJaques@BoiseState.edu [Department of Materials Science and Engineering, Boise State University, 1910 University Dr., Boise, ID 83725 (United States); Center for Advanced Energy Studies, 995 University Blvd., Idaho Falls, ID 83401 (United States); Watkins, Jennifer; Croteau, Joseph R.; Alanko, Gordon A. [Department of Materials Science and Engineering, Boise State University, 1910 University Dr., Boise, ID 83725 (United States); Center for Advanced Energy Studies, 995 University Blvd., Idaho Falls, ID 83401 (United States); Tyburska-Püschel, Beata [Department of Engineering Physics, University of Wisconsin–Madison, 1500 Engineering Dr., Madison, WI 53706 (United States); Meyer, Mitch [Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Xu, Peng; Lahoda, Edward J. [Westinghouse Electric Company LLC, Pittsburgh, PA 15235 (United States); Butt, Darryl P., E-mail: DarrylButt@BoiseState.edu [Department of Materials Science and Engineering, Boise State University, 1910 University Dr., Boise, ID 83725 (United States); Center for Advanced Energy Studies, 995 University Blvd., Idaho Falls, ID 83401 (United States)

2015-11-15

The design and development of an economical, accident tolerant fuel (ATF) for use in the current light water reactor (LWR) fleet is highly desirable for the future of nuclear power. Uranium mononitride has been identified as an alternative fuel with higher uranium density and thermal conductivity when compared to the benchmark, UO{sub 2}, which could also provide significant economic benefits. However, UN by itself reacts with water at reactor operating temperatures. In order to reduce its reactivity, the addition of UO{sub 2} to UN has been suggested. In order to avoid carbon impurities, UN was synthesized from elemental uranium using a hydride-dehydride-nitride thermal synthesis route prior to mixing with up to 10 wt% UO{sub 2} in a planetary ball mill. UN and UN – UO{sub 2} composite pellets were sintered in Ar – (0–1 at%) N{sub 2} to study the effects of nitrogen concentration on the evolved phases and microstructure. UN and UN-UO{sub 2} composite pellets were also sintered in Ar – 100 ppm N{sub 2} to assess the effects of temperature (1700–2000 °C) on the final grain morphology and phase concentration.

Densification Behavior of BN-added UO2

International Nuclear Information System (INIS)

Rhee, Young Woo; Kim, Keonsik; Kim, Dong Joo; Kim, Jong Hun; Oh, Jang Soo; Yang, Jae Ho

2013-01-01

Local wall thinning in pipelines affects the structural integrity of industries like nuclear power plants (NPPs). In the present study a pulsed eddy current (PEC) technology to detect the wall thing of carbon steel pipe covered with insulation is developed. Boron is commercially used as a neutron absorber fuel. A neutron absorber fuel is burned out or depleted during reactor operation. Westinghouse have been produced the Integral Fuel Burnable Absorber (IFBA) which is enriched UO 2 fuel pellets with a thin coating of zirconium diboride (ZrB 2 ) on the outer surface. Standard sintered fuel pellets are sputter coated with ZrB 2 . It is known that IFBA fuel can incur 20% to 30% additional fabrication costs. Boron-dispersed UO 2 fuel pellet made by the conventional pressing and sintering process of a powder mixture of UO 2 and B compound might be more cost-effective than IFBAs. M. G. Andrew et al. tried to sinter boron-dispersed UO 2 green pellet. However, they reported that boron-dispersed UO 2 fuel pellet is very difficult to be fabricated with a sufficient level of boron retention and high sintered density (greater than 90 % of theoretical density) because of the volatilization of boron oxide. We have investigated the densification behavior of mixtures of UO 2 and various boron compounds, such as B 4 C, BN, TiB 2 , ZrB 2 , SiB 6 , and HfB 2 . Boron compounds seemed to act as a sintering additive for UO 2 at a certain low temperature range. In this study, the densification behavior of BN-added UO 2 pellet has been investigated by sintering green pellets of a mixture of UO 2 powder and BN powder in H 2 atmosphere. A high density BN-added UO 2 pellet can be fabricated after sintering at 1200 .deg. C for more than 1 h in a H 2 atmosphere. The sintered density of BN-added UO 2 pellet can be increased up to about 95 %TD

A study of UO2 wafer fuel for very high-power research reactors

International Nuclear Information System (INIS)

Hsieh, T.C.; Jankus, V.Z.; Rest, J.; Billone, M.C.

1983-01-01

The Reduced Enrichment Research and Test Reactor Program is aimed at reducing fuel enrichment to 2 caramel fuel is one of the most promising new types of reduced-enrichment fuel for use in research reactors with very high power density. Parametric studies have been carried out to determine the maximum specific power attainable without significant fission-gas release for UO 2 wafers ranging from 0.75 to 1.50 mm in thickness. The results indicate that (1) all the fuel designs considered in this study are predicted not to fail under full power operation up to a burnup, of 1.9x10 21 fis/cm 3 ; (2) for all fuel designs, failure is predicted at approximately the same fuel centerline temperature for a given burnup; (3) the thinner the wafer, the wider the margin for fuel specific power between normal operation and increased-power operation leading to fuel failure; (4) increasing the coolant pressure in the reactor core could improve fuel performance by maintaining the fuel at a higher power level without failure for a given burnup; and (5) for a given power level, fuel failure will occur earlier at a higher cladding surface temperature and/or under power-cycling conditions. (author)

Performance of highly rated UO₂ fuel in the WR-1 organic-cooled reactor

Energy Technology Data Exchange (ETDEWEB)

Schankula, M. H.; Hastings, I. J.

1977-07-15

Information on oxide fuel behaviour in organic coolant was required as part of the organic-cooled power reactor (OCR) study. Of major interest were data on the release of fission gases from fuel operating at high fuel surface temperatures and low external restraint; features which are peculiar to the OCR. To provide these and other data, UO₂ fuel with cold-worked Zr-2.5wt%Nb sheathing was irradiated in the WR-1 organic-cooled reactor to burnups of 135-154 MWh/kgU at a time-averaged linear power of 60-63 kW/m. Elements with 0.38 and 0.69 mm thick sheathing showed maximum diametral increases averaging 3.7 and 1.7% respectively at pellet mid-planes. Reduced fuel/sheath heat transfer resulting from a difference between internal gas pressure and coolant pressure produced high operating temperatures, and there was evidence of central melting in some elements. Fission gas releases were 30-60%. In the heat affected zone adjacent to brazed appendages, the diametral increases were lower, averaging 0.9 and 0.5% for 0.38 and 0.69 mm thick sheathing respectively. Heat treatment during the brazing process produced a local improvement in sheath creep strength. Highly rated oxide fuel irradiated in organic coolant will require sheathing with improved high temperature creep properties; heat-treated Zr-2.5 wt% Nb may provide this improvement.

BURNY-SQUID, 2-D Burnup of UO{sub 2} and Mix UO{sub 2} PuO{sub 2} Fuel in X-Y or R-Z Geometry

Energy Technology Data Exchange (ETDEWEB)

Rosa, I; Zara, G; Guidotti, R [ENEL-DCO, Via G.B. Martini, 3, 00198 Rome (Italy)

1974-08-01

1 - Nature of physical problem solved: - Multigroup neutron diffusion and burnup equations for two- to five- energy groups over a rectangular region of the x-y or r-z plane. - For a given geometry and initial enrichment, it calculates the two- to five- group flux distributions, the nuclides burnt in a time step t, and then the flux distribution again. This process is repeated until the maximum burn-up is reached. - Criticality search by uniform variation of a control isotope. - Solution of problems with fuel having different geometrical parameters, by means of super-compositions. - Recycle and restart options are available. - UO{sub 2} and PUO{sub 2}-UO{sub 2} fuel can be handled. 2 - Method of solution: The zero-dimension burn-up program RIBOT-5 is coupled with the two-dimension program SQUID and alternately executed. The differential equations are solved by the difference method. 3 - Restrictions on the complexity of the problem: 200 maximum number of compositions 10,000 maximum number of mesh points 5 maximum Number of groups. 4 maximum number of super-compositions. Diagonal symmetry allowed.

Measurements of thermal disadvantage factors in light-water moderated PuO2-UO2 and UO2 lattices

International Nuclear Information System (INIS)

Ohno, Akio; Kobayashi, Iwao; Tsuruta, Harumichi; Hashimoto, Masao; Suzaki, Takenori

1980-01-01

The disadvantage factor for thermal neutrons in light-water moderated PuO 2 -UO 2 and UO 2 square lattices were obtained from measurements of thermal neutron density distributions in a unit lattice cell, measured with Dy-Al wire detectors. The lattices consisted of 3.4 w/o PuO 2 .UO 2 and 2.6 w/o UO 2 fuel rods, and the water-to-fuel volume ratio within the unit cell was parametrically changed. The PuO 2 .UO 2 and UO 2 fuel rods were designed to realize equal fissile atomic number density. The disadvantage factors thus measured were 1.36 +- 0.07, 1.37 +- 0.08, 1.40 +- 0.06 and 1.38 +- 0.06 in the PuO 2 .UO 2 fuel lattices, and 1.30 +- 0.06, 1.31 +- 0.08, 1.30 +- 0.08 and 1.33 +- 0.06 in the UO 2 , for water-to-fuel volume ratios, of 1.76, 2.00, 2.38 and 2.95, respectively. This difference in disadvantage factor between PuO 2 .UO 2 and UO 2 fuel lattices corresponds to about 8%. Calculated results obtained by multigroup transport code LASER agreed well with the measured ones. (author)

Pulsed irradiation of enriched UO{sub 2} in the Annular Core Pulse Reactor (ACPR)

Energy Technology Data Exchange (ETDEWEB)

Schmidt, T R; Lucoff, D M; Reil, K O; Croucher, D W [Sandia Laboratories (United States)

1974-07-01

A series of experiments have been conducted in the Annular Core Pulse Reactor (ACPR) to determine the energy deposition and behavior of enriched UO{sub 2} under pulse conditions. In the experiment single unirradiated pellets with enrichments up to 25 percent were pulse heated to melt temperatures. Temperature and fission product inventory measurements were made and compared with neutron transport calculations. (author)

UO2/magnetite concrete interaction and penetration study

International Nuclear Information System (INIS)

Farhadieh, R.; Purviance, R.; Carlson, N.

1983-01-01

The concrete structure represents a line of defense in safety assessment of containment integrity and possible minimization of radiological releases following a reactor accident. The penetration study of hot UO 2 particles into limestone concrete and basalt concrete highlighted some major differences between the two concretes. These included penetration rate, melting and dissolution phenomena, released gases, pressurization of the UO 2 chamber, and characteristics of post-test concrete. The present study focuses on the phenomena associated with core debris interaction with and penetration into magnetite type concrete. The real material experiment was carried out with UO 2 particles and magnetite concrete in a test apparatus similar to the one utilized in the UO 2 /limestone experiment

State of the art of UO2 fuel fabrication processes

International Nuclear Information System (INIS)

Henke, M.; Klemm, U.

1980-01-01

Starting from the need of UO 2 for thermal power reactors in the period from 1980 to 1990 and the role of UF 6 conversion into UO 2 within the fuel cycle, the state-of-the-art of the three established industrial processes - ADU process, AUC process, IDR process - is assessed. The number of process stages and requirements on process management are discussed. In particular, the properties of the fabricated UO 2 powders, their influence on the following pellet production and on operational behaviour of the fuel elements under reactor conditions are described. Hence, an evaluation of the three essential conversion processes is derived. (author)

Specification of PWR UO2 pellet design parameters with the fuel performance code FRAPCON-1

International Nuclear Information System (INIS)

Silva, A.T.; Marra Neto, A.

1988-08-01

UO 2 pellet design parameters are analysed to verify their influence in the fuel basic properties and in its performance under irradiation in pressurized water reactors. Three groups of parameters are discussed: 1) content of fissionable and impurity materials; 2) stoichiometry; 3) density pore morpholoy, and microstructure. A methodology is applied with the fuel performance program FRAPCON-1 to specify these parameters. (author [pt

Fabrication and testing of ceramic UO{sub 2} fuel - I-III. Part II, Fabrication of sintered pressed samples UO{sub 2} (Final report); Izrada i ispitivanje keramickog goriva na bazi UO{sub 2}- I-III, II Deo - Dobijanje sinterovanih ispresaka UO{sub 2} (zavrsni izvestaj)

Energy Technology Data Exchange (ETDEWEB)

Novakovic, M; Ristic, M M [Institute of Nuclear Sciences Boris Kidric, Laboratorija za termotehniku reaktora, Vinca, Beograd (Serbia and Montenegro)

1961-12-15

Procedure for fabrication of sintered ceramic UO{sub 2} pellets was developed in the Department of reactor materials. The tasks described in this report deal with design and construction of laboratory equipment for treatment of ceramic materials, and fabrication of UO{sub 2} pellets. The procedure was based on cold pressing of appropriately prepared powder and sintering of the of thus obtained pressed samples.

Irradiation of UO2+x fuels in the TANOX device

International Nuclear Information System (INIS)

Dehaudt, P.; Caillot, L.; Delette, G.; Eminet, G.; Mocellin, A.

1998-01-01

The TANOX analytical irradiation device is presented and the first results concerning stoichiometric and hyper stoichiometric uranium dioxide fuels with two different grain sizes are given. The TANOX device is designed to obtain rapidly significant burnups in fuels at relatively low temperatures. It is placed at the periphery of the SILOE reactor and translated to adjust the irradiation power. The continuous measure of the centre-line temperature allows to control the experiment and to evaluate the thermal behaviour of the rods. A TANOX fuel rod has a length of 100 mm with 20 fuel pellets in a stainless steel cladding and is inserted in a thick aluminium alloy overcladding which is cooled by the primary water circuit reactor. These conditions of small size pellets and improved thermal exchanges have been designed to dissipate the heat power due to fission densities three to five times higher than in a PWR. The first analytical irradiation was devoted to the study of UO 2.00 , UO 2.01 and UO 2.02 fuels with standard and large grain sizes obtained by annealing. A burnup of about 9000 MWd.t -1 U was reached in these fuels. The thermal analysis shows a degraded conductivity for the UO 2.02 fuel rod due to the hyper stoichiometry. The released fractions of 85 Kr during irradiation are negligible as expected (lower than 0,1%). Some of the pellets were heat treated at 1700 deg. C for 5 hours. The gas release was analysed after 30 minutes and at the end of the treatment. The main results are as follows: the fission gas release (FGR) of the standard UO 2 varies from one sample to another; the FGR of the hyper stoichiometric fuels is of the same order of magnitude than that of the stoichiometric UO 2 fuel of normal grain sizes; the grain size increase has no effect on FGR for UO 2.00 but considerably decreases the FGR for UO 2.01 and UO 2.02 fuels. These heat treated samples are also observed to characterize the inter- and intragranular fission gas bubbles. (author)

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

International Nuclear Information System (INIS)

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

2005-01-01

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

Thermal-mechanical properties of cracked UO2 pellets

International Nuclear Information System (INIS)

Williford, R.E.; Mohr, C.L.; Lanning, D.D.

1980-11-01

A series of experiments (IFA-431, 432, 513, and 527) sponsored by the Fuel Behavior Research Branch of the USNRC are being irradiated in the Halden Boiling Water Reactor to better define LWR fuel behavior over the normal operating range of power reactor fuel rods. One fuel behavior variable of interest is the thermally induced cracking of UO 2 fuel pellets. The effects of pellet cracking on the effective thermal conductivity and elastic moduli for the fragmented fuel were found to be primarily dependent on the free area in the r, theta plane of the fuel rod. The free area is defined as the area within the cladding inner surface that is not occupied by the fuel fragments themselves

Interim results from UO2 fuel oxidation tests in air

International Nuclear Information System (INIS)

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

1987-08-01

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

Irradiation experiments of recycled PuO2-UO2 fuels by SAXTON reactor, (1)

International Nuclear Information System (INIS)

Yumoto, Ryozo; Akutsu, Hideo

1975-01-01

Seventy two mixed oxide fuel rods made by PNC were irradiated in Saxton Core 3. This paper generally describes the fuel specifications, the power history of the fuel and the post-irradiation examination of the PNC fuel. The specifications of the 4.0 w/o and 5.0 w/o enriched PuO 2 fuel rods with zircaloy-4 cladding are presented in a table and a figure. The positions of PNC fuel rods in the Saxton reactor are shown in a figure. Sixty eight 5.0 w/o PuO 2 -UO 2 fuel rods were assembled in a 9 x 9 rod array together with zircaloy-4 bars, a flux thimble, and a Sb-Be source. The power history of the Saxton Core 3 and the irradiation history of the PNC fuel rods are summarized in tables. The peak power and burnup of each fuel rod and the axial power profile are also presented. The maximum linear power rate and burnup attained were 512W/cm and 8700 MWD/T, respectively. As for the post irradiation examination, the items of nondestructive test, destructive test, and cladding test are presented together with the working flow diagram of the examination. It is concluded that the performance of all fuel rods was safe and satisfactory throughout the power history. (Aoki, K.)

Brandon mathematical model describing the effect of calcination and reduction parameters on specific surface area of UO{sub 2} powders

Energy Technology Data Exchange (ETDEWEB)

Hung, Nguyen Trong; Thuan, Le Ba [Institute for Technology of Radioactive and Rare Elements (ITRRE), 48 Lang Ha, Dong Da, Ha Noi (Viet Nam); Van Khoai, Do [Micro-Emission Ltd., 1-1 Asahidai, Nomi, Ishikawa, 923-1211 (Japan); Lee, Jin-Young, E-mail: jinlee@kigam.re.kr [Convergence Research Center for Development of Mineral Resources (DMR), Korea Institute of Geoscience and Mineral Resources (KIGAM), Daejeon, 305-350 (Korea, Republic of); Jyothi, Rajesh Kumar, E-mail: rkumarphd@kigam.re.kr [Convergence Research Center for Development of Mineral Resources (DMR), Korea Institute of Geoscience and Mineral Resources (KIGAM), Daejeon, 305-350 (Korea, Republic of)

2016-06-15

Uranium dioxide (UO{sub 2}) powder has been widely used to prepare fuel pellets for commercial light water nuclear reactors. Among typical characteristics of the powder, specific surface area (SSA) is one of the most important parameter that determines the sintering ability of UO{sub 2} powder. This paper built up a mathematical model describing the effect of the fabrication parameters on SSA of UO{sub 2} powders. To the best of our knowledge, the Brandon model is used for the first time to describe the relationship between the essential fabrication parameters [reduction temperature (T{sub R}), calcination temperature (T{sub C}), calcination time (t{sub C}) and reduction time (t{sub R})] and SSA of the obtained UO{sub 2} powder product. The proposed model was tested with Wilcoxon's rank sum test, showing a good agreement with the experimental parameters. The proposed model can be used to predict and control the SSA of UO{sub 2} powder.

Construction of the vibrator for UO{sub 2} powder compacting; Izrada vibratora za kompaktiranje praha UO{sub 2}

Energy Technology Data Exchange (ETDEWEB)

Vrgora, M [Institute of Nuclear Sciences Boris Kidric, Laboratorija za termotehniku reaktora, Vinca, Beograd (Serbia and Montenegro)

1961-12-15

This report contains the description and the scheme of the device for vibration compacting of sintered UO{sub 2} powder. This device was designed and constructed in the Department for reactor materials.

Geometrical dimensioning of PWR UO2 pellets

International Nuclear Information System (INIS)

Silva, A.T.

1988-08-01

The finite element structural program SAP-IV is used to calculate UO 2 pellet strains developed under thermal gradients in pressurized water reactors. The applied procedure allows to analyse the influence of various aspects of pelet geometry on cladding strains and can be utilized for the dimensioning of UO 2 pellets. Pellets purchased with flat ends, with dishes pressed into both ends, shouders, and a 45-deg edge chamfer are analysed. The analyse results are compared with experiemtnal data. (author) [pt

Interaction and penetration of heated UO2 with limestone concrete

International Nuclear Information System (INIS)

Farhadieh, R.; Pedersen, D.R.; Purviance, R.; Carlson, N.

1982-01-01

To safeguard the environment against radiological releases, the major question of concern in PAHR safety assessment, following an HCDA, involves confinement and dilution of the molten core-debris. Significant to the study is the directional growth of the core-debris in the concrete foundation of the reactor building or the concrete below the reactor cavity. The real material experiments were carried out in the test apparatus shown. Casts of CRBRP limestone concrete were prepared in graphite cylinders, each having an internal diameter of 8.9 cm and a depth of 30.5 cm. The 17.8-cm-deep concrete samples were allowed to cure for at least 28 days. Experiments were conducted within two months of curing time. The cavity above concrete was packed with 3 kg of pure UO 2 particles (1 to 3 mm). A uranothermic mixture was placed on the top of UO 2 powder. Heating and possible melting of UO 2 was achieved resistively after the ignition of the thermite. Total experimental time was about 60 minutes, during which time a maximum electrical power input of 1.8 watts/gr was applied to the UO 2 . Three gas samples were taken at temperatures of 100, 600, and 950 0 C, measured in the plane of the No. 2 thermocouple. Selection of three temperatures were to study the amount and the type of gases released from different phases of concrete

Methods of modification and investigations of properties of fuel UO2

International Nuclear Information System (INIS)

Kurina, I.; Popov, V.; Rogov, S.; Dvoryashin, A.; Serebrennikova, O.

2009-01-01

In the SSC RF-IPPE the researches are carried out directed towards the uranium dioxide fuel pellets modification with the purpose of improvement of their performance parameters (increase of thermal conductivity, growth of grain for decrease gas release, decrease of interaction with coolant). The following technological methods of manufacturing of modified pellets UO 2 were used: 1) The water method including precipitation of Ammonium Polyuranate (APU) with manufacturing of simultaneously coarse and super dispersed particles, and also coprecipitation APU with additives (Cr, Ti, etc.), with the after calcination of powders, reduction to UO 2 pressing and sintering of pellets; 2) A method including addition of chemical reagent containing ammonia to the powder UO 2 manufactured under the dry or water technology; mechanical mixture; pressing and sintering of pellets. Application of the specified up methods makes manufacturing the UO 2 fuel pellets having the properties differing from pellets manufactured by industrial technology. Conclusions: 1) Properties of powders and the pellets manufactured by different technologies considerably differ; 2) Precipitate manufactured by water industrial technology, consists of phase NH 3 ·3UO 3 ·5H 2 O whereas the precipitate manufactured by nanotechnology contains in addition phase NH 3 ·2UO 3 ·3H 2 O; 3) Powders of U 3 O 8 manufactured by water nanotechnology have particles size 300-500 nm and ultra dispersive particles size ∼70-75 nm; 4) Powder UO 2 obtained by water nanotechnology differs by greater activity because all phase changes under oxidation result at lower temperatures; 5) Basic differences of properties of modified UO 2 pellets was established: decreasing of defects inside and on grains boundaries, minor porosity (pore size 0,05-0,5 μm), presence of pore of spherical form, presence of additional chemical bond U-U (presence of metal clusters), polyvalence of U; 6) Methods including addition of Cr and Ti under

Development of AUC-based process at BARC for production of free-flowing and sinterable UO2 powder

International Nuclear Information System (INIS)

Keni, V.S.; Ghosh, S.K.; Ganguly, C.; Majumdar, S.

1994-01-01

Ammonium uranium carbonate (AUC) process has been developed and industrially used in Germany for preparation of free-flowing and sinterable UO 2 powder for fabrication of UO 2 fuel pellets for light water reactors (LWR). Efforts are underway at Bhabha Atomic Research Centre (BARC) for developing AUC-based process which would yield free-flowing UO 2 powder suitable for direct pelletisation and sintering to very high density (> 96% T.D.) UO 2 fuel pellets for pressurised heavy water reactors (PHWRs) in India. The first phase of this work has been completed jointly by Chemical Engineering Division (ChED) and Radiometallurgy Division (RMD) in batches of 1.5 kg. It was possible to fabricate UO 2 pellets of density 93-95% T.D. on a reproducible basis. At ChED, process parameters have been optimised for fabrication of AUC with suitable physical properties in batches of 1.5 kg (U), starting with nuclear pure uranyl nitrate solution. At RMD calcination parameters of AUC was optimised in batches of 500 g for obtaining free-flowing UO 2 powder, suitable for direct pelletisation and sintering. The pelletisation and sintering have been carried out at Radiometallurgy Division in batches of 1-1.5 kg. The maximum achievable density of UO 2 pellets has been in the range of 95.5-96% T.D. (author). 11 refs

Variable dimensionality and new uranium oxide topologies in the alkaline-earth metal uranyl selenites AE[UO2)(SeO3)2] (AE=Ca, Ba) and Sr[UO2)(SeO3)2] · 2H2O

International Nuclear Information System (INIS)

Almond, Philip M.; Peper, Shane M.; Bakker, Eric; Albrecht-Schmitt, Thomas E.

2002-01-01

Three new alkaline-earth metal uranyl selenites, Ca[UO 2 )(SeO 3 ) 2 ] (1), Sr[UO 2 )(SeO 3 ) 2 ] · 2H 2 O (2), and Ba[UO 2 )(SeO 3 ) 2 ] (3), have been prepared from the reactions of CaCO 3 and Ca(OH) 2 , SrCl 2 and Sr(OH) 2 , or BaCl 2 and Ba(OH) 2 with UO 3 and SeO 2 under mild hydrothermal conditions. Single-crystal X-ray diffraction experiments reveal that the structures of 1-3 differ in both connectivity and dimensionality even though all contain the same fundamental building unit, namely [UO 2 (SeO 3 ) 4 ]. This polyhedron consists of a linear uranyl unit that is bound by one chelating and three bridging selenite anions creating a pentagonal bipyramidal environment around the U(VI) center. The crystal structure of 1 contains one-dimensional ribbons where the edges are terminated by monodentate selenite anions. The interior of the ribbons are constructed from edge-sharing pentagonal bipyramidal UO 7 units. The structure of 2 is also one-dimensional; however, here there are chains of edge-sharing pentagonal bipyramidal UO 7 dimers that are connected by bridging selenite anions. Ba[(UO 2 )(SeO 3 ) 2 ] (3) is two-dimensional, and the highly ruffled anionic sheets present in this structure are formed from both bridging and chelating/bridging selenite anions bound to uranyl moieties. The anionic substructures in 1-3 are separated by Ca 2+ , Sr 2+ , or Ba 2+ cations. Crystallographic data (193 K, MoKα, λ=0.71073): 1, triclinic, space group P1-bar, a=5.5502(6) A, b=6.6415(7) A, c=11.013(1) A, α=104.055(2) deg., β=93.342(2) deg., γ=110.589(2) deg. , Z=2, R(F)=4.56% for 100 parameters with 1530 reflections with I>2σ(I); 2, triclinic, space group P1-bar, a=7.0545(5) A, b=7.4656(5) A, c=10.0484(6) A, α=106.995(1) deg., β=108.028(1) deg., γ=98.875(1) deg., Z=2, R(F)= 2.43% for 128 parameters with 2187 reflections with I>2σ(I); 3, monoclinic, space group P2 1 /c, a=7.3067(6) A, b=8.1239(7) A, c=13.651(1) A, β=100.375(2) deg., Z=4, R(F)=4.31% for 105 parameters

Chemical and spectrochemical production analysis of ThO2 and 233UO2-ThO2 pellets for the light water breeder reactor core for Shippingport (LWBR development program)

International Nuclear Information System (INIS)

Bukowski, J.F.; Hollis, E.D.

1975-06-01

The Bettis Atomic Power Laboratory has utilized wet chemical, emission spectrochemical, and mass spectrometric analytical techniques for the production analysis of the ThO 2 and 233 UO 2 -ThO 2 (1 to 6 wt percent 233 UO 2 ) pellets for the Light Water Breeder Reactor (LWBR) core for Shippingport. Proof of the fuel breeding concept necessitates measurement of precise and accurate chemical characterization of all fuel pellets before core life. Chemistry's efforts toward this goal are presented in three main sections: (1) general discussions relating the chemical requirements for ThO 2 and 233 UO 2 -ThO 2 core materials to the analytical capabilities, (2) technical discussions of the chemical and instrumental technology applied for the analysis of aluminum, boron, calcium, carbon, chloride plus bromide, chromium, cobalt, copper, dysprosium, europium, fluoride, gadolinium, iron, magnesium, manganese, mercury, molybdenum, nickel, nitrogen, samarium, silicon, titanium, vanadium, thorium, and uranium (total, trace, and uranium VI), and (3) a formal presentation of the analytical procedures as applied to the LWBR Development Program. (U.S.)

In-pile vapor pressure measurements on UO2 and (U,Pu)O2

International Nuclear Information System (INIS)

Breitung, W.; Reil, K.O.

1985-08-01

The Effective-Equation-of-State (EEOS) experiments investigated the saturation vapor pressures of ultra pure UO 2 , reactor grade UO 2 , and reactor grade (Usub(.77)Pusub(.23))O2 using newly developed in-pile heating techniques. For enthalpies between 2150 and 3700 kJ/kg (about 4700 to 8500 K) vapor pressures from 1.3 to 54 MPa were measured. The p-h curves of all three fuel types were identical within the experimental uncertainties. An assessment of all published p-h measurements showed that the p-h saturation curve of UO 2 appears now well established by the EEOS and the CEA in-pile data. Using an estimate for the heat capacity of liquid UO 2 , the in-pile results were also compared to earlier p-T measurements. The assessments lead to proposal of two equations. Equation I, which includes a factor-of-2 uncertainty band, covers all p-T equilibrium evaporation measurements. Equation I yields 3817 K for the normal boiling point, 415.4 kJ/mol for the corresponding heat of vaporization, and 1.90 MPa for the vapor pressure at 5000 K. Equations I and II, which represent a parametric form of the p-h curve (T=parameter), also give a good description of the EEOS and CEA in-pile data. Thus the proposed equations allow a consistent representation of both p-T and p-h measurements, they are sufficiently precise for CDA analyses and cover the whole range of interest (3120-8500 K, 1400-3700 kJ/kg). (orig./HP) [de

The Influences of Uranium Concentration and Polyvinyl Alcohol on the Quality UO2 Microsphere for Fuel of High Temperature Reactor

International Nuclear Information System (INIS)

Damunir; Sukarsono; Bangun-Wasito; Endang Nawangsih

2000-01-01

The influences of uranium concentration and PVA on the quality of UO 2 microspheres for fuel of high temperature reactor have been investigated. The UO 2 particles were prepared by gel precipitation using internal gelation process. Uranyl nitrate solution containing uranium of 100 g/l was neutralized using NH 4 OH 1 M. The solution was changed into sol by adding 60 g PVA/l solution while stirred and heated up to 80 o C for 20 minutes. In order to find gels in spherical shape, the sol solution was dropped into 5 M NH 4 OH medium. The formed gels were small spheres, was washed, screened and heated up to 120 o C. After that, the gels were calcined at 800 o C for 4 hours, resulting in U 3 O 8 spheres. The U 3 O 8 particles were reduced using H 2 gas in a N 2 media at 800 o C for 4 hours, yielded in UO 2 spheres. Using a similar procedure, the influence of uranium concentration of 150-250 g/l and PVA 40-80 g/l were studied. The qualities of UO 2 particles were obtained by their physical properties, i.e. density, specific surface area, total volume of pores and pore radius using surface area meter and N 2 gas used as absorbent, and the particle size was observed using optical microscope. The result showed that the changing of uranium and PVA concentrations on the internal gelation affected the density, specific surface area, total volume of pores and pore radius of UO 2 particles. (author)

Fissile fuel production and usage of thermal reactor waste fueled with UO2 by means of hybrid reactor system

International Nuclear Information System (INIS)

Ipek, O.

1997-01-01

The use of Fast Breeder Reactors to produce fissile fuel from nuclear waste and the operation of these reactors with a new neutron source are becoming today' topic. In the thermonuclear reactors, it is possible to use 2.45-14.1 MeV - neutrons which can be obtained by D-T, D-D Semicatalyzed (D-D) and other fusion reactions. To be able to do these, Hybrid Reactor System, which still has experimental and theoretical studies, have to be taken into consideration.In this study, neutronic analysis of hybrid blanket with grafit reflector, is performed. D-T driven fusion reaction is surrounded by UO 2 fuel layer and the production of ''2''3''9Pu fissile fuel from waste ''2''3''8U is analyzed. It is also compared to the other possible fusion reactions. The results show that 815.8 kg/year ''2''3''8Pu with D-T reaction and 1431.6 kg/year ''2''3''8Pu with semicatalyzed (D-D) reaction can be produced for 1000 MW fusion power. This means production of 2.8/ year and 4.94/ year LWR respectively. In addition, 1000 MW fusion flower is is multiplicated to 3415 MW and 4274 MW for D-T and semicatalyzed (D-D) reactions respectively. The system works subcritical and these values are 0.4115 and 0.312 in order. The calculations, ANISN-ORNL code, S 16 -P 3 approach and DLC36 data library are used

Heat transfer coefficient between UO2 and Zircaloy-2

International Nuclear Information System (INIS)

Ross, A.M.; Stoute, R.L.

1962-06-01

This paper provides some experimental values of the heat-transfer coefficient between UO 2 and Zircaloy-2 surfaces in contact under conditions of interfacial pressure, temperature, surface roughness and interface atmosphere, that are relevant to UO 2 /Zircaloy-2 fuel elements operating in pressurized-water power reactors. Coefficients were obtained from eight UO 2 / Zircaloy-2 pairs in atmospheres of helium, argon, krypton or xenon, at atmosphere pressure and in vacuum. Interfacial pressures were varied from 50 to 550 kgf/cm 2 while surface roughness heights were in the range 0.2 x 10 -4 to 3.5 x 10 -4 cm. The effect on the coefficients of cycling the interfacial pressure, of interface gas pressure and of temperature were examined. The experimental values of the coefficients were used to test the predictions of expressions for the heat-transfer between two solids in contact. For the particular UO 2 / Zircaloy-2 pairs examined, numerical values were assigned to several parameters that related the surface roughnesses to either the radius of solid/solid contact spots or to the mean thickness of the interface voids and that accounted for the imperfect accommodation of the void gas on the test surfaces. (author)

Critical sizes of light-water moderated UO2 and PuO2-UO2 lattices

International Nuclear Information System (INIS)

Tsuruta, Harumichi; Kobayashi, Iwao; Suzuki, Takenori; Ohno, Akio; Murakami, Kiyonobu

1978-02-01

Experimental critical sizes are presented for a total of about 250 lattices with 2.6 w/o UO 2 and 3.0 w/o PuO 2 -natural UO 2 fuel rods. The moderator was H 2 O and water-to-fuel volume ratios in the lattice cells ranged from 1.50 to 3.00 in the UO 2 lattices and from 2.42 to 5.55 in the PuO 2 -UO 2 lattices. The critical sizes were determined with the number of the fuel rods and a water level which were required to make the lattice critical in the shape of a rectangular parallelepiped over the temperature range from room temperature to 80 0 C. Reactivity variations of the PuO 2 -UO 2 lattices due to decaying of 241 Pu to 241 Am were traced during 3 years. Some critical sizes of the UO 2 and PuO 2 -UO 2 lattices with a water gap and of the UO 2 lattices with liquid poison in the moderator are also reported. Some physics parameters, such as the temperature coefficient of reactivity, the water-level worth, the reflector saving, the ratio between a migration area and an infinite multiplication factor and the critical buckling, are shown in relation to the critical sizes of the unperturbed lattices without the water gap and liquid poison. (auth.)

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

Energy Technology Data Exchange (ETDEWEB)

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

2005-07-01

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

Geometric dimensioning of UO2 pellets for PWR

International Nuclear Information System (INIS)

Teixeira e Silva, A.

1988-01-01

The finite element structural program SAP-IV is used to calculate UO 2 pellet strains developed under thermal gradients in pressurized water reactors. The applied procedure allows to analyse the influence of various aspects of pellet geometry on cladding strains and can be utilized for the dimensioning of UO 2 pellets. Pellets purchased with flat ends, with dishes pressed into both ends, shouders, and a 45-deg edge chamfer are analysed. The analyse results are compared with experimental data.(autor) [pt

Oxidative dissolution of ADOPT compared to standard UO2 fuel

International Nuclear Information System (INIS)

Nilsson, Kristina; Roth, Olivia; Jonsson, Mats

2017-01-01

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

Thermal and in-reactor densification of UO2; mechanism and experimental results

International Nuclear Information System (INIS)

Assmann, H.; Stehle, H.

1980-01-01

Suggested is a generalized model of UO 2 densification kinetics under irradiation in a reactor which takes into account the peculiarities of small and large pores behaviour in four temperature ranges (450 deg C, 450-750 deg C; 750-1300 deg C; 1300 deg C ) determining the process. It is pointed out that one of the most important parameters influencing the speed of densification is an initial distribution of pores according to dimensions. Summary VO 2 volume change under irradiation is shown to be obtained from decreasing the volume at the expense of pore shrinkage and matrix swelling. The model also takes into account such parameters as irradiation time and temperature, burning, and initial fuel density. The densification model suggested is confirmed by many experimental data

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

Energy Technology Data Exchange (ETDEWEB)

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

1978-07-01

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

Unirradiated UO2 in irradiated zirconium alloy sheathing

International Nuclear Information System (INIS)

MacDonald, R.D.; Hardy, D.G.; Hunt, C.E.L.; Scoberg, J.A.

1979-07-01

Zircaloy-clad UO 2 fuel elements have defected in power reactors when element power outputs were raised significantly after a long irradiation at low power. We have irradiated fuel elements fabricated from fresh UO 2 pellets and zirconium alloy sheaths previously irradiated without fuel. This gave a fuel element with radiation-damaged low-ductility sheathing but with no fission products in the fuel. The elements were power boosted in-reactor to linear power outputs up to 84 kW/m for two five-day periods. No elements defected despite sheath strains of 0.82 percent at circumferential ridge postions. Half of these elements were subsequently soaked at low power to build up the fission product inventory in the fuel and then power boosted to 63 kW/m for a third time. Two elements defected on this final boost. We conclude that these defects were caused by fission product induced stress-corrosion cracking and that this mechanism plays an importent role in power reactor fuel defects. (auth)

Hydrothermal synthesis and crystal structures of new uranyl oxalate hydroxides: α- and β-[(UO2)2(C2O4)(OH)2(H2O)2] and [(UO2)2(C2O4)(OH)2(H2O)2].H2O

International Nuclear Information System (INIS)

Duvieubourg, Laurence; Nowogrocki, Guy; Abraham, Francis; Grandjean, Stephane

2005-01-01

Two modifications of the new uranyl oxalate hydroxide dihydrate [UO 2 ) 2 (C 2 O 4 )(OH) 2 (H 2 O) 2 ] (1 and 2) and one form of the new uranyl oxalate hydroxide trihydrate [(UO 2 ) 2 (C 2 O 4 )(OH) 2 (H 2 O) 2 ].H 2 O (3) were synthesized by hydrothermal methods and their structures determined from single-crystal X-ray diffraction data. The crystal structures were refined by full-matrix least-squares methods to agreement indices R(wR)=0.0372(0.0842) and 0.0267(0.0671) calculated for 1096 and 1167 unique observed reflections (I>2σ(I)), for α (1) and β (2) forms, respectively and to R(wR)=0.0301(0.0737) calculated for 2471 unique observed reflections (I>2σ(I)), for 3. The α-form of the dihydrate is triclinic, space group P1-bar , Z=1, a=6.097(2), b=5.548(2), c=7.806(3)A, α=89.353(5), β=94.387(5), γ=97.646(5) o , V=260.88(15)A 3 , β-form is monoclinic, space group C2/c, Z=4, a=12.180(3), b=8.223(2), c=10.777(3)A, β=95.817(4), V=1073.8(5)A 3 . The trihydrate is monoclinic, space group P2 1 /c, Z=4, a=5.5095(12), b=15.195(3), c=13.398(3)A, β=93.927(3), V=1119.0(4)A 3 . In the three structures, the coordination of uranium atom is a pentagonal bipyramid composed of dioxo UO 2 2+ cation perpendicular to five equatorial oxygen atoms belonging to one bidentate oxalate ion, one water molecule and two hydroxyl ions in trans configuration in 2 and in cis configuration in 1 and 3. The UO 7 polyhedra are linked through hydroxyl oxygen atoms to form different structural building units, dimers [U 2 O 10 ] obtained by edge-sharing in 1, chains [UO 6 ] ∼ and tetramers [U 4 O 26 ] built by corner-sharing in 2 and 3, respectively. These units are further connected by oxalate entities that act as bis-bidentate to form one-dimensional chains in 1 and bi-dimensional network in 2 and 3. These chains or layers are connected in frameworks by hydrogen-bond arrays

Ceramic UO2 powder production at Cameco Corporation

International Nuclear Information System (INIS)

Kwong, A.K.; Kuchurean, S.M.

1997-01-01

This presentation covers the various aspects of ceramic grade uranium dioxide (UO 2 ) powder production at Cameco Corporation and its use as fuel and blanket fuel for heavy-water and light-water reactors, respectively. In addition, it discusses the significant production variables that affect production and product quality. It also provides an insight into how various support groups such as Quality Assurance, Analytical Services, and Technology Development fit into the quality cycle and contribute to a successful operation. The ability of Cameco to identify, measure and control the physical and chemical properties of ceramic grade UO 2 has resulted in the production of uniform quality powder. This has meant that 100% of Cameco's ceramic grade UO 2 powder produced since mid-1989 has been accepted by the fuel manufacturers. (author)

Synthesis report of a reactor-physics programme devoted to the qualification of the project calculation methods applied to the mixed (U-Pu) fuelled reloadings

International Nuclear Information System (INIS)

Golinelli, C.

1982-01-01

The Commission of the European Communities has sponsored three sets of experiments with a contract in the framework of the LWR plutonium recycling programme. The aim of this experimental programme was to supplement the qualification of the computer codes used in connection with control and safety studies relating to plutonium recycling in LWR power stations. The experiments are performed at the Cadarache centre by CEA and EdF in the MINERVE and EOLE facilities. The power distributions in various kinds of assemblies with some heterogeneities (water holes, absorbers, UO 2 /UO 2 -PuO 2 interfaces) are studied in MINERVE as well as the DOPPLER effect. This second programme consisted in oscillating the UO 2 and UO 2 -PuO 2 samples which were heated between 20 0 C and 800 0 C. The CREOLE experiments in EOLE dealt with the moderator temperature effect from 20 0 C to 300 0 C. All the results show that the calculation schema for the hot point values must be sophisticated and that it is possible to reduce the uncertainties concerning the feedback factors

Study on Reactor Physics Characteristic of the PWR Core Using UO2

International Nuclear Information System (INIS)

Tukiran Surbakti

2009-01-01

Study on reactor physics characteristic of the PWR core using UO 2 fuel it is necessary to be done to know the characteristic of geometry, condition and configuration of pin cell in the fuel assembly Because the geometry, configuration and condition of the pin cell in fuel core determine the loading strategy of in-core fuel management Calculation of k e ff is a part of the neutronic core parameter calculation to know the reactor physics characteristic. Generally, core calculation is done using computer code starts from modelling one unit fuel lattice cell, fuel assembly, reflector, irradiation facility and until core reactor. In this research, the modelling of pin cell and fuel assembly of the PWR 17 ×17 is done homogeneously. Calculation of the k-eff is done with variation of the fuel volume fraction, fuel pin diameter, fuel enrichment. The calculation is using by NITAWL and CENTRM, and then the results will be compared to KENOVI code. The result showed that the value of k e ff for pin cell and fuel assembly PWR 17 ×17 is not different significantly with homogenous and heterogenous models. The results for fuel volume fraction of 0.5; rod pitch 1.26 cm and fuel pin diameter of 9.6 mm is critical with burn up of 35,0 GWd/t. The modeling and calculation method accurately is needed to calculation the core physic parameter, but sometimes, it is needed along time to calculate one model. (author)

Physical and chemical characterization of the (Th, U)O2 mixed oxide fuel

International Nuclear Information System (INIS)

Santos, A.M.M. dos; Avelar, M.M.; Palmieri, H.E.L.; Lameiras, F.S.; Ferreira, R.A.N.

1986-01-01

The NUCLEBRAS R and D Center (Centro de Desenvolvimento da Tecnologia Nuclear - CDTN) has been performing, together with german institutions (Kernforschungsanlage Julich GmbH - KFA, Krafwerk Union A.G. - KWU and NUKEM GmbH), a program for utilization of thorium in pressurized water reactors. In this paper are presented the physical and chemical characterizations necessary to quality the (Th, U)O 2 fuel and the respective methods. (Author) [pt

Analysis of flux standards in a fluized bed for AUC - UO2 convertion

International Nuclear Information System (INIS)

Juanico, L.E.; Clausse, A.; Guido Lavalle, G.

1990-01-01

One of the fuel cycle stages is the convertion (reduction) of ammonium uranyl carbonate (AUC) in UO 2 which, after being directly compacted, allows pellet obtainment acquire the correct density to be used as nuclear fuel during sintering. AUC's reduction in UO 2 is made on a fluidized bed in which AUC powder going into the upper part at a countercurrent to the gas flux (superheated steam), is converted into UO 2 ; after the reaction, UO 2 is collected at the lower part of the reactor. (Author) [es

Interactions with Small and Large Sodium to UO2 Mass Ratios

International Nuclear Information System (INIS)

Clerici, G.; Holtbecker, H.; Schins, H.; Schlittenbardt, P.

1976-01-01

This paper is divided into the following three parts: - Presentation of final results of the Ispra dropping experiments; - Discussion of preliminary Na entrapment tests; - Presentation of the Press I and II codes. The experiments for which the Ispra UO 2 dropping facility was originally designed were completed in 1975. The experimental facility which initially had had difficulties in reaching the predefined working conditions gave in the last year a series of results. For this reason Ispra decided to built a similar plant for dropping experiments into water which started working in 1975. Concerning the entrapment tests it was originally foreseen to built in collaboration with GfK Karlsruhe a test section having subassembly geometry and in which the UO 2 would have been violently dispersed into the surrounding Na by the expansion of a small quantity of superheated sodium. Preliminary tests and the design work for the facility could be completed. The Press I + II codes were developed to support the above mentioned experiment - al activity. A 1-D analysis is made to investigate phenomena like UO 2 crust formation and calculate delay times between the time of the Na injection into UO 2 and the violent expansion of superheated Na. An estimate was also made of the available mechanical work in such a process which should allow to get an idea of possible energy release in a reactor core. First conclusions can be drawn from this estimate concerning the mechanical energy release in a WCA due to SPI. The result is that considerably lower energies are calculated from Na entrapment in a reactor core due to the limited amount of molten UO 2 present in the core

An improved slow neutron spectrometer at nuclear research reactor et-r r-1. Vol. 2

Energy Technology Data Exchange (ETDEWEB)

Abu El-Ela, M A [Reactor and Neutron Physics, Nuclear Research Center, AEA, Cairo (Egypt)

1996-03-01

An improved slow neutron selector has been aligned at channel number 6 of the nuclear research reactor ET-R R-1 Inshas. The flight path is 4 meter. The collimator-rotor-collimator system has the dimensions 0.3 x 2.5 x 70 cm with the rotor diameter 16 cm and 3 slits of 0.3 x 2.5 cm cross section. The rotor rotation rate varies between 600 r.p.m. the counting system has one of the best modern high electronic advanced technology time analyzer with minimum dwell time 2 sec, 8192 channels and a double detector inputs of TTL and NEG NIM standard pulses. The analyzer external triggering signals are of TTL standard type. A special design {sup 3} He detector for time of flight spectrometry has been used in the SNS. The reactor bare thermal neutron spectrum has been successfully measured, to show good agreement with the previous data. 6 figs.

Perovskite phases in the systems Asup(II)O-UO/sub 3/. 1. Tetragonal perovskite Ba/sub 2/Basub(7/8)vacantsub(1/8)UO/sub 5/sub(7/8)

Energy Technology Data Exchange (ETDEWEB)

Griffiths, A J; Kemmler-Sack, S [Tuebingen Univ. (Germany, F.R.)

1979-10-01

The new tetragonal compound Ba/sub 2/Basub(7/8)vacantsub(1/8)UO/sub 5/sub(7/8) (a = 2 x 6.31/sub 2/ A; c = 2 x 8.76/sub 7/ A) has been found besides Ba/sub 3/UO/sub 6/ (triclinic) in the BaO-UO/sub 3/ system. It crystallizes with a superstructure of perovskite type. The differences in properties between Ba/sub 3/UO/sub 6/ and Ba/sub 2/Basub(7/8)vacantsub(1/8) UO/sub 5/sub(7/8) are discussed.

Electron probe micro-analysis of irradiated Triso-coated UO2 particles, (1)

International Nuclear Information System (INIS)

Ogawa, Toru; Minato, Kazuo; Fukuda, Kosaku; Ikawa, Katsuichi

1983-11-01

The Triso-coated low-enriched UO 2 particles were subjected to the post-irradiation electron probe micro-analysis. Observations and analyses on the amoeba effect, inclusions and solutes in the UO 2 matrix were made. In the cooler side of the particle which suffered extensive kernel migration, two significant features were observed: (1) the wake of minute particles, presumably UO 2 , left by the moving kernel in the carbon phase and (2) carbon precipitation in the pores and along the grain boundaries of the UO 2 kernel. Both features could be hardly explained by the gas-phase mechanism of carbon transport and rather suggest the solid state mechanism. Two-types of 4d-transition metal inclusions were observed: the one which was predominantly Mo with a fraction of Tc and another which was enriched with Ru and containing significant amount of Si. The Mo and Si were also found in the UO 2 matrix; the observation led to the discussion of the oxygen potential in the irradiated Triso-coated UO 2 particle. (author)

Contribution to the qualification of Gd calculation in PWR reactors

International Nuclear Information System (INIS)

Chaucheprat, Patrick.

1982-06-01

This thesis presents the state of knowledge on gadolinium and the advantages of its use as burnable poison. A study on the behaviour of gadolinium makes it possible to bring out the essential parameters to which it is sensitive. The most important part of this work is devoted to the measurements by oscillations carried out in Minerve in 1981. The conceiving and implementation of this campaign are reported. The experimental results and the amending factors linked to the interpretation are presented. To complete this study at zero time, it seemed useful to process configurations with fuel clusters of UO 2 - Gd 2 O 3 in order to see the effect of UO 2 - Gd 2 O 3 rods in interaction. To this end, efficiency determinations of UO 2 - Gd 2 O 3 rod clusters were carried out in the Melodie lattice. The second part of this work involves the change in the gadolinium. Two main points are tackled here. The first concerns the determinations by oscillations of ''reconstituted'' samples that are composed of two concentric rings with various 235 U enrichments and gadolinium levels so as to simulate irradiated UO 2 - Gd 2 O 3 fuel. The second point is devoted to the description of the GEDEON experiment. UO 2 - Gd 2 O 3 rods will be irradiated in a 13 x 13 lattice of which the spectrum is representative of that of a PWR. This experiment will take place in the centre of the Melusine reactor at Grenoble [fr

A Characterization Research of UO2 Powder for UO2 Pellet Fabrication of Candu Type

International Nuclear Information System (INIS)

Rachmawati, M.

1998-01-01

A characterization research of of UO 2 powder for UO 2 pellet fabrication of Candu type is reported in this paper. The research has been conducted by characterizing sinterability, compactibility, and compressibility of UO 2 (Cameco) without a pre-compacting and UO 2 powder the result of a pre-compacting. The pre-compacting UO 2 powder has been done to have particle size to less than 150 mu (150-800) mu, and more than 800 mu with distribution varied. Sinterability of each group of particle sizes is analyzed using Thermogravimetric-Differential Thermal Analysis (TG-DTA). Then the final compacting to the powder is done using compaction pressure varied from 1 MP to 4 MP to the all groups of the particle sizes to find the optimum pressure by measuring the density and mechanical strength of the UO 2 green pellet. Both measurements are performed using Micrometer and Universal Testing Machine respectively. The result of this investigation shows that the group of UO 2 powder with no pre-compacting with particle size of less than 150 mu with 60% distribution and (150-800) mu size with 40% distribution are the UO 2 pellets which are eligible in terms of their density and mechanical strength

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-05-15

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

IEA-R1 reactor - Spent fuel management

International Nuclear Information System (INIS)

Mattos, J.R.L. De

1996-01-01

Brazil currently has one Swimming Pool Research Reactor (IEA-R1) at the Instituto de Pesquisas Energeticas e Nucleares - Sao Paulo. The spent fuel produced is stored both at the Reactor Pool Storage Compartment and at the Dry Well System. The present situation and future plans for spent fuel storage are described. (author). 3 refs, 2 figs, 2 tabs

Complexing in (NH4)2SeO4-UO2SeO4 H2O system

International Nuclear Information System (INIS)

Serezhkina, L.B.

1994-01-01

Isotherm of solubility in the (NH 4 ) 2 SeO 4 -UO 2 SeO 4 -H 2 O system has been constructed at 25 deg C. (NH 4 ) 2 (UO 2 ) 2 (SeO 4 ) 3 x6H 2 O formation is established for the first time and certain its physicochemical properties are determined. Regularities of complexing in the R 2 Se) 4 -UO 2 SeO 4 -H 2 O systems, where R-univalent cation are under discussion. 6 refs.; 3 tabs

Leaching Studies on ACR-1000{sup R} Fuel Under Reactor Operating Conditions

Energy Technology Data Exchange (ETDEWEB)

Sunder, S. [Atomic Energy of Canada Limited, Fuel and Fuel Channel Safety Branch, Chalk River, Ontario, K0J 1J0 (Canada)

2009-06-15

ACR-1000{sup R} is the latest nuclear power reactor being developed by AECL. The ACR-1000 fuel uses a modified CANFLEX{sup R} fuel bundle that contains low-enriched uranium and pellets of burnable neutron absorbers (BNA) in a central element. Dysprosium and gadolinium are used as the burnable neutron absorbers and are present as oxides in a 'fully-stabilized' zirconia matrix. The BNA material in the centre element is designed to limit the coolant void reactivity of the reactor core during postulated loss-of-coolant accidents. As part of the ACR-1000 fuel development, the stability of the BNA material under conditions associated with defects of the Zircaloy sheathing of the BNA central element has been investigated. The results of these tests can be used to demonstrate the phase stability and leaching behaviour of the ACR-1000 fuel under reactor operating conditions. The samples were disks, about 3-4 mm thick, obtained from BNA pellets and Candu fuel (natural uranium UO{sub 2}) pellets (the UO{sub 2} measurements provide a reference point). Leaching tests were carried out in light water at 325 deg. C, above the maximum coolant temperature in an ACR-1000 fuel channel during normal operating conditions (319 deg. C). This temperature also bounds the maximum operating temperature for the current Candu reactors (311 deg. C). The initial pH of the solution (measured at room temperature) used in the leaching tests was 10.3. The leach rates were determined by monitoring the amount of metals leached into solutions. Leaching tests were also carried out with BNA pellet samples in the presence of Zr-2.5%Nb pressure tube coupons to determine the effects, if any, of the presence of pressure tube material on leach rates. Other leaching tests with BNA pellet samples and UO{sub 2} pellets were conducted at 80 deg. C to study the effects of temperature on the leach rates. The temperature of 80 deg. C was selected as representative of typical shutdown temperatures

A study on etching of UO2, Co, and Mo surface with R.F. plasma using CF4 and O2

International Nuclear Information System (INIS)

Kim, Yong Soo; Seo, Yong Dae

2003-01-01

Recently dry decontamination/surface-cleaning technology using plasma etching has been focused in the nuclear industry. In this study, the applicability of this new dry processing technique are experimentally investigated by examining the etching reaction of UO 2 , Co, and Mo in r.f. plasma with the etchant gas of CF 4 /O 2 mixture. UO 2 is chosen as a representing material for uranium and TRU (TRans-Uranic) compounds while metallic Co and Mo are selected because they are the principal contaminants in the used metallic nuclear components such as valves and pipes made of stainless steel or Inconel. Results show that in all cases maximum etching rate is achieved when the mole fraction of O 2 in CF 4 /O 2 mixture gas is 20%, regardless of temperature and r.f. power. In case of UO 2 , the highest etching reaction rate is greater than 1000 monolayers/min. at 370 .deg. C under 150 W r.f. power which is equivalent to 0.4 μm/min. As for Co, etching reaction begins to take place significantly when the temperature exceeds 350 .deg. C. Maximum etching rate achieved at 380 .deg. C is 0.06 μm/min. Mo etching reaction takes place vigorously even at relatively low temperature and the reaction rate increases drastically with increasing temperature. Highest etching rate at 380 .deg. C is 1.9 μm /min. According to OES (Optical Emission Spectroscopy) and AES (Auger Electron Spectroscopy) analysis, primary reaction seems to be a fluorination reaction, but carbonyl compound formation reaction may assist the dominant reaction, especially in case of Co and Mo. Through this basic study, the feasibility and the applicability of plasma decontamination technique are demonstrated

Sensitivity and uncertainty analysis for UO2 and MOX fueled PWR cells

International Nuclear Information System (INIS)

Foad, Basma; Takeda, Toshikazu

2015-01-01

Highlights: • A method for calculating sensitivity coefficients has been improved. • The IR approximation was used in order to get accurate results. • Sensitivities and uncertainties are calculated using the improved method. • The method is applied for UO 2 and MOX fueled PWR cells. • The verification was performed by comparing our results with MCNP6 and TSUNAMI-1D. - Abstract: This paper discusses the improvement of a method for calculating sensitivity coefficients of neutronics parameters relative to infinite dilution cross-sections because the conventional method neglects resonance self-shielding effect. In this study, the self-shielding effect is taken into account by using the intermediate resonance approximation in order to get accurate results in both high and low energy groups. The improved method is applied to calculate sensitivity coefficients and uncertainties of eigenvalue responses for UO 2 and MOX (ThO 2 –UO 2 and PuO 2 –UO 2 ) fueled pressurized water reactor cells. The verification of the improved method was performed by comparing the sensitivities with MCNP6 and TSUNAMI-1D. For uncertainty, calculation comparisons were done with TSUNAMI-1D, and we demonstrate that the differences are caused by the use of different covariance matrices

Thermodynamic properties and behaviour of A2[(UO2)(MoO4)2] compounds with A = Li, Na, K, Rb, and Cs

International Nuclear Information System (INIS)

Lelet, Maxim I.; Suleimanov, Evgeny V.; Golubev, Aleksey V.; Geiger, Charles A.; Depmeier, Wulf; Bosbach, Dirk; Alekseev, Evgeny V.

2014-01-01

Highlights: • Low temperature heat capacity of A 2 [(UO 2 )(MoO 4 ) 2 ] (A = Li, Na, K, Rb, and Cs) series was determined. • Enthalpy of formation of Li 2 [(UO 2 )(MoO 4 ) 2 ] was determined by HF solution calorimetry. • Δ f G° (T = 298 K) of all phases from studied series were calculated. - Abstract: A thermodynamic investigation of five alkali-metal uranyl molybdates of the general formula A 2 [(UO 2 )(MoO 4 ) 2 ], where A = Li, Na, K, Rb, and Cs, was undertaken. The various phases were synthesized by solid-state reaction of ANO 3, with A = Li, Na, K, Rb, or Cs, MoO 3 and γ-UO 3 . The synthetic products were characterized by X-ray powder diffraction and X-ray fluorescence methods. The low-temperature heat capacity, S r °, was measured using adiabatic calorimetry from T = (6 to 335) K. Based on these data, the third law entropy at T = 298.15 K, S°, is (345 ± 1) J · K −1 · mol −1 for Li 2 [(UO 2 )(MoO 4 ) 2 ], (373 ± 1) J · K −1 · mol −1 for Na 2 [(UO 2 )(MoO 4 ) 2 ], (390 ± 1) J · K −1 · mol −1 for K 2 [(UO 2 )(MoO 4 ) 2 ], (377 ± 1) J · K −1 · mol −1 for Rb 2 [(UO 2 )(MoO 4 ) 2 ] and (394 ± 1) J · K −1 · mol −1 for Cs 2 [(UO 2 )(MoO 4 ) 2 ]. The enthalpy of formation of Li 2 [(UO 2 )(MoO 4 ) 2 ] was determined using HF solution calorimetry giving Δ f H°(T = 298 K, Li 2 [(UO 2 )(MoO 4 ) 2 ], cr) = −(3456 ± 9) kJ · mol −1 . Using these new experimental results, together with literature data, the Gibbs free energy of formation of each compound was calculated, giving: Δ f G°(T = 298 K, Li 2 [(UO 2 )(MoO 4 ) 2 ], cr) = −(3204 ± 9) kJ · mol −1 , Δ f G°(T = 298 K, Na 2 [(UO 2 )(MoO 4 ) 2 ], cr) = −(3243 ± 2) kJ · mol −1 , Δ f G°(T = 298 K, K 2 [(UO 2 )(MoO 4 ) 2 ], cr) = −(3269 ± 3) kJ · mol −1 , Δ f G°(T = 298 K, Rb 2 [(UO 2 )(MoO 4 ) 2 ], cr) = −(3262 ± 3) kJ · mol −1 , and Δ f G°(T = 298 K, Cs 2 [(UO 2 )(MoO 4 ) 2 ], cr) = −(3259 ± 3) kJ · mol −1 . Smoothed S r °(T) values

Development of Innovative Accident Tolerant High Thermal Conductivity UO₂-Diamond Composite Fuel Pellets

Energy Technology Data Exchange (ETDEWEB)

Tulenko, James [Univ. of Florida, Gainesville, FL (United States); Subhash, Ghatu [Univ. of Florida, Gainesville, FL (United States)

2016-01-01

The University of Florida (UF) evaluated a composite fuel consisting of UO₂ powder mixed with diamond micro particles as a candidate as an accident-tolerant fuel (ATF). The research group had previous extensive experience researching with diamond micro particles as an addition to reactor coolant for improved plant thermal performance. The purpose of this research work was to utilize diamond micro particles to develop UO₂-Diamond composite fuel pellets with significantly enhanced thermal properties, beyond that already being measured in the previous UF research projects of UO₂ – SiC and UO₂ – Carbon Nanotube fuel pins. UF is proving with the current research results that the addition of diamond micro particles to UO₂ may greatly enhanced the thermal conductivity of the UO₂ pellets producing an accident-tolerant fuel. The Beginning of life benefits have been proven and fuel samples are being irradiated in the ATR reactor to confirm that the thermal conductivity improvements are still present under irradiation.

Determination of uranium content and its impurities in the AUC and UO2 powders

International Nuclear Information System (INIS)

Boybul; Arif Nugroho

2012-01-01

The analysis of uranium (U) content and its impurities in the ammonium uranyl carbonate (AUC) and uranium dioxide (UO 2 ) produced from research reactor fuel element production installation, PT. BATAN Teknologi have been carried out. Uranium content in the powders was analyzed by potentiometric titration methods and impurity contents was analyzed by atomic absorption spectrophotometer (AAS) and by inductively coupled plasma-atomic emission spectroscopy (ICP-AES). The purpose of this study was to determine of impurity elements in the AUC and UO 2 powder resulting from the production process if it meets the required specifications. It is reported that U content in the AUC is 48.62 wt% and that in the UO 2 is 88.08 wt%. The precision and accuracy analysis of the U content is 0,235% and 0,151%. In case of impurities in the AUC powders, it is reported that the analytical results of Zn, Ni, Cd, Co, Mn, Mg, Fe, Cu and Cr at 10.15 ppm, 1.12 ppm, not detection, not detection, not detection, 0.30 ppm, 216.07 ppm, not detection, and 31.36 ppm, respectively, while that UO 2 are 11.31 ppm, 72.14 ppm, not detection, not detection, 6.25 ppm, 8.65 ppm, 298.24 ppm, 12.75 ppm and 32, 23 ppm. The U and impurity contents in both the AUC and UO 2 fulfill the specification of nuclear fuel for RSG-GAS research reactor. (author)

Effect of titania addition on the thermal conductivity of UO2 fuel [Paper IIIB-C

International Nuclear Information System (INIS)

Sengupta, A.K.; Kumar, A.; Arora, K.B.S.; Pandey, V.D.; Nair, M.R.; Kamath, H.S.

1986-01-01

Pellet clad interaction in nuclear reactor fuel elements can be reduced by the use of higher grain size UO 2 fuel. This is achieved by the addition of dopant like titania, niobia etc. However, these dopants are considered as impurities which may affect the thermophysical and thermomechanical properties of the fuel. Thermal Conductivity which is one of the important properties controlling the inpile performance of the fuel has been measured for pure UO 2 and UO 2 containing 0.05wt per cent and 0.1wt per cent TiO 2 in the temperature range 900K to 1900K in vacuum. Thermal conductivity was obtained from thermal diffusivity data measured by laser flash method. The paper highlights the experimental results and discusses the effect of TiO 2 on the thermal conductivity of UO 2 fuel. (author)

Effect of titania addition on the thermal conductivity of UO2 fuel (Paper IIIB-C)

Energy Technology Data Exchange (ETDEWEB)

Sengupta, A K; Kumar, A; Arora, K B.S.; Pandey, V D; Nair, M R; Kamath, H S

1986-01-01

Pellet clad interaction in nuclear reactor fuel elements can be reduced by the use of higher grain size UO2 fuel. This is achieved by the addition of dopant like titania, niobia etc. However, these dopants are considered as impurities which may affect the thermophysical and thermomechanical properties of the fuel. Thermal Conductivity which is one of the important properties controlling the inpile performance of the fuel has been measured for pure UO2 and UO2 containing 0.05wt per cent and 0.1wt per cent TiO2 in the temperature range 900K to 1900K in vacuum. Thermal conductivity was obtained from thermal diffusivity data measured by laser flash method. The paper highlights the experimental results and discusses the effect of TiO2 on the thermal conductivity of UO2 fuel. 5 figures.

Experimental Determination of the Neutron Characteristics of UO{sub 2}-PuO{sub 2}-H{sub 2}O Lattices; Determination Experimentale Des Caracteristiques Neutroniques De Reseaux UO{sub 2}-PuO{sub 2}-H{sub 2}O

Energy Technology Data Exchange (ETDEWEB)

Debrue, J.; Fabry, A.; Leenders, L.; Motte, F.; Van Den Broeck, H. [Centre d' Etude de l' Energie Nucleaire, Mol (Belgium)

1967-09-15

As part of the investigation, in the VENUS test facility, of the variably moderated core of the BR3/VULCAIN reactor, a fuel assembly consisting of 37 UO{sub 2}-PuO{sub 2} pins (94% natural UO{sub 2}, 6% PuO{sub 2} ) was substituted for one of the enriched (to 7% {sup 235}U) UO{sub 2} fuel assemblies constituting the reactor core. Experiments were carried out with the object of refining the mathematical models for calculating the performance of this special assembly; inter alia, the fission density distribution and the changing ratio of the effective cross-sections for fission in the {sup 233}Pu and {sup 235}U were measured. Using the same critical facility, the authors are carrying out a critical experiment related directly to the problems of plutonium recycling in pressurized light-water thermal reactors. Three types of fuel are being used: UO{sub 2}-PuO{sub 2} with 3% {sup 235}U and 1% fissile plutonium, UO{sub 2}-PuO{sub 2} with 2% {sup 235}U and 2% fissile plutonium, and UO{sub 2} with 4% {sup 235}U. The two UO{sub 2}-PuO{sub 2} mixtures have completely different isotopic contents of {sup 240}Pu: 7% and 17%. In the first part of the experimental programme, a study is being made of regular lattices in cores having two co-axial cylindrical zones: a UO{sub 2}-PuO{sub 2} zone and a UO{sub 2} zone. Particular attention is being paid to investigating the region on either side of the interface separating the two zones, where the neutron spectrum reflects the characteristic energy distributions in each of the two lattices. The experimental results are to be used in calibrating the computational methods. In the second part of the experimental programme, parts of the core of the SENA power reactor will be simulated with a view to studying the problems of reloading one third of the core with mixed UO{sub 2}-PuO{sub 2} fuel. Among the experimental techniques employed in these various experiments emphasis is given to those most specifically related to the presence of

Study of UO2 radioinduced densification

International Nuclear Information System (INIS)

Stora, J.P.; Bruet, M.

1975-01-01

Measurements of radioinduced densification were performed on UO 2 DCN (intergranular fine porosity) and UO 2 DCI (interaggregate coarse porosity) in the Anemone device. The densification kinetics was followed by measuring the shrinkage of the oxide column on neutron radiographic plates. UO 2 DCI was found stable in regard to densification. At power near 450Wcm -1 , densification is hitten by restructuring phenomena [fr

Inspection of the UO2 special fuel for the prototype heavy water reactor 'FUGEN'

International Nuclear Information System (INIS)

Miura, Makoto; Ohmori, Takuro; Yoshino, Hiroyuki; Matsui, Hiromasa; Hirosawa, Naonori

1979-01-01

UO 2 special fuel assemblies are the fuel for material irradiation incorporating irradiation specimens, for the prototype heavy water reactor ''FUGEN''. In order to monitor the behavior of the pressure tube material irradiated with neutrons for long time, monitoring specimens were equipped in the core. This special fuel was fabricated by the Nuclear Fuel Industries, Ltd. (NFI), and the fuel cladding tubes, the capsule guide tubes and the capsule tubes were furnished by PNC. The irradiation specimens were prepared by PNC, and incorporated into the assemblies by NFI. The inspection by PNC on the special fuel assemblies was conducted following the inspection by the maker, which was made on UO 2 pellets, fuel element and assembly parts except cladding tubes, after completing the fabrication. The specifications of the special fuel, especially for the outer and inner layer pellets, the outer and inner layer fuel elements and the fuel assemblies, are presented. The flow sheet for the inspection process and surveillance test of special fuel assemblies is illustrated. The inspection items, the materials and the quantity inspection are tabulated for the fuel elements, the fuel assemblies and the irradiation capsules, respectively. The structure of a special type fuel assembly is shown. For each inspection, the inspection methods and items and the results are explained. As for the results of inspection of the special fuel, the UO 2 pellets, fuel element parts, fuel elements, fuel assembly parts, fuel assemblies, capsules and irradiation specimens were in accordance with the specifications. Regarding the situation of the quality control in the processes, check was made with many documents, and it was recognized that the quality control was performed in the quality assurance program. (Nakai, Y.)

Fission and explosive energy releases of PuO2, PuO2--UO2, UO2, and UO3 assemblies

International Nuclear Information System (INIS)

Koelling, J.J.; Hansen, G.E.; Byers, C.C.

1977-01-01

The critical masses and fission and explosive energy releases of PuO 2 , PuO 2 --UO 2 , UO 2 , and UO 3 assemblies have been calculated. The parameters selected for the model are conservative. They were chosen after review of appropriate plants that have been and are proposed for construction in the future. The resulting data envelopes are intended to include any conceivable set of circumstances that could ultimately lead to a nuclear incident. All energy release analysis was performed for initial fission spikes only: recriticality mechanisms were not considered

Recent Results from MINERvA

Energy Technology Data Exchange (ETDEWEB)

Fields, Laura [Fermilab

2016-12-21

The MINERvA collaboration is currently engaged in a broad program of neutrino-nucleus interaction measurements. Several recent measurements of interest to the accelerator-based oscillation community are presented. These include measurements of quasi-elastic scattering, diffractive pion production, kaon production and comparisons of interaction cross sections across nuclei. A new measurement of the NuMI neutrino beam flux that incorporates both external hadro-production data and MINERvA detector data is also presented.

Recent Results from MINERvA

Energy Technology Data Exchange (ETDEWEB)

Miller, Jonathan [Santa Maria U., Valparaiso

2014-09-01

MINERvA (Main INjector ExpeRiment for v-A) is a few-GeV neutrino nucleus scattering experiment at Fermilab using various nuclei as targets. The experiment provides measurements of neutrino and anti-neutrino cross sections off of nuclear targets which are important for neutrino oscillation experiments and the probing of the nuclear medium.Presented are recent results from MINERvA on quasi-elastic, inclusive charged-current neutrino scattering, and pion production processes.

AGR-2 irradiation test final as-run report, Rev. 1

International Nuclear Information System (INIS)

2014-01-01

This document presents the as-run analysis of the AGR-2 irradiation experiment. AGR-2 is the second of the planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the U.S. Department of Energy as part of the Very High Temperature Reactor (VHTR) Technical Development Office (TDO) program. The objectives of the AGR-2 experiment are to: (a) Irradiate UCO (uranium oxycarbide) and UO 2 (uranium dioxide) fuel produced in a large coater. Fuel attributes are based on results obtained from the AGR-1 test and other project activities; (b) Provide irradiated fuel samples for post-irradiation experiment (PIE) and safety testing; and, (c) Support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. The primary objective of the test was to irradiate both UCO and UO 2 TRISO (tri-structural isotropic) fuel produced from prototypic scale equipment to obtain normal operation and accident condition fuel performance data. The UCO compacts were subjected to a range of burnups and temperatures typical of anticipated prismatic reactor service conditions in three capsules. The test train also includes compacts containing UO 2 particles produced independently by the United States, South Africa, and France in three separate capsules. The range of burnups and temperatures in these capsules were typical of anticipated pebble bed reactor service conditions. The results discussed in this report pertain only to U.S. produced fuel. In order to achieve the test objectives, the AGR-2 experiment was irradiated in the B-12 position of the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) for a total irradiation duration of 559.2 effective full power days (EFPD). Irradiation began on June 22, 2010, and ended on October 16, 2013, spanning 12 ATR power cycles and approximately three and a half calendar years. The

Hydrothermal synthesis, structure, and catalytic properties of UO2Sb2O4

International Nuclear Information System (INIS)

Sykora, Richard E.; King, Joseph E.; Illies, Andreas J.; Albrecht-Schmitt, Thomas E.

2004-01-01

A new uranyl antimonite, UO 2 Sb 2 O 4 (1), has been prepared from the hydrothermal reaction of UO 3 with Sb 2 O 3 and KCl. The structure of 1 consists of neutral two-dimensional ∞ 2 [UO 2 Sb 2 O 4 ] layers. The U(VI) centers are ligated by two trans oxo ligands and four square pyramidal antimonite anions. In addition, the U(VI) also forms long contacts with two additional oxygen atoms that are distorted by 12.7(2) degree sign out of the equatorial plane perpendicular to the uranyl unit. These long interactions are significant owing to evidence supplied by bond valence sum calculations. The two-dimensional layers found in 1 are built from one-dimensional chains formed from edge-sharing UO 6 octahedra that run along the b-axis, and are linked together by [Sb 2 O 4 ] 2- chains. A flow microreactor system has been used to study the catalytic activity of 1, and these results show that it can be used as a catalyst in the conversion of propene and O 2 to acrolein. Crystallographic data: 1, monoclinic, space group C2/m, a=13.490(2) A, b=4.0034(6) A, c=5.1419(8) A, β=104.165(3) deg., Z=2, MoKα, λ=0.71073, R(F)=1.74% for 30 parameters with 365 reflections with I>2σ(I)

Thermal conductivity thermal diffusivity of UO{sub 2}-BeO nuclear fuel pellets

Energy Technology Data Exchange (ETDEWEB)

Mansur, Fábio A.; Camarano, Denise M.; Santos, Ana M. M.; Ferraz, Wilmar B.; Silva, Mayra A.; Ferreira, Ricardo A.N., E-mail: fam@cdtn.br, E-mail: dmc@cdtn.br, E-mail: amms@cdtn.br, E-mail: ferrazw@cdtn.br, E-mail: mayra.silva@cdtn.br, E-mail: ricardoanf@yahoo.com.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2017-07-01

The temperature distribution in nuclear fuel pellets is of vital importance for the performance of the reactor, as it affects the heat transfer, the mechanical behavior and the release of fission gas during irradiation, reducing safety margins in possible accident scenarios. One of the main limitation for the current uranium dioxide nuclear fuel (UO{sub 2}) is its low thermal conductivity, responsible for the higher temperature of the pellet center and, consequently, for a higher radial temperature gradient. Thus, the addition of another material to increase the UO{sub 2} fuel thermal conductivity has been considered. Among the additives that are being investigated, beryllium oxide (BeO) has been chosen due to its high thermal conductivity, with potential to optimize power generation in pressurized light water reactors (PWR). In this work, UO{sub 2}-BeO pellets were obtained by the physical mixing of the powders with additions of 2wt% and 3wt% of BeO. The thermal diffusivity and conductivity of the pellets were determined from room temperature up to 500 °C. The results were normalized to 95% of the theoretical density (TD) of the pellets and varied according to the BeO content. The range of the values of thermal diffusivity and conductivity were 1.22 mm{sup 2}∙s{sup -1} to 3.69 mm{sup 2}∙s{sup -1} and 3.80 W∙m{sup -}'1∙K{sup -1} to 9.36 W∙m{sup -1}∙K{sup -1}, respectively. (author)

Possible effects of UO2 oxidation on light water reactor spent fuel performance in long-term geologic disposal

International Nuclear Information System (INIS)

Almassy, M.Y.; Woodley, R.E.

1982-08-01

Disposal of spent nuclear fuel in a conventionally mined geologic formation is the nearest-term option for permanently isolating radionuclides from the biosphere. Because irradiated uranium dioxide (UO 2 ) fuel pellets retain 95 to 99% of the radionuclides generated during normal light water reactor operation, they may represent a significant barrier to radionuclide release. This document presents a technical assessment of published literature representing the current level of understanding of spent fuel characteristics and conditions that may degrade pellet integrity during a geologic disposal sequence. A significant deterioration mechanism is spent UO 2 oxidation with possible consequences identified as fission gas release, rod diameter increases, cladding breach extension, and release of solid fuel particles containing radionuclides. Areas requiring further study to support development of a comprehensive spent fuel performance prediction model are highlighted. A program and preliminary schedule to obtain the information needed to develop model correlations are also presented

Thermal Shock Tests on UO{sub 2} Small Spheres; Essais de choc thermique sur des elements spheriques de UO{sub 2}; Ispytaniya nebol'shikh sharikov iz UO{sub 2} teplovykh udarom; Ensayo de pequenas esferas de UO{sub 2} por choque.termico

Energy Technology Data Exchange (ETDEWEB)

Perona, G.; Brutto, E.; Galbusera, U.; Palladino, G.; Sesini, R. [Centro Informazioni Studi Esperienze, Milan (Italy)

1963-11-15

If UO{sub 2} small spheres are used as fuel in a reactor in contact with the cooler, it is necessary to know the maximum value of the thermal stress, due to the work conditions in the reactor, which the small spheres are able to withstand without breaking. These conditions can be calculated if the physical properties of the material are known. Owing to the considerable number of properties involved, and in consideration of the uncertainty which always exists in each of them, it is preferable to test directly the spheres, submitting them to the same kind of stresses that they undergo in thereactor. In this work a thermal shock method for the small spheres has been studied, while conditions are indicated in which this method can reproduce stress conditions directly comparable with those existing in the reactor. As for small spheres, the difficulty consists in producing coolings with very high values of the coefficient of surface heat transfer. The experimental methods are described and the results obtained are indicated. The application of this method seems to be very interesting particularly in the field of the technological research for improving the characteristics of the UO{sub 2} small spheres by means of additives. In fact it allows the control of the total interesting effect with a single measurement. (author) [French] Si l'on veut utiliser comme combustible dans un reacteur des elements spheriques de UO{sub 2} en contact avec le refroidisseur, il faut au prealable determiner la valeur maximum de la contrainte thermique - due aux conditions regnant dans le reacteura laquelle les elements sont capables de resister sans se fissurer. Il est possible de calculer ces conditions si l'on connait les proprietes physiques du materiau utilise. En raison du nombre important des proprietes a prendre en consideration, et compte tenu de l'incertitude qui existe toujours pour chacune d'elles, il est preferable de faire des essais thermiques en soumettant directement les

Characterization of UO{sub 2}, a) Characterization of UO{sub 2} powder; b) Investigation of U-O system by DDK and TGA methods; Karakterizacija UO{sub 2}, a) Karakterizacija praha UO{sub 2}; b) Ispitivanje sistema U-O metodama DDK i TGA

Energy Technology Data Exchange (ETDEWEB)

Ristic, M M [Institute of Nuclear Sciences Vinca, Laboratorija za reaktorske materijale, Beograd (Serbia and Montenegro)

1962-10-15

The objectives of the study of U-O powder system were: detailed characterization of the UO{sub 2} powder which will be used for studying the sintering process, and more detailed properties of the U-O system (thermodynamic aspects of oxidation kinetics). Study of the physical and chemical properties of UO{sub 2} powder were performed and then oxidation kinetics of UO{sub 2} {yields}U{sub 3}O{sub 7} was investigated. Detailed qualitative DDK analysis was done. Owing to the TGA equipment there was a possibility to obtain U{sub 3}O{sub 7} study of U{sub 3}O{sub 7} {yields} U{sub 3}O{sub 8} oxidation was possible.

Investigation of very high burnup UO{sub 2} fuels in Light Water Reactors

Energy Technology Data Exchange (ETDEWEB)

Cappia, Fabiola

2017-03-27

Historically, the average discharge burnup of Light Water Reactor (LWR) fuel has increased almost continuously. On one side, increase in the average discharge burnup is attractive because it contributes to decrease part of the fuel cycle costs. On the other side, it raises the practical problem of predicting the performance, longevity and properties of reactor fuel elements upon accumulation of irradiation damage and fission products both during in-reactor operation and after discharge. Performance of the fuel and structural components of the core is one of the critical areas on which the economic viability and public acceptance of nuclear energy production hinges. Along the pellet radius, the fuel matrix is subjected to extremely heterogeneous alteration and damage, as a result of temperature and burnup gradients. In particular, in the peripheral region of LWR UO{sub 2} fuel pellets, when the local burnup exceeds 50-70 GWd/tHM, a microstructural transformation starts to take place, as a consequence of enhanced accumulation of radiation damage, fission products and limited thermal recovery. The newly formed structure is commonly named High Burnup Structure (HBS). The HBS is characterised by three main features: (a) formation of submicrometric grains from the original grains, (b) depletion of fission gas from the fuel matrix, (c) steep increase in the porosity, which retains most of the gas depleted from the fuel matrix. The last two aspects rose significant attention because of the important impact of the fission gas behaviour on integral fuel performance. The porosity increase controls the gas-driven swelling, worsening the cladding loading once the fuel-cladding gap is closed. Another concern is that the large retention of fission gas within the HBS could lead to significant release at high burnups through the degradation of thermal conductivity or contribute to fuel pulverisation during accidental conditions. Need of more experimental investigations about the

Effect of alpha irradiation on UO2 surface reactivity in aqueous media

International Nuclear Information System (INIS)

Jegou, C.; Muzeau, B.; Broudic, V.; Poulesquen, A.; Roudil, D.; Jorion, F.; Corbel, C.

2005-01-01

The option of direct disposal of spent nuclear fuel in a deep geological formation raises the need to investigate the long-term behavior of the UO 2 matrix in aqueous media subjected to α-β-γ radiation. The β-γ emitters account for most of the activity of spent fuel at the moment it is removed from the reactor, but diminish within a millennial time frame by over three orders of magnitude to less than the long-term activity. The latter persists over much longer time periods and must therefore be taken into account over a geological disposal time scale. Leaching experiments with solution renewal were carried out on UO 2 pellets doped with alpha emitters ( 238 Pu and 239 Pu) to quantify the impact of alpha irradiation on UO 2 matrix alteration. Three batches of doped UO 2 pellets with different alpha flux levels (3.30 x 10 4 , 3.30 x 10 5 , and 3.2 x 10 6 α cm -2 s -1 ) were studied. The results obtained in aerated and deaerated media immediately after sample annealing or interim storage in air provide a better understanding of the UO 2 matrix alteration mechanisms under alpha irradiation. Interim storage in air of UO 2 pellets doped with alpha emitters results in variations of the UO 2 surface reactivity, which depends on the alpha particle flux at the interface and on the interim storage duration. The variation in the surface reactivity and the greater uranium release following interim storage cannot be attributed to the effect of alpha radiolysis in aerated media since the uranium release tends toward the same value after several leaching cycles for the doped UO 2 pellet batches and spent fuel. Oxygen diffusion enhanced by alpha irradiation of the extreme surface layer and/or radiolysis of the air could account for the oxidation of the surface UO 2 to UO 2+x . However, leaching experiments performed in deaerated media after annealing the samples and preleaching the surface suggest that alpha radiolysis does indeed affect the dissolution, which varies with the

Sintering of nonstoichiometric UO2

International Nuclear Information System (INIS)

Susnik, D.; Holc, J.

1983-01-01

Activated sintering of UO 2 pellets at 1100 deg C is described. In CO 2 atmosphere is UO 2 is nonstoichiometric and pellets from active UO 2 powders sinter at 900 deg C to high density. At 1100 deg C the final sintered density is practically achieved at heating on sintering temperature. After reduction and cooling in H 2 atmosphere which is followed sintering in CO 2 the structure is identical to the structured UO 2 pellets sintered at high temperature in H 2 . Density of activated sintered UO 2 pellets is stable, even after additional sintering at 1800 deg C. (author)

Fission gas release from the sintered UO{sub 2} fuel; Oslobadjanje fisionih gasova iz goriva od sinterovanog UO{sub 2}

Energy Technology Data Exchange (ETDEWEB)

Sigulinski, F; Stevanovic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

1966-11-15

This paper shoes the phenomena which control fission gases release from the sintered UO{sub 2} dependent of the burnup rate: ejection, release, diffusion, increased fission gas accumulation causing structural changes in the fuel. release of fission gases from the fuel for power reactors was studied as well. The influence of factors as temperature, characteristics of fuel, burnup rate and burnup level was analyzed. Prikazani su mehanizmi koji kontrolisu izdvajanje fisionih gasova iz sinterovanog UO{sub 2} pri razlicitim brzinama izgaranja: izletanje, izbijanje, difuzija, povecano izdvajanje fisionih gasova koje prati strukturne promene u gorivu. Razmatrano je proucavanje izdvajanja fisionih gasova iz goriva za reaktore snage. Analiziran je uticaj faktora kao sto su temperatura, karakteristike goriva, brzina i stepen izgaranja (author)

Interesting Developments in UO{sub 2} Technology; Progres interessants dans la technologie du bioxyde d'uranium; Interesnye usovershenstvovaniya tekhnologii UO{sub 2}; Recientes progresos en la tecnologia del UO{sub 2}

Energy Technology Data Exchange (ETDEWEB)

Robertson, J. A.L. [Atomic Energy of Canada Ltd., Chalk River, Ontario (Canada)

1963-11-15

Now that several UO{sub 2}-fuelled reactors are operating routinely, good irradiation performance of UO{sub 2} is taken for granted. It is therefore stimulating to find that significant developments are still occurring. Most exciting was the recent discovery by Battelle Memorial Institute workers that a particular single crystal of UO{sub 2} had a very high thermal conductivity at elevated temperatures. Following controversy over the matter, an irradiation at Chalk River demonstrated that the large grains formed in operating fuel elements do not necessarily exhibit this enhanced conductivity. Our laboratory experiments have shown that the enhancement is only present in hypostoichiometric compositions and depends little, if any, on the absence of grain boundaries. Indeed, the high conductivity can be obtained in polycrystalline sinters by controlling the stoichiometry. It has long been known that sheath elongation could be reduced by fabricating the UO{sub 2} pellets with depressions in their end faces. Later it was shown that movement of the fuel into a void at the end of the pellet stack was impeded by diametral expansion of the fuel and its mechanical interaction with the sheath. The biggest advance in minimizing sheath distensions has been the realization that longitudinal and diametral expansions are interrelated through the volume expansion of the fuel whose hot core is appreciably plastic. Our empirical knowledge of the factors determining the release of fission-product gases from UO{sub 2} has improved. In particular, increasing the irradiation exposure from 10{sup 15} to 10{sup 18} fissions/cm{sup 3} can reduce the apparent diffusion rates for xenon in UO{sub 2} during subsequent anneals by a factor of 10{sup 3}. The gas is probably immobilized in minute traps, some existing in the original material and some generated by irradiation damage. Detailed analysis indicated slow escape from the traps, presumably from the finite solubility of the xenon in UO{sub 2

MINERvA neutrino detector response measured with test beam data

International Nuclear Information System (INIS)

Aliaga, L.; Altinok, O.; Araujo Del Castillo, C.; Bagby, L.; Bellantoni, L.; Bergan, W.F.; Bodek, A.; Bradford, R.; Bravar, A.; Budd, H.; Butkevich, A.; Martinez Caicedo, D.A.; Carneiro, M.F.; Christy, M.E.; Chvojka, J.; Motta, H. da; Devan, J.

2015-01-01

The MINERvA collaboration operated a scaled-down replica of thesolid scintillator tracking and sampling calorimeter regions of the MINERvA detector in a hadron test beam at the Fermilab Test Beam Facility. This paper reports measurements with samples of protons, pions, and electrons from 0.35 to 2.0 GeV/c momentum. The calorimetric response to protons, pions, and electrons is obtained from these data. A measurement of the parameter in Birks' law and an estimate of the tracking efficiency are extracted from the proton sample. Overall the data are well described by a Geant4-based Monte Carlo simulation of the detector and particle interactions with agreements better than 4% for the calorimetric response, though some features of the data are not precisely modeled. These measurements are used to tune the MINERvA detector simulation and evaluate systematic uncertainties in support of the MINERvA neutrino cross-section measurement program

MINERvA neutrino detector response measured with test beam data

Energy Technology Data Exchange (ETDEWEB)

Aliaga, L. [Department of Physics, College of William & Mary, Williamsburg, VA 23187 (United States); Altinok, O. [Physics Department, Tufts University, Medford, MA 02155 (United States); Araujo Del Castillo, C. [Sección Física, Departamento de Ciencias, Pontificia Universidad Católica del Perú, Apartado 1761, Lima (Peru); Bagby, L.; Bellantoni, L. [Fermi National Accelerator Laboratory, Batavia, IL 60510 (United States); Bergan, W.F. [Department of Physics, College of William & Mary, Williamsburg, VA 23187 (United States); Bodek, A.; Bradford, R. [University of Rochester, Rochester, New York 14627 (United States); Bravar, A. [University of Geneva, 1211 Geneva 4 (Switzerland); Budd, H. [University of Rochester, Rochester, New York 14627 (United States); Butkevich, A. [Institute for Nuclear Research of the Russian Academy of Sciences, 117312 Moscow (Russian Federation); Martinez Caicedo, D.A. [Centro Brasileiro de Pesquisas Físicas, Rua Dr. Xavier Sigaud 150, Urca, Rio de Janeiro, Rio de Janeiro 22290-180 (Brazil); Fermi National Accelerator Laboratory, Batavia, IL 60510 (United States); Carneiro, M.F. [Centro Brasileiro de Pesquisas Físicas, Rua Dr. Xavier Sigaud 150, Urca, Rio de Janeiro, Rio de Janeiro 22290-180 (Brazil); Christy, M.E. [Hampton University, Department of Physics, Hampton, VA 23668 (United States); Chvojka, J. [University of Rochester, Rochester, New York 14627 (United States); Motta, H. da [Centro Brasileiro de Pesquisas Físicas, Rua Dr. Xavier Sigaud 150, Urca, Rio de Janeiro, Rio de Janeiro 22290-180 (Brazil); Devan, J. [Department of Physics, College of William & Mary, Williamsburg, VA 23187 (United States); and others

2015-07-21

The MINERvA collaboration operated a scaled-down replica of thesolid scintillator tracking and sampling calorimeter regions of the MINERvA detector in a hadron test beam at the Fermilab Test Beam Facility. This paper reports measurements with samples of protons, pions, and electrons from 0.35 to 2.0 GeV/c momentum. The calorimetric response to protons, pions, and electrons is obtained from these data. A measurement of the parameter in Birks' law and an estimate of the tracking efficiency are extracted from the proton sample. Overall the data are well described by a Geant4-based Monte Carlo simulation of the detector and particle interactions with agreements better than 4% for the calorimetric response, though some features of the data are not precisely modeled. These measurements are used to tune the MINERvA detector simulation and evaluate systematic uncertainties in support of the MINERvA neutrino cross-section measurement program.

Safe dismantling of the SVAFO research reactors R2 and R2-0 in Sweden

International Nuclear Information System (INIS)

ARNOLD, Hans-Uwe; BROY, Yvonne; Dirk Schneider

2017-01-01

The R2 and R2-0 reactors were part of the Swedish government's research program on nuclear power from the early 1960's. Both reactors were shut down in 2005 following a decision by former operator Studsvik Nuclear AB. The decommissioning of the R2 and R2-0 reactors is divided into three phases. The first phase - awarded to AREVA - involved dismantling of the reactors and associated systems in the reactor pool, treatment of the disassembled components as well as draining, cleaning and emptying the pool. In the second phase, the pool structure itself will be dismantled, while removal of remaining reactor systems, treatment and disposal of materials and clean-up will be carried out in the third stage. The entire work is planned to be completed before the end of this decade. The paper describes the several steps of phase 1 - starting with the team building, followed by the dismantling operations and covers challenges encountered and lessons learned as well. The reactors consist of 5.400 kg aluminum, 6.000 kg stainless steel restraint structures as well as, connection elements of the mostly flanged components (1.000 kg). The most demanding - from a radiological point of view - was the R2-0 reactor that was limited to ∼ 1 m"3 construction volumes but with an extremely heterogeneous activation profile. Based on the calculated radiological entrance data and later sampling, nuclide vectors for both reactors depending on the real placement of the single component and on the material (aluminum and stainless steel) were created. Finally, for the highest activated component from R2 reactor, 85 Sv/h were measured. The dismantling principles - adopted on a safety point of view - were the following: The always protected base area of the ponds served as a flexible buffer area for waste components and packaging. Specific protections were also installed on the walls to protect them from mechanical stress which may occur during dismantling work. A specific work platform was

Pile noise experiment in MINERVE reactor to estimate kinetic parameters using various data processing methods

Energy Technology Data Exchange (ETDEWEB)

Geslot, Benoit; Gruel, Adrien; Pepino, Alexandra; Di Salvo, Jacques; Izarra, Gregoire de; Jammes, Christian; Destouches, Christophe; Blaise, Patrick [CEA, DEN, DER/SPEx, Cadarache, F-13108 St Paul Lez Durance (France)

2015-07-01

MINERVE is a two-zone pool type zero power reactor operated by CEA (Cadarache, France). Kinetic parameters of the core (prompt neutron decay constant, delayed neutron fraction, generation time) have been recently measured using various pile noise experimental techniques, namely Feynman-α, Rossi-α and Cohn-α. Results are discussed and compared to each other's. The measurement campaign has been conducted in the framework of a tri-partite collaboration between CEA, SCK.CEN and PSI. Results presented in this paper were obtained thanks to a time-stamping acquisition system developed by CEA. PSI performed simultaneous measurements which are presented in a companion paper. Signals come from two high efficiency fission chambers located in the graphite reflector next to the core driver zone. Experiments were conducted at critical state with a reactor power of 0.2 W. The core integral fission rate is obtained from a calibrated miniature fission chamber located at the center of the core. Other results obtained in two sub-critical configurations will be presented elsewhere. Best estimate delayed neutron fraction comes from the Cohn-α method: 747 ± 15 pcm (1σ). In this case, the prompt decay constant is 79 ± 0.5 s{sup -1} and the generation time is 94.5 ± 0.7 μs. Other methods give consistent results within the confidence intervals. Experimental results are compared to calculated values obtained from a full 3D core modeling with the CEA-developed Monte Carlo code TRIPOLI4.9 associated with its continuous energy JEFF3.1.1-based library. A very good agreement is observed for the calculated delayed neutron fraction (748.7 ± 0.4 pcm at 1σ), that is a difference of -0.3% with the experiment. On the contrary, a 10% discrepancy is observed for the calculated generation time (104.4 ± 0.1 μs at 1σ). (authors)

A Study of the Temperature Distribution in UO2 Reactor Fuel Elements

International Nuclear Information System (INIS)

Devold, I.

1968-05-01

Thermal conductivity is one of the most important properties of nuclear reactor fuels. Accurate knowledge of this property is vital because, among other things, it determines the maximum power that can be taken out of the fuel element per unit length of the material without exceeding the safety limits of the fuel elements. This report consists of a study of the thermal behaviour of uranium dioxide in the form of reactor fuel. The experimental part of the report describes measurements performed at the OECD Halden Reactor Project, Halden, Norway. The experiment was originally set up in order to measure the temperature at the center of a UO 2 fuel element as a function of element power, in order to determine the safe operation limit of the fuel assembly. However, in analysing the data obtained, very interesting thermal conductivity values were obtained and comparison with existing correlations could be performed. This comparison shows that a certain agreement is obtained between the measured data at Halden and a theory published by J.L. Bates in 1961, which predicts an increase in the thermal conductivity above 1500 deg C. The data obtained below 1300 deg C are also in good agreement with measurements performed by Vogt, Grandell and Runfors in 1964. The report contains a mathematical description of the heat transfer mechanisms in cylindrical fuel elements. The model is coded in FORTRAN IV-code and referred to as FUELTEMP

TRX and UO2 criticality benchmarks with SAM-CE

International Nuclear Information System (INIS)

Beer, M.; Troubetzkoy, E.S.; Lichtenstein, H.; Rose, P.F.

1980-01-01

A set of thermal reactor benchmark calculations with SAM-CE which have been conducted at both MAGI and at BNL are described. Their purpose was both validation of the SAM-CE reactor eigenvalue capability developed by MAGI and a substantial contribution to the data testing of both ENDF/B-IV and ENDF/B-V libraries. This experience also resulted in increased calculational efficiency of the code and an example is given. The benchmark analysis included the TRX-1 infinite cell using both ENDF/B-IV and ENDF/B-V cross section sets and calculations using ENDF/B-IV of the TRX-1 full core and TRX-2 cell. BAPL-UO2-1 calculations were conducted for the cell using both ENDF/B-IV and ENDF/B-V and for the full core with ENDF/B-V

First identification and thermodynamic characterization of the ternary U(VI) species, UO2(O2)(CO3)2(4-), in UO2-H2O2-K2CO3 solutions.

Science.gov (United States)

Goff, George S; Brodnax, Lia F; Cisneros, Michael R; Peper, Shane M; Field, Stephanie E; Scott, Brian L; Runde, Wolfgang H

2008-03-17

In alkaline carbonate solutions, hydrogen peroxide can selectively replace one of the carbonate ligands in UO2(CO3)3(4-) to form the ternary mixed U(VI) peroxo-carbonato species UO2(O2)(CO3)2(4-). Orange rectangular plates of K4[UO2(CO3)2(O2)].H2O were isolated and characterized by single crystal X-ray diffraction studies. Crystallographic data: monoclinic, space group P2(1)/ n, a = 6.9670(14) A, b = 9.2158(10) A, c = 18.052(4) A, Z = 4. Spectrophotometric titrations with H 2O 2 were performed in 0.5 M K 2CO 3, with UO2(O2)(CO3)2(4-) concentrations ranging from 0.1 to 0.55 mM. The molar absorptivities (M(-1) cm(-1)) for UO2(CO3)3(4-) and UO2(O2)(CO3)2(4-) were determined to be 23.3 +/- 0.3 at 448.5 nm and 1022.7 +/- 19.0 at 347.5 nm, respectively. Stoichiometric analyses coupled with spectroscopic comparisons between solution and solid state indicate that the stable solution species is UO2(O2)(CO3)2(4-), which has an apparent formation constant of log K' = 4.70 +/- 0.02 relative to the tris-carbonato complex.

Oxidation of UO2 at 150 to 3500C

International Nuclear Information System (INIS)

Gilbert, E.R.; White, G.D.; Knox, C.A.

1985-02-01

Tests were performed on nonirradiated UO 2 pellets from 150 to 350 0 C in atmospheric air and controlled environments and on spent light-water reactor (LWR) fuel fragments at 200 and 230 0 C in atmospheric air to determine the variables that affect oxidation behavior under dry storage conditions. The weight of spent fragments increased 50 to 100 times faster than the weight of nonirradiated UO 2 pellets at 230 0 C. Non-irradiated pellet fragments gained weight 5 to 7 times faster than nonirradiated pellets. The fragments simulated fuel fragmented by thermal gradients during reactor power changes. Low-density powder (U 3 O 8 ) formed at 0.05 and 0.3% weight gain for nonirradiated pellets and fragments, respectively, but had not formed at 3% weight gain for spent fuel fragments with a burnup of 29,000 MWd/MTU. Canadian investigators had found that powder formed at intermediate levels of weight gain in CANDU spent fuel fragments with an approximate burnup of 8000 MWd/MTU. The combined effects of the high rate of weight gain in spent fuel and the burnup dependence of weight gain at powder formation resulted in a minimum in a plot of the time for the onset of powder formation versus burnup. The minimum in powder induction time occurs at or below burnup levels typical of CANDU spent fuel and spent fuel at the ends of some LWR rods. The results are described in terms of thermal and neutron irradiation-induced changes in UO 2 pellet structure and chemical composition. Other tests were performed at up to 275 0 C with spent fuel fragments and nonirradiated UO 2 pellets in moist nitrogen to determine the suitability of nitrogen as a cover gas. No measurable weight gain or visible physical changes occurred during the first 2 months of testing. 22 figures, 7 tables

Enhancement of actinide incineration and transmutation rates in Ads EAP-80 reactor core with MOX PuO2 and UO2 fuel

International Nuclear Information System (INIS)

Kaltcheva-Kouzminava, S.; Kuzminov, V.; Vecchi, M.

2001-01-01

Neutronics calculations of the accelerator driven reactor core EAP-80 with UO 2 and PuO 2 MOX fuel elements and Pb-Bi coolant are presented in this paper. Monte Carlo optimisation computations of several schemes of the EAP-80 core with different types of fuel assemblies containing burnable absorber B4 C or H 2 Zr zirconium hydride moderator were performed with the purpose to enhance the plutonium and actinide incineration rate. In the first scheme the reactor core contains burnable absorber B4 C arranged in the cladding of fuel elements with high enrichment of plutonium (up to 45%). In the second scheme H2 Zr zirconium hydride moderated zones were located in fuel elements with low enrichment (∼20%). In both schemes the incineration rate of plutonium is about two times higher than in the reference EAP-80 core and at the same time the power density distribution remains significantly unchanged compared to the reference core. A hybrid core containing two fuel zones one of which is the inner fuel region with UO 2 and PuO 2 high enrichment plutonium fuel and the second one is the outer region with fuel elements containing zirconium hydride layer was also considered. Evolution of neutronics parameters and actinide transmutation rates during the fuel burn-up is presented. Calculations were performed using the MCNP-4B code and the SCALE 4.3 computational system. (author)

High-temperature irradiation of niobium-1 w/o zirconium-clad UO/sub 2/. [Compatibility with lithium

Energy Technology Data Exchange (ETDEWEB)

Kangilaski, M.; Fromm, E.O.; Lozier, D.H.; Storhok, V.W.; Gates, J.E.

1965-06-28

Twenty-four 0.225-in.-diameter and six 0.290-in.-diameter UO/sub 2/ specimens clad with 80 mils of niobium-1 w/o zirconium were irradiated to burnups of 1.4 to 6.0 at. % of uranium at surface temperatures of 900 to 1400/sup 0/C. UO/sub 2/ and lithium were found to be incompatible at these temperatures, and the thick cladding was used primarily to minimize the chances of contact of UO/sub 2/ and the lithium coolant. The thickly clad specimens did not undergo any dimensional changes as a result of irradiation, although it was found that movement of UO/sub 2/ took place in the axial direction by a vaporization-redeposition mechanism. It was found that 32 to 87% of the fission gases was released from the fuel, depending on the temperature of the specimen. Metallographic examination of longitudinal and transverse sections of the specimens indicated the usual UO/sub 2/ microstructure with columnar grains. Grain-boundary thickening was observed in the UO/sub 2/ at higher burnups. The oxygen/uranium ratio of UO/sub 2/ increased with increasing burnup.

The uranium(VI) oxoazides [UO{sub 2}(N{sub 3}){sub 2}.CH{sub 3}CN], [(bipy){sub 2}(UO{sub 2}){sub 2}(N{sub 3}){sub 4}], [(bipy)UO{sub 2}(N{sub 3}){sub 3}]{sup -}, [UO{sub 2}(N{sub 3}){sub 4}]{sup 2-}, and [(UO{sub 2}){sub 2}(N{sub 3}){sub 8}]{sup 4-}

Energy Technology Data Exchange (ETDEWEB)

Haiges, Ralf; Christe, Karl O. [Loker Hydrocarbon Research Institute and Department of Chemistry, University of Southern California, Los Angeles, CA (United States); Vasiliu, Monica; Dixon, David A. [Department of Chemistry, The University of Alabama, Tuscaloosa, AL (United States)

2017-01-12

The reaction between [UO{sub 2}F{sub 2}] and an excess of Me{sub 3}SiN{sub 3} in acetonitrile solution results in fluoride-azide exchange and the uranium(VI) dioxodiazide adduct [UO{sub 2}(N{sub 3}){sub 2}.CH{sub 3}CN] was isolated in quantitative yield. The subsequent reaction of [UO{sub 2}(N{sub 3}){sub 2}.CH{sub 3}CN] with 2,2{sup '}-bipyridine (bipy) resulted in the formation of the azido-bridged binuclear complex [(bipy){sub 2}(UO{sub 2}){sub 2}(N{sub 3}){sub 4}]. The triazido anion [(bipy)UO{sub 2}(N{sub 3}){sub 3}]{sup -} was obtained by the reaction of [UO{sub 2}(N{sub 3}){sub 2}.CH{sub 3}CN] with stoichiometric amounts of bipy and the ionic azide [PPh{sub 4}][N{sub 3}]. The reaction of [UO{sub 2}(N{sub 3}){sub 2}] with two equivalents of the [PPh{sub 4}][N{sub 3}] resulted in the formation of the mononuclear tetraazido anion [UO{sub 2}(N{sub 3}){sub 4}]{sup 2-} as well as the azido-bridged binuclear anion [(UO{sub 2}){sub 2}(N{sub 3}){sub 8}]{sup 4-}. The novel uranium oxoazides were characterized by their vibrational spectra and in the case of [(bipy){sub 2}(UO{sub 2}){sub 2}(N{sub 3}){sub 4}].CH{sub 3}CN, [PPh{sub 4}][(bipy)UO{sub 2}(N{sub 3}){sub 3}], [PPh{sub 4}]{sub 2}[UO{sub 2}(N{sub 3}){sub 4}], [PPh{sub 4}]{sub 2}[UO{sub 2}(N{sub 3}){sub 4}].2CH{sub 3}CN, and [PPh{sub 4}]{sub 4}[(UO{sub 2}){sub 2}(N{sub 3}){sub 8}].4CH{sub 3}CN by their X-ray crystal structures. (copyright 2017 Wiley-VCH Verlag GmbH and Co. KGaA, Weinheim)

Effect of alpha irradiation on UO{sub 2} surface reactivity in aqueous media

Energy Technology Data Exchange (ETDEWEB)

Jegou, C.; Muzeau, B.; Broudic, V.; Poulesquen, A.; Roudil, D. [Commissariat a l' Energie Atomique (CEA), Rhone Valley Research Center, DIEC/SESC/LMPA, Bagnols-sur-Ceze (France); Jorion, F. [Commissariat a l' Energie Atomique (CEA), Rhone Valley Research Center, DRCP/SE2A/LEMA, Bagnols-sur-Ceze (France); Corbel, C. [Commissariat a l' Energie Atomique (CEA), Saclay Research Center, DSM/DRECAM/SCM, Gif sur Yvette (France)

2005-07-01

The option of direct disposal of spent nuclear fuel in a deep geological formation raises the need to investigate the long-term behavior of the UO{sub 2} matrix in aqueous media subjected to {alpha}-{beta}-{gamma} radiation. The {beta}-{gamma} emitters account for most of the activity of spent fuel at the moment it is removed from the reactor, but diminish within a millennial time frame by over three orders of magnitude to less than the long-term activity. The latter persists over much longer time periods and must therefore be taken into account over a geological disposal time scale. Leaching experiments with solution renewal were carried out on UO{sub 2} pellets doped with alpha emitters ({sup 238}Pu and {sup 239}Pu) to quantify the impact of alpha irradiation on UO{sub 2} matrix alteration. Three batches of doped UO{sub 2} pellets with different alpha flux levels (3.30 x 10{sup 4}, 3.30 x 10{sup 5}, and 3.2 x 10{sup 6} {alpha} cm{sup -2} s{sup -1}) were studied. The results obtained in aerated and deaerated media immediately after sample annealing or interim storage in air provide a better understanding of the UO{sub 2} matrix alteration mechanisms under alpha irradiation. Interim storage in air of UO{sub 2} pellets doped with alpha emitters results in variations of the UO{sub 2} surface reactivity, which depends on the alpha particle flux at the interface and on the interim storage duration. The variation in the surface reactivity and the greater uranium release following interim storage cannot be attributed to the effect of alpha radiolysis in aerated media since the uranium release tends toward the same value after several leaching cycles for the doped UO{sub 2} pellet batches and spent fuel. Oxygen diffusion enhanced by alpha irradiation of the extreme surface layer and/or radiolysis of the air could account for the oxidation of the surface UO{sub 2} to UO{sub 2+x}. However, leaching experiments performed in deaerated media after annealing the samples and

Analysis of UO{sub 2}-BeO fuel under transient using fuel performance code

Energy Technology Data Exchange (ETDEWEB)

Gomes, Daniel S.; Abe, Alfredo Y.; Muniz, Rafael O.R.; Giovedi, Claudia, E-mail: dsgomes@ipen.br, E-mail: alfredo@ctmsp.mar.mil.br, E-mail: rafael.orm@gmail.com, E-mail: claudia.giovedi@ctmsp.mar.mil.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Universidade de São Paulo (USP), São Paulo, SP (Brazil). Departamento de Engenharia Naval e Oceânica

2017-11-01

Recent research has appointed the need to replace the classic fuel concept, used in light water reactors. Uranium dioxide has a weak point due to the low thermal conductivity, that produce high temperatures on the fuel. The ceramic composite fuel formed of uranium dioxide (UO{sub 2}), with the addition of beryllium oxide (BeO), presents high thermal conductivity compared with UO{sub 2}. The oxidation of zirconium generates hydrogen gas that can create a detonation condition. One of the preferred options are the ferritic alloys formed of iron-chromium and aluminum (FeCrAl), that should avoid the hydrogen release due to oxidation. In general, the FeCrAl alloys containing 10 - 20Cr, 3 - 5Al, and 0 - 0.12Y in weight percent. The FeCrAl alloys should exhibit a slow oxidation kinetics due to chemical composition. Resistance to oxidation in the presence of steam is improved as a function of the content of chromium and aluminum. In this way, the thermal and mechanical properties of the UO{sub 2}-BeO-10%vol, composite fuel were coupled with FeCrAl alloys and added to the fuel codes. In this work, we examine the fuel rod behavior of UO{sub 2}-10%vol-BeO/FeCrAl, including a simulated transient of reactivity. The fuels behavior shown reduced temperature with UO{sub 2}-BeO/Zr, UO{sub 2}-BeO/FeCrAl also were compared with UO{sub 2}/Zr system. The case reactivity initiated accident analyzed, reproducing the fuel rod called VA-1 using UO{sub 2}/Zr alloys and compared with UO{sub 2}-BeO/FeCrAl. (author)

The uranium recovery from UO{sub 2} kernel production effluent

Energy Technology Data Exchange (ETDEWEB)

Chen, Xiaotong, E-mail: chenxiaotong@tsinghua.edu.cn; He, Linfeng; Liu, Bing; Tang, Yaping; Tang, Chunhe

2016-12-15

Graphical abstract: In this study, a flow sheet including evaporation, flocculation, filtration, adsorption, and reverse osmosis was established for the UO{sub 2} kernel production effluent of HTR spherical fuel elements. The uranium recovery could reach 99.9% after the treatment, with almost no secondary pollution produced. Based on the above experimental results, the treating flow process in this study would be feasible for laboratory- and engineering-scale treatment of UO{sub 2} kernel production effluent of HTR spherical fuel elements. - Highlights: • A flow sheet including evaporation, flocculation, filtration, adsorption, and reverse osmosis was established for the UO{sub 2} kernel production effluent. • The uranium recovery could reach 99.9% after the treatment, with almost no secondary pollution produced. • The treating flow process would be feasible for laboratory- and engineering-scale treatment of UO{sub 2} kernel production effluent. - Abstract: For the fabrication of coated particle fuel elements of high temperature gas cooled reactors, the ceramic UO{sub 2} kernels are prepared through chemical gelation of uranyl nitrate solution droplets, which produces radioactive effluent with components of ammonia, uranium, organic compounds and ammonium nitrate. In this study, a flow sheet including evaporation, flocculation, filtration, adsorption, and reverse osmosis was established for the effluent treating. The uranium recovery could reach 99.9% after the treatment, with almost no secondary pollution produced.

Minerve: thermal-hydraulic phenomena simulation and virtual reality

International Nuclear Information System (INIS)

Laffont, A.; Pentori, B.

2003-01-01

MINERVE is a 3D interactive application representing the thermal-hydraulic phenomena happening in a nuclear plant. Therefore, the 3D geometric model of the French 900 MW PWR installations has been built. The users can interact in real time with this model to see at each step of the simulation what happens in the pipes. The thermal-hydraulic simulation is made by CATHARE-2, which calculates at every time step data on about one thousand meshes (the whole primary circuit, a part of the second circuit, and the Residual Heat Removal System). The simulation covers incidental and accidental cases on these systems. There are two main innovations in MINERVE: In the domain of nuclear plant's visualization, it is to introduce interactive 3D software mechanisms to visualize results of a physical simulation. In the domain of real-time 3D, it is to visualize fluids in a pipe, while they can have several configurations, like bubbles or single liquid phase. These mechanisms enable better comprehension and better visual representation of the possible phenomena. This paper describes the functionalities of MINERVE, and the difficulties to represent fluids with several characteristics like speed, configuration,..., in 3D. On the end, we talk about the future of MINERVE, and more widely of the possible futures of such an application in scientific visualization. (authors)

Minerve: thermal-hydraulic phenomena simulation and virtual reality

Energy Technology Data Exchange (ETDEWEB)

Laffont, A.; Pentori, B. [EDF R and D, EDF SEPTEN Electricity of France - Research and Development, Department SINETICS, 92 - Clamart (France)

2003-07-01

MINERVE is a 3D interactive application representing the thermal-hydraulic phenomena happening in a nuclear plant. Therefore, the 3D geometric model of the French 900 MW PWR installations has been built. The users can interact in real time with this model to see at each step of the simulation what happens in the pipes. The thermal-hydraulic simulation is made by CATHARE-2, which calculates at every time step data on about one thousand meshes (the whole primary circuit, a part of the second circuit, and the Residual Heat Removal System). The simulation covers incidental and accidental cases on these systems. There are two main innovations in MINERVE: In the domain of nuclear plant's visualization, it is to introduce interactive 3D software mechanisms to visualize results of a physical simulation. In the domain of real-time 3D, it is to visualize fluids in a pipe, while they can have several configurations, like bubbles or single liquid phase. These mechanisms enable better comprehension and better visual representation of the possible phenomena. This paper describes the functionalities of MINERVE, and the difficulties to represent fluids with several characteristics like speed, configuration,..., in 3D. On the end, we talk about the future of MINERVE, and more widely of the possible futures of such an application in scientific visualization. (authors)

SANDPIPER I (A comprehensive analysis programme for liquid moderated UO2 lattices)

International Nuclear Information System (INIS)

Alpiar, R.A.

1962-04-01

Methods of calculation for light water moderated reactors have recently been reviewed in AEEW R64. Calculation schemes for lattice parameters were presented which depended on the use of a number of IBM 704 and Perranti MERCURY Computer Programmes. SANDPIPER I is a comprehensive MERCURY programme designed to cover all the operations with a degree of accuracy adequate for survey calculations. The present version is restricted to regular or near regular UO 2 pin type lattices moderated by H 2 O, D 2 O, or organic liquids; it is planned to allow for greater flexibility in later versions of the programme. The present version is written in Autocode and requires a 4 drum machine. (author)

Atomic transport properties in UO2 and mixed oxides (U,Pu)O2

International Nuclear Information System (INIS)

Matzke, H.

1987-01-01

Atomic diffusion processes in UO 2 and in the fast-breeder reactor fuel, (U,Pu)O 2 are reviewed. Emphasis is given to the slower-moving species, i.e. U and Pu. Self-diffusion, chemical diffusion, diffusion in a thermal gradient, enhancement of diffusion by radiation and fission and the operative diffusion mechanisms are discussed. The main parameter, besides the temperature, is the oxygen-to-metal ratio (O/M ratio) of the oxide. The experimental results are compared with recent calculations reported elsewhere in this volume. Also treated are effects of the possible lambda-transition at ca.2600 K in UO 2 on high-temperature kinetic processes. The present knowledge on the diffusion and mobility of fission products with emphasis on volatile and gaseous elements, and of other actinides with emphasis on their valence states are treated. Gaps in our knowledge are pointed out and the relevance of the available results for oxide fuel during reactor operation is discussed. Whereas much is known for the as-produced 'virgin' fuel, more results are urgently needed for oxides with higher burn-ups containing a few per cent fission products. Finally, technological applications of the diffusion results are treated. As an example, important savings in cost, energy and time in fuel sintering were recently achieved based on basic studies of diffusion properties of UO 2 . (author)

Hypoeutectic melting in the UO{sub 2-x}-Gd{sub 2}O{sub 3} system

Energy Technology Data Exchange (ETDEWEB)

Journeau, Christophe, E-mail: christophe.journeau@cea.fr [CEA, DEN, SMTA, LPMA, Cadarache, F13108 St Paul lez Durance (France); Fouquart, Pascal [CEA, DEN, SMTA, LPMA, Cadarache, F13108 St Paul lez Durance (France); Domenger, Renaud; Allegri, Patrick [CEA, DEN, SGCS, LMAC, Marcoule, F30207 Bagnols sur Cèze (France)

2017-05-15

Gadolinium is one of the best neutron absorber materials and its use can be considered as a sacrificial material in a Sodium Fast Reactor core catcher in view of preventing recriticallity. A series of experiments have been conducted in the VITI induction-heated facility to study the melting in the UO{sub 2-x}-Gd{sub 2}O{sub 3} system with 60–87 mol% gadolinia. These experiments have indicated that the eutectic composition is around 92 mol% Gd{sub 2}O{sub 3} – 8 mol% UO{sub 2-x} and that the liquidus line is close to that of Popov et al. [Atom. Energ. 110 (2011) pp. 221–229] phase diagram. - Highlights: •Melting/Solidification experiments with UO{sub 2-x} and Gd{sub 2}O{sub 3} in reducing environment. •Eutectic composition around 92 mol% Gd{sub 2}O{sub 3}-8 mol% UO{sub 2-x}. •UO{sub 2-x} - Gd{sub 2}O{sub 3} liquidus line seems close to that of the pseudobinary phase diagram proposed by Popov et al. •Results will support the assessment of Gd{sub 2}O{sub 3} as a sacrificial material to mitigate criticality risk in SFR core catchers.

Ultrasonic analysis of UO{sub 2} pellets

Energy Technology Data Exchange (ETDEWEB)

Bittencourt, Marcelo de S.Q.; Baroni, Douglas B.; Martorelli, Daniel S., E-mail: bittenc@ien.gov.br, E-mail: douglasbaroni@ien.gov.br, E-mail: daniel@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil). Lab. de Ultrassom; Dias, Fabio C.; Silva, Jose W.S. da, E-mail: fabio@ird.gov.br, E-mail: wanderley@ird.gov.br [Instituto de Radioprotecao e Dosimetria (IRD/CNEN-RJ), Rio de Janeiro, RJ (Brazil). Lab. de Salvaguardas

2013-07-01

Ceramic materials have been widely used for various purposes in many different industries due to certain characteristics, such as high melting point and high resistance to corrosion. In the nuclear area, ceramics are of great importance due to the process of fabrication of fuel pellets for nuclear reactors. Generally, high accuracy destructive techniques are used to characterize nuclear materials for fuel fabrication. These techniques usually require costly equipment and facilities, as well as experienced personnel. This paper aims at presenting an analysis methodology for UO2 pellets using a non-destructive ultrasonic technique for porosity measurement. This technique differs from traditional ultrasonic techniques in the sense it uses ultrasonic pulses in frequency domain instead of time domain. Therefore, specific characteristics of the analyzed material are associated with the obtained frequency spectrum. In the present work, four fuel grade UO2 pellets were analyzed and the corresponding results evaluated. (author)

High temperature interaction between UO2 and Zircaloy-4/silver mixture

International Nuclear Information System (INIS)

Uetsuka, Hiroshi; Nagase, Fumihisa; Otomo, Takashi

1995-12-01

The reaction between UO 2 and Zircaloy is a main material interaction in the reactor core during a severe accident of LWR. With a view of examining the influence of the core materials having low melting temperatures on the reaction, the effect of silver that is main component of PWR control rod alloy was investigated in the temperature range from 1373 to 1703K. Zircaloy was completely liquefied by the same weight of liquid silver at tested temperatures. The reaction between UO 2 and (Zircaloy+silver) mixture roughly obeyed a parabolic rate law. The determined reaction rate below about 1600K was much lower than that obtained by Hofmann et al. for the reaction between UO 2 and Zircaloy. However, it sharply increased with temperature and became comparable with the rate of UO 2 /Zircaloy reaction at about 1700K. Metallurgical examination including EPMA analysis revealed that Zr(O) layer formed at the reaction interface only for the tests below about 1600K correlated with the discontinuity of the temperature dependence of reaction rate. (author)

Thoria-fuel irradiation. Program to irradiate 80% ThO2/20% UO2 ceramic pellets at the Savannah River Plant

International Nuclear Information System (INIS)

Pickett, J.B.

1982-02-01

This report describes the fabrication of proliferation-resistant thorium oxide/uranium oxide ceramic fuel pellets and preparations at the Savannah River Laboratory (SRL) to irradiate those materials. The materials were fabricated in order to study head end process steps (decladding, tritium removal, and dissolution) which would be required for an irradiated proliferation-resistant thorium based fuel. The thorium based materials were also to be studied to determine their ability to withstand average commercial light water reactor (LWR) irradiation conditions. This program was a portion of the Thorium Fuel Cycle Technology (TFCT) Program, and was coordinated by the Oak Ridge National Laboratory (ORNL) under the Consolidated Fuel Reprocessing Program (CFRP). The fuel materials were to be irradiated in a Savannah River Plant (SRP) reactor at conditions simulating the heat ratings and burnup of a commercial LWR. The program was terminated due to a de-emphasis of the TFCT Program, following completion of the fabrication of the fuel and the modified assemblies which were to be used in the SRP reactor. The reactor grade ceramic pellets were fabricated for SRL by Battelle, Pacific Northwest Laboratories. Five fuel types were prepared: 100% UO 2 pellets (control); 80% ThO 2 /20% UO 2 pellets; approximately 80% ThO 2 /20% UO 2 + 0.25 CaO (dissolution aid) pellets; 100% UO 2 hybrid pellets (prepared from sol-gel microspheres); and 100% ThO 2 pellets (control). All of the fuel materials were transferred to SRL from PNL and were stored pending a subsequent reactivation of the TFCT Programs

Modeling conversion of ammonium diuranate (ADU) into uranium dioxide (UO{sub 2}) powder

Energy Technology Data Exchange (ETDEWEB)

Hung, Nguyen Trong; Thuan, Le Ba [Institute for Technology of Radioactive and Rare Elements (ITRRE), 48 Lang Ha, Dong Da, Ha Noi (Viet Nam); Khoai, Do Van [Institute for Technology of Radioactive and Rare Elements (ITRRE), 48 Lang Ha, Dong Da, Ha Noi (Viet Nam); Current Postdoctoral Fellow at Tokai Reprocessing Technology Development Center, Japan Atomic Energy Agency (JAEA), 4-33 Tokaimura, Nakagun, Ibaraki, 319-1194 (Japan); Lee, Jin-Young, E-mail: jylee@kigam.re.kr [Convergence Research Center for Development of Mineral Resources (DMR), Korea Institute of Geoscience and Mineral Resources (KIGAM), Daejeon, 34132 (Korea, Republic of); Jyothi, Rajesh Kumar, E-mail: rkumarphd@kigam.re.kr [Convergence Research Center for Development of Mineral Resources (DMR), Korea Institute of Geoscience and Mineral Resources (KIGAM), Daejeon, 34132 (Korea, Republic of)

2016-10-15

In the paper, Brandon mathematical model that describes the relationship between the essential fabrication parameters [reduction temperature (T{sub R}), calcination temperature (T{sub C}), calcination time (t{sub C}) and reduction time (t{sub R})] and specific surface area of ammonium diuranate (ADU)-derived UO{sub 2} powder products was established. The proposed models can be used to predict and control the specific surface area of UO{sub 2} powders prepared through ADU route. Suitable temperatures for conversion of ADU and ammonium uranyl carbonate (AUC) was examined with the proposed model through assessment of the sinterability of UO{sub 2} powders.

Experimental studies of Micro- and Nano-grained UO₂: Grain Growth Behavior, Sufrace Morphology, and Fracture Toughness

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); Jamison, Laura M. [Argonne National Lab. (ANL), Argonne, IL (United States); Lian, Jie [Rensselaer Polytechnic Inst., Troy, NY (United States); Yao, Tiankai [Rensselaer Polytechnic Inst., Troy, NY (United States); Bhattacharya, Sumit [Argonne National Lab. (ANL), Argonne, IL (United States); Northwestern Univ., Evanston, IL (United States)

2016-01-01

This activity is supported by the US Nuclear Energy Advanced Modeling and Simulation (NEAMS) Fuels Product Line (FPL) and aims at providing experimental data for the validation of the mesoscale simulation code MARMOT. MARMOT is a mesoscale multiphysics code that predicts the coevolution of microstructure and properties within reactor fuel during its lifetime in the reactor. It is an important component of the Moose-Bison-Marmot (MBM) code suite that has been developed by Idaho National Laboratory (INL) to enable next generation fuel performance modeling capability as part of the NEAMS Program FPL. In order to ensure the accuracy of the microstructure-based materials models being developed within the MARMOT code, extensive validation efforts must be carried out. In this report, we summarize the experimental efforts in FY16 including the following important experiments: (1) in-situ grain growth measurement of nano-grained UO₂; (2) investigation of surface morphology in micrograined UO₂; (3) Nano-indentation experiments on nano- and micro-grained UO₂. The highlight of this year is: we have successfully demonstrated our capability to in-situ measure grain size development while maintaining the stoichiometry of nano-grained UO₂ materials; the experiment is, for the first time, using synchrotron X-ray diffraction to in-situ measure grain growth behavior of UO₂.

Recycling process of Mn-Al doped large grain UO2 pellets

International Nuclear Information System (INIS)

Nam, Ik Hui; Yang, Jae Ho; Rhee, Young Woo; Kim, Dong Joo; Kim, Jong Hun; Kim, Keon Sik; Song, Kun Woo

2010-01-01

To reduce the fuel cycle costs and the total mass of spent light water reactor (LWR) fuels, it is necessary to extend the fuel discharged burn-up. Research on fuel pellets focuses on increasing the pellet density and grain size to increase the uranium contents and the high burnup safety margins for LWRs. KAERI are developing the large grain UO 2 pellet for the same purpose. Small amount of additives doping technology are used to increase the grain size and the high temperature deformation of UO 2 pellets. Various promising additive candidates had been developed during the last 3 years and the MnO-Al 2 O 3 doped UO 2 fuel pellet is one of the most promising candidates. In a commercial UO 2 fuel pellet manufacturing process, defective UO 2 pellets or scraps are produced and those should be reused. A common recycling method for defective UO 2 pellets or scraps is that they are oxidized in air at about 450 .deg. C to make U 3 O 8 powder and then added to UO 2 powder. In the oxidation of a UO 2 pellet, the oxygen propagates along the grain boundary. The U 3 O 8 formation on the grain boundary causes a spallation of the grains. So, size and shape of U 3 O 8 powder deeply depend on the initial grain size of UO 2 pellets. In the case of Mn-Al doped large grain pellets, the average grain size is about 45μm and about 5 times larger than a typical un-doped UO 2 pellet which has grain size of about 8∼10μm. That big difference in grain size is expected to cause a big difference in recycled U 3 O 8 powder morphology. Addition of U 3 O 8 to UO 2 leads to a drop in the pellet density, impeding a grain growth and the formation of graph- like pore segregates. Such degradation of the UO 2 pellet properties by adding the recycled U 3 O 8 powder depend on the U 3 O 8 powder properties. So, it is necessary to understand the property and its effect on the pellet of the recycled U 3 O 8 . This paper shows a preliminary result about the recycled U 3 O 8 powder which was obtained by

Thermodynamic Behaviour of Hypostoichiometric UO{sub 2}; Comportement Thermodynamique de UO{sub 2} HypostoeChiometrique; Termodinamicheskoe povedenie gipostekhiometricheskoj UO{sub 2}; Comportamiento Termodinamico del UO{sub 2} Subestequiometrico

Energy Technology Data Exchange (ETDEWEB)

Aitken, E. A.; Brassfield, H. C.; Fryxell, R. E. [General Electric Company, Nuclear Materials and Propulsion Operation, Cincinnati, OH (United States)

1966-02-15

The ability of the UO{sub 2}-type structure to accomodate excess oxygen is well known. Recent evidence has indicated that this structure is stable also in the hypostoichiometric state at high temperatures and low oxygen partial pressures, but its manifestation occurs as a uranium metal precipitate in the oxide after cooling from high temperatures. This paper presents further evidence of the existence, at high temperatures, of a stable hypostoichiometric urania and describes in part the variation in thermodynamic properties across its homogeneity range. Hypostoichiometric UO{sub 2} evaporates congruently during free vaporization in slowly flowing hydrogen (-40 Degree-Sign C dew point) at 2400 Degree-Sign C at a composition having oxygen-to-uranium ratio of 1.88. If the temperature is decreased or the moisture content (oxygen partial pressure) increased, the congruent composition increases. The water content of the hydrogen at 2400 Degree-Sign C must be at least one per cent to maintain stoichiometric uranium dioxide. When UO{sub 2} pellets are sealed in tantalum cans and heated above 1700 Degree-Sign C, the O/U ratio of the pellet changes and reaches an equilibrium value which is governed by the oxygen activity of the atmosphere surrounding the can. UO{sub 2} does not react with tantalum but, because of the high solubility of oxygen in tantalum, the latter functions as a membrane. Using the data from congruent evaporation, and tantalum capsule tests, conducted in various argon-hydrogen mixtures, the oxygen activity in urania as a function of stoichiometry has been determined. The partial molar free energy of oxygen, G(O{sub 2} ), increases almost linearly on the oxygen deficient side with increasing oxygen-to-uranium ratio. Near the stoichiometric composition G(O{sub 2}) rises steeply. Using these results together with estimated G(O{sub 2}) values on the oxygen excess side obtained from the literature, it is shown that the data at a given temperature are consistent

A Study of the Temperature Distribution in UO{sub 2} Reactor Fuel Elements

Energy Technology Data Exchange (ETDEWEB)

Devold, I

1968-05-15

Thermal conductivity is one of the most important properties of nuclear reactor fuels. Accurate knowledge of this property is vital because, among other things, it determines the maximum power that can be taken out of the fuel element per unit length of the material without exceeding the safety limits of the fuel elements. This report consists of a study of the thermal behaviour of uranium dioxide in the form of reactor fuel. The experimental part of the report describes measurements performed at the OECD Halden Reactor Project, Halden, Norway. The experiment was originally set up in order to measure the temperature at the center of a UO{sub 2} fuel element as a function of element power, in order to determine the safe operation limit of the fuel assembly. However, in analysing the data obtained, very interesting thermal conductivity values were obtained and comparison with existing correlations could be performed. This comparison shows that a certain agreement is obtained between the measured data at Halden and a theory published by J.L. Bates in 1961, which predicts an increase in the thermal conductivity above 1500 deg C. The data obtained below 1300 deg C are also in good agreement with measurements performed by Vogt, Grandell and Runfors in 1964. The report contains a mathematical description of the heat transfer mechanisms in cylindrical fuel elements. The model is coded in FORTRAN IV-code and referred to as FUELTEMP.

Fabrication of Cr-doped UO2 Fuel Pellet using Liquid Phase Sintering

International Nuclear Information System (INIS)

Kim, Dong Joo; Yang, Jae Ho; Kim, Keon Sik; Rhee, Young Woo; Kim, Jong Hun; Oh, Jang Soo; Koo, Yang Hyun

2013-01-01

An enhancement of the thermal conductivity of a pellet can be obtained by the addition of a higher thermal conductive material in the pellet. In addition, the resistance to the PCI can be increased through a plasticity increase of the pellet. Thermal conductivity of ceramic materials is generally lower than that of metallic materials. The thermal conductivity of uranium oxide which is a typical ceramic material is low as well. The steep temperature gradient in the fuel pellet results from the low thermal conductivity. Therefore, the thermal conductivity improvement of a nuclear fuel pellet can enhance the fuel performance in various aspects. The lower centerline temperature of a fuel pellet affects the enhancement of fuel safety as well as fuel pellet integrity during nuclear reactor operation. Besides, the nuclear reactor power can be uprated due to the higher safety margin. So, many researches to enhance the thermal conductivity of nuclear fuel pellet have been performed in various ways. To improve the thermal conductivity of UO 2 pellet, an appropriate arrangement of the high thermal conductive material in UO 2 matrix is one of the various methods. We intended to control a placement of chromium as the high thermal conductive material. The metallic chromium and chromium oxide were arranged in a grain boundary of UO 2 using a liquid phase sintering method. The liquid phase sintering of Cr-doped UO 2 pellet could be adjusted using a control of an oxygen potential in sintering atmosphere

Analysis of a MOX-UO2 interface by the method of characteristics

International Nuclear Information System (INIS)

Chetaine, A.; Erradi, L.; Sanchez, R.; Zmijarevic, I.; Aniel-Buchheit, S.

2005-01-01

In the last few years many studies have been done to improve the ability of core reactors (PWR and BWR) to burn Plutonium fuel, either in mixed UO 2 /MOX pattern or full MOX pattern. The analysis of a MOX-UO 2 interface with the method of characteristics has been carried out. Comparisons with Monte Carlo and collision-probability calculations show that our results are in good agreement with those obtained by reference methods and qualify the method of characteristic as a reliable technique for such calculations. (authors)

Experimental Observation of Densification Behavior of UO2 Annular Pellet

International Nuclear Information System (INIS)

Kim, Dong-Joo; Rhee, Young-Woo; Kim, Jong-Hun; Yang, Jae-Ho; Kang, Ki-Won; Kim, Keon-Sik

2007-01-01

Recently, in the nuclear industry, one of the major issues is the improvement of a fuel economy. And many efforts have been made to develop a nuclear fuel for a high burnup and extended cycle. In the development of a high performance fuel, in-reactor fuel behavior (fission gas release, pellet-clad interaction, stress corrosion cracking, cladding corrosion, etc.) must be seriously reconsidered. Also, fuel fabrication (high enriched UO 2 powder handling, fuel rod and assembly manufacturing, fabricated fuel rod and assembly storage and transport, etc.) and an enrichment process (5 w/o criticality limit, etc.) must be discussed. A modification and an improvement of the nuclear fuel system will be also required. The typical fuel geometry of a PWR (Pressurized Water Reactor) is composed of a cylindrical pellet with a tubular cladding. And the outer surface of the cladding is cooled with water. However, to allow a substantial increase in the power density, an additional cooling is needed. One of the best ways is the application of the new fuel geometry that is of annular shape and has both internal and external cooling. From this point of view, the double cooled fuel is being developed by KAERI (Korea Atomic Energy Research Institute), and as a part of the project, the development of a fabrication process of a UO 2 annular pellet is now in progress. The dimensional behavior of UO 2 fuel is an important parameter in an irradiation performance. Various investigations (resintering test, model calculation, in-pile dimensional change measuring, etc.) had been performed. In designing a double cooled fuel, the importance of the dimensional behavior of a fuel pellet is higher, because the gap distance between a pellet and cladding can considerably affect on the in reactor fuel performance (gap conductance). And the dimensional behavior of an inner/outer gap is different with a cylindrical pellet, when the pellet shrinks (densification), the inner gap distance decreases and the

Determination of organic phosphorus in UO2C2O4·TRPO complex

International Nuclear Information System (INIS)

Guo Yifei; Yuan Jianhua; Liang Junfu; Jiao Rongzhou; Liu Xiuqin

2001-01-01

Organic phosphorus in UO 2 C 2 O 4 ·TRPO complex is converted to inorganic phosphorous with H 2 SO 4 -HNO 3 -H 2 O 2 wet cinefaction method. In 0.14 mol/L H 2 SO 4 solution containing water soluble poly vinylalcohol as stabilizing agent, the highly sensitive ion-associates are formed by the reaction of basic dye ethyl violet with heteropoly molybdophosphoric blue. Spectrophotometric method is used for determination of phosphorus with these ion-associates. The absorbance maximum is at 620 nm. Determination of phosphorus is not affected with mass ratios R(UO 2 2+ /P) ≤ 1.4 x 10 3 , R(C 2 O 4 2- /P) ≤ 8.8 x 10 2 and R(C 2 O 4 2- /P ≤ 3.6 x 10 4 (one time wet cinefaction must be carried out). In aqueous phase, phosphorus can be directly developed and determined. This method is contrasted with poly vinylalcohol-Rodamine B-heteropoly molybdophosphoric blue, analytical results are in good coincidence. Conversion ratio of phosphorus is 99.8% - 101.1%. The minimum detection limit is 0.02 mg/L. The relative standard deviation is 3%. The recovery ratio is 97% - 103%

Behavior of fission gases in nuclear fuel: XAS characterization of Kr in UO{sub 2}

Energy Technology Data Exchange (ETDEWEB)

Martin, P.M., E-mail: Philippe-m.martin@cea.fr [CEA, DEN, Cadarache DEC/SESC, F-13108 St-Paul-Lez-Durance Cedex (France); Vathonne, E.; Carlot, G.; Delorme, R.; Sabathier, C.; Freyss, M.; Garcia, P.; Bertolus, M. [CEA, DEN, Cadarache DEC/SESC, F-13108 St-Paul-Lez-Durance Cedex (France); Glatzel, P. [European Synchrotron Radiation Facility, 6 Rue Jules Horowitz, 38043 Grenoble (France); Proux, O. [OSUG, Observatoire des Sciences de l’Univers de Grenoble, CNRS and Université Joseph Fourier, BP 53, 38041 Grenoble Cedex 9 (France)

2015-11-15

X-ray Absorption Spectroscopy (XAS) was used to study the behavior of krypton as a function of its concentration in UO{sub 2} samples implanted with Kr ions. For a 0.5 at.% krypton local concentration, by combining XAS results and DFT + U calculations, we show that without any thermal treatment Kr atoms are mainly incorporated in the UO{sub 2} lattice as single atoms inside a neutral bound Schottky defect with O vacancies aligned along the (100) direction (BSD1). A thermal treatment at 1273 K induces the precipitation of dense Kr nano-aggregates, most probably solid at room temperature. In addition, 26 ± 2% of the Kr atoms remain inside BSD1 showing that Kr-BSD1 complex is stable up to this temperature. Consequently, the (in-)solubility of krypton in UO{sub 2} has to be re-evaluated. For high Kr concentration (8 at.%), XAS signals show that Kr atoms have precipitated in nanometer-sized aggregates with internal densities ranging between 4.15(7) g cm{sup −3} and 3.98(5) g cm{sup −3} even after annealing at 873 K. By neglecting the effect due to the UO{sub 2} matrix, the corresponding krypton pressures at 300 K were equal to 2.6(3) GPa and 2.0(2) GPa, respectively. After annealing at 1673 K, regardless of the initial Kr concentration, a bi-modal distribution is observed with solid nano-aggregates even at room temperature and larger cavities only partially filled with Kr. These results are very close to those observed in UO{sub 2} fuel irradiated in reactor. In this study we show that a rare gas can be used as a probe to investigate the defect creation and their stability in UO{sub 2}.

Recycling of nuclear fuel swarf at the fabrication of UO sub(2)-pellets and its influence on the irradiation behavior

International Nuclear Information System (INIS)

Dias, M.S.; Lameiras, F.S.; Santos, A.M.M. dos

1991-01-01

From the fabrication of UO sub(2) pellets for light water reactor fuel rods, nuclear fuel scraps results in form of UO sub(2) grinding swarf and UO sub(2) sinter scraps oxidized to U sub(3)O sub(8) powder. Detailed investigations on five types of UO sub(2) pellets fabricated with different portions of this scrap kinds added to the UO sub(2) press powder showed that there is only a small influence of such scrap additions on the irradiation behavior, especially for the fission gas release. This allows to recycle the fabrication scrap in a simple and economic way. (author)

Improving the Thermal Conductivity of UO2 Fuel with the Addition of Graphene

International Nuclear Information System (INIS)

Cho, Byoung Jin; Kim, Young Jin; Sohn, Dong Seong

2012-01-01

Improvement of fuel performances by increasing the fuel thermal conductivity using the BeO or W were reported elsewhere. In this paper, some major fuel performances of improved thermal conductivity oxide (ICO) nuclear fuel with the addition of 10 v/o graphene have been compared to those of standard UO 2 fuel. The fuel thermal conductivity affects many performance parameters and thus is an important parameter to determine the fuel performance. Furthermore, it also affects the performance of the fuel during reactor accidents. The improved thermal conductivity of the fuel would reduce the fuel temperature at the same power condition and would improve the fission gas release, rod internal pressure and fuel stored energy. Graphene is well known for its excellent electrical conductivity, strength and thermal conductivity. The addition of graphene to the UO 2 fuel could increase the thermal conductivity of the ICO fuel. Although the graphene material is extensively studied recently, the characteristics of the graphene material, especially the thermal properties, are not well-known yet. In this study, we used the Light Water Reactor fuel performance analysis code FRAPCON-3.2 to analyze the performance of standard UO 2 and ICO fuel

Concept and nuclear performance of direct-enrichment fusion breeder blanket using UO2 powder

International Nuclear Information System (INIS)

Oka, Yoshiaki; Kasahara, Takayasu; An, Shigehiro

1985-01-01

A new concept is presented for direct enrichment of fissile fuel in the blanket of a fusion-fission hybrid reactor. The enriched fuel produced by this means can be used in fission reactors without reprocessing. The outstanding feature of the concept is the powdered form in which UO 2 fuel is placed in the reactor blanket, where it is irradiated to the requisite enrichment for use as fuel in burner reactor, e.g. 3%. After removal from blanket, the powder is mixed to homogenize the enrichment. Fuel pellets and assemblies are then fabricated from the powder without reprocessing. The concept of irradiating UO 2 in powder eliminates the problems of spatial nonuniformity in fissile enrichment, and of radiation damage to fuel clad, encountered in attempting to enrich prefabricated fuel. Powder mixing for homogenization brings the additional benefit of removing volatile fission products. Also burnable poison can be added, as necessary, after irradiation. An extensive neutronic parameter survey showed that the optimum blanket arrangement for this enrichment concept is one presenting a fission suppressing configuration and with beryllium adopted as moderator. By this arrangement, the average 239 Pu enrichment obtained on the natural UO 2 fuel in the blanket reaches 3% after only 0.56 MW.yr/m"2 exposure. A conceptual design is presented of the blanket, together with associated fusion breeder, from which, practical application of the concept is shown to be promising. (author)

Microspheres of UO2, ThO2 and PuO2 for the high temperature reactor

International Nuclear Information System (INIS)

Brandau, T.; Brandau, E.

2010-01-01

Up to the end of the eighties of last century, the so called ''Kernels'', microspheres with a diameter of about 300 μm as sintered out of ThO 2 and UO 2 have been produced by a special vibrational dropping process. After coating and embedding in carbon the pebble fuel balls with a diameter of 60 mm included 40.000 UO 2 - or ThO 2 -microspheres in the core. Since the early nineties BRACE is developing the processing of microspheres with a broad range of materials for applications in chemical, pharmaceutical, electronic, cosmetic and food industries. One of the developing areas is the production of microspheres out of metal oxides, where different processes as sol-gel-, suspension- or mixed processes are used. (orig.)

High Pressure Low Temperature X-Ray Diffraction Studies of UO2 and UN single crystals.

Science.gov (United States)

Antonio, Daniel; Mast, Daniel; Lavina, Barbara; Gofryk, Krzysztof

Uranium dioxide is the most commonly used nuclear fuel material in commercial reactors, while uranium nitride also has many thermal and physical properties that make it attractive for potential use in reactors. Both have a cubic fcc lattice structure at ambient conditions and transition to antiferromagnetic order at low temperature. UO2 is a Mott insulator that orders in a complex non-collinear 3k magnetic structure at about 30 K, while UN has appreciable conductivity and orders in a simpler 1k magnetic structure below 52 K. Both compounds are characterized by strong magneto-structural interactions, understanding of which is vital for modeling their thermo-physical properties. While UO2 and UN have been extensively studied at and above room temperature, little work has been done to directly study the structure of these materials at low temperatures where magnetic interactions are dominant. In the course of our systematic studies on magneto vibrational behavior of UO2 and UN, here we present our recent results of high pressure X-Ray Diffraction (up to 35 GPa) measured below the Neel temperature using synchrotron radiation. Work supported by the Department of Energy, Office of Basic Energy Sciences, Materials Sciences, and Engineering Division.

Diffusion and release of noble gas and halogen fission products with several days half-life in UO2 particle

International Nuclear Information System (INIS)

Fang Chao

2013-01-01

The exact solutions of diffusion and release model of noble gas and halogen fission products in UO 2 particle of HTGR were built under the conditions of adsorption effect and other physical processes. The corresponding release fractions (F(t)) and the ratio of release and productive amounts (R(t)/B (t)) of fission products were also derived. Furthermore, the F(t) and R(t)/B(t) of 131 I, 131 IXe m , 133 Xe and 133 Xe m whose half-lifes are several days in UO 2 particle with the exact solutions, approximate solutions and corresponding numerical solutions under different temperature histories of reactor core were investigated. The results show that the F(t) and R(t)/B(t) are different in numerical values unless the time of release is long enough. The properties of conservation of exact solutions are much more reasonable than the ones of approximate solutions. It is also found that the results of exact solutions approach the actual working conditions more than the approximate and numerical solutions. (author)

Quality assurance and control in the manufacture of metalclad UO2 reactor fuels

International Nuclear Information System (INIS)

1976-01-01

The International Atomic Energy Agency has carried out a programme since its earliest days that includes the collection and dissemination of information on nuclear fuels. Since the 1960 symposium on Fuel Element Fabrication with Special Emphasis on Cladding Materials there has been an average of one meeting a year reviewing some aspect of fuel fabrication technology. A recent meeting dealing with the fabrication of UO 2 fuels was the Study Group on the Facilities and Technology needed for Nuclear Fuel Manufacture, held in Grenoble in 1972 (Rep. IAEA-158). After that meeting it became apparent that the quality of fuel production was an important aspect that had received inadequate coverage so far, and the Panel on Quality Assurance and Control in Nuclear Fuel Manufacture was convened by the Agency in Vienna in November 1974. In the working papers and discussions at the Panel meeting the viewpoints of different countries and of various interested parties, such as manufacturers, reactor operators and government authorities, were presented

A prediction of the inert gas solubilities in stoichiometric molten UO2

International Nuclear Information System (INIS)

Gunnerson, F.S.; Cronenberg, A.W.

1975-01-01

To analyze the effect of fission gas behaviour on fast reactor fuels during a hypothetical overpower transient, the solubility characteristics of the noble gases in molten UO 2 have been assessed. To accomplish this, a theoretical estimation of such solubilities is made by determining the reversible work required to introduce a hard sphere, the size of the gas atom, into the liquid solvent. Results indicate that the solubility of the noble gases in molten UO 2 is quite low, the molar fraction of gas-to-liquid being approximately 10 -6 . Such a low solubility of fission gases suggests that for preirradiated fuels, added swelling or formation may occur upon melting. In addition, such low solubility potential indicates that the fission gases do not play an appreciable role in the fragmentation of molten UO 2 upon quenching in sodium coolant. (Auth.)

Model Uncertainties for Valencia RPA Effect for MINERvA

Energy Technology Data Exchange (ETDEWEB)

Gran, Richard [Univ. of Minnesota, Duluth, MN (United States)

2017-05-08

This technical note describes the application of the Valencia RPA multi-nucleon effect and its uncertainty to QE reactions from the GENIE neutrino event generator. The analysis of MINERvA neutrino data in Rodrigues et al. PRL 116 071802 (2016) paper makes clear the need for an RPA suppression, especially at very low momentum and energy transfer. That published analysis does not constrain the magnitude of the effect; it only tests models with and without the effect against the data. Other MINERvA analyses need an expression of the model uncertainty in the RPA effect. A well-described uncertainty can be used for systematics for unfolding, for model errors in the analysis of non-QE samples, and as input for fitting exercises for model testing or constraining backgrounds. This prescription takes uncertainties on the parameters in the Valencia RPA model and adds a (not-as-tight) constraint from muon capture data. For MINERvA we apply it as a 2D ($q_0$,$q_3$) weight to GENIE events, in lieu of generating a full beyond-Fermi-gas quasielastic events. Because it is a weight, it can be applied to the generated and fully Geant4 simulated events used in analysis without a special GENIE sample. For some limited uses, it could be cast as a 1D $Q^2$ weight without much trouble. This procedure is a suitable starting point for NOvA and DUNE where the energy dependence is modest, but probably not adequate for T2K or MicroBooNE.

High temperature thermal conductivity measurements of UO2 by Direct Electrical Heating. Final report

International Nuclear Information System (INIS)

Bassett, B.

1980-10-01

High temperature properties of reactor type UO 2 pellets were measured using a Direct Electrical Heating (DEH) Facility. Modifications to the experimental apparatus have been made so that successful and reproducible DEH runs may be carried out while protecting the pellets from oxidation at high temperature. X-ray diffraction measurements on the UO 2 pellets have been made before and after runs to assure that sample oxidation has not occurred. A computer code has been developed that will model the experiment using equations that describe physical properties of the material. This code allows these equations to be checked by comparing the model results to collected data. The thermal conductivity equation for UO 2 proposed by Weilbacher has been used for this analysis. By adjusting the empirical parameters in Weilbacher's equation, experimental data can be matched by the code. From the several runs analyzed, the resulting thermal conductivity equation is lambda = 1/4.79 + 0.0247T/ + 1.06 x 10 -3 exp[-1.62/kT/] - 4410. exp[-3.71/kT/] where lambda is in w/cm K, k is the Boltzman constant, and T is the temperature in Kelvin

Development of ceramics based fuel, Phase I, Kinetics of UO{sub 2} sintering by vibration compacting of UO{sub 2} powder (Introductory report); Razvoj goriva na bazi keramike, I faza, Kinetika sinterovanja UO{sub 2} vibraciono kompaktiranje praha UO{sub 2} (Uvodni izvestaj)

Energy Technology Data Exchange (ETDEWEB)

Ristic, M M [Institute of Nuclear Sciences Vinca, Laboratorija za reaktorske materijale, Beograd (Serbia and Montenegro)

1962-10-15

After completing the Phase I of the task related to development of ceramics nuclear fuel the following reports are presented: Kinetics of UO{sub 2} sintering; Vibrational compacting and sintering of UO{sub 2}; Characterisation of of UO{sub 2} powder by DDK and TGA methods; Separation of UO{sub 2} powder.

Optimization of process parameters in precipitation for consistent quality UO2 powder production

International Nuclear Information System (INIS)

Tiwari, S.K.; Reddy, A.L.V.; Venkataswamy, J.; Misra, M.; Setty, D.S.; Sheela, S.; Saibaba, N.

2013-01-01

Nuclear reactor grade natural uranium dioxide powder is being produced through precipitation route, which is further processed before converting into sintered pellets used in the fabrication of PHWR fuel assemblies of 220 and 540 MWe type reactors. The process of precipitating Uranyl Nitrate Pure Solution (UNPS) is an important step in the UO 2 powder production line, where in soluble uranium is transformed into solid form of Ammonium Uranate (AU), which in turn reflects and decides the powder characteristics. Precipitation of UNPS with vapour ammonia is being carried out in semi batch process and process parameters like ammonia flow rate, temperature, concentration of UNPS and free acidity of UNPS are very critical and decides the UO 2 powder quality. Variation in these critical parameters influences powder characteristics, which in turn influences the sinterability of UO 2 powder. In order to get consistent powder quality and sinterability the critical parameter like ammonia flow rate during precipitation is studied, optimized and validated. The critical process parameters are controlled through PLC based automated on-line data acquisition systems for achieving consistent powder quality with increased recovery and production. The present paper covers optimization of process parameters and powder characteristics. (author)

Separation of UO{sub 2} powder; Separacija praha UO{sub 2}

Energy Technology Data Exchange (ETDEWEB)

Ristic, M M [Institute of Nuclear Sciences Vinca, Laboratorija za reaktorske materijale, Beograd (Serbia and Montenegro)

1962-10-15

This report deals with theoretical approach to separation process and describes the constructed separator with liquid medium. The separator was calibrated and tested with Al{sub 3}O{sub 3} and UO{sub 2}. it has been concluded that it can be used for separation of powders with sufficient accuracy if the separation is performed for a longer period of time. The separated fractions were characterised by microscopic method and the UO{sub 2} fraction additionally by sedimentation method.

Antiferromagnetic-ferromagnetic crossover in UO2-TiOx multi-phase systems

International Nuclear Information System (INIS)

Nakamura, Akio; Tsutsui, Satoshi; Yoshii, Kenji

2001-01-01

An antiferromagnetic (AF)-weakly ferromagnetic (WF) crossover has been found for UO 2 -TiO x multi-phase systems, (1-y)UO 2 +yTiO x (y=0.05-0.72, x=0, 1.0, 1.5 and 2.0), when these mixtures are heat treated at high temperature in vacuum. From the powder X-ray diffraction and electron-microprobe analyses, their phase assemblies were as follows: for x=0, 1.0 and 1.5, a heterogeneous two-phase mixture of UO 2 +TiO x ; for x=2.0, that of UO 2 +UTi 2 O 6 for y 0.67 that of UTi 2 O 6 +TiO 2 (plus residual minor UO 2 ). Magnetic susceptibility (χ) of the present UO 2 powder was confirmed to exhibit an antiferromagnetic sharp drop at T N (=30.5 K). In contrast, χ of these multi-phase systems was found to exhibit a sharp upturn at the respective T N , while their T N values remained almost constant with varying y. This χ upturn at T N is most pronounced for UO 2 +Ti-oxide (titania) systems (x=1.0, 1.5 and 2.0) over the wide mixture ratio above y∼0.10. These observations indicate that an AF-WF crossover is induced for these multi-phase systems, plausibly due to the interfacial magnetic modification of UO 2 in contact with the oxide partners

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)

Tests of the Hardware and Software for the Reconstruction of Trajectories in the Experiment MINERvA

International Nuclear Information System (INIS)

Palomino Gallo, Jose Luis

2009-01-01

MINERvA experiment has a highly segmented and high precision neutrino detector able to record events with high statistic (over 13 millions in a four year run). MINERvA uses FERMILAB NuMI beamline. The detector will allow a detailed study of neutrino-nucleon interactions. Moreover, the detector has a target with different materials allowing, for the first time, the study of nuclear effects in neutrino interactions. We present here the work done with the MINERvA reconstruction group that has resulted in: (a) development of new codes to be added to the RecPack package so it can be adapted to the MINERvA detector structure; (b) finding optimum values for two of the MegaTracker reconstruction package variables: PEcut = 4 (minimum number of photo electrons for a signal to be accepted) and Chi2Cut = 200 (maximum value of χ 2 for a track to be accepted); (c) testing of the multi anode photomultiplier tubes used at MINERvA in order to determine the correlation between different channels and for checking the device's dark counts.

Tests of the Hardware and Software for the Reconstruction of Trajectories in the Experiment MINERvA

Energy Technology Data Exchange (ETDEWEB)

Palomino Gallo, Jose Luis; /Rio de Janeiro, CBPF

2009-05-01

MINERvA experiment has a highly segmented and high precision neutrino detector able to record events with high statistic (over 13 millions in a four year run). MINERvA uses FERMILAB NuMI beamline. The detector will allow a detailed study of neutrino-nucleon interactions. Moreover, the detector has a target with different materials allowing, for the first time, the study of nuclear effects in neutrino interactions. We present here the work done with the MINERvA reconstruction group that has resulted in: (a) development of new codes to be added to the RecPack package so it can be adapted to the MINERvA detector structure; (b) finding optimum values for two of the MegaTracker reconstruction package variables: PEcut = 4 (minimum number of photo electrons for a signal to be accepted) and Chi2Cut = 200 (maximum value of {chi}{sup 2} for a track to be accepted); (c) testing of the multi anode photomultiplier tubes used at MINERvA in order to determine the correlation between different channels and for checking the device's dark counts.

The bare uranyl(2+) ion, UO22+

International Nuclear Information System (INIS)

Cornehl, H.H.; Heinemann, C.; Marcalo, J.; Pires de Matos, A.; Schwarz, H.

1996-01-01

Ion-molecule reactions between U 2+ and oxygen donors or charge-stripping collisions between singly charged UO 2 2 ions and O 2 collision partners generate uranyl(2+) ions in the gas phase. These do not readily dissociate into singly charged fragments. The standard enthalpy of formation for UO 2 2+ is estimated to be 371±60 kcal mol -1 , in accord with the results of ab initio calculations. (orig.)

Kinetics of UO{sub 2} sintering; Kinetika sinterovanja UO{sub 2}

Energy Technology Data Exchange (ETDEWEB)

Ristic, M M [Institute of Nuclear Sciences Vinca, Laboratorija za reaktorske materijale, Beograd (Serbia and Montenegro)

1962-10-15

Detailed conclusions related to the UO{sub 2} sintering can be drawn from investigating the kinetics of the sintering process. This report gives an thorough analysis of the the data concerned with sintering available in the literature taking into account the Jander and Arrhenius laws. This analysis completes the study of influence of the O/U ratio and the atmosphere on the sintering. Results presented are fundamentals of future theoretical and experimental work related to characterisation of the UO{sub 2} sintering process.

Role of nitrous acid during the dissolution of UO2 in nitric acid

International Nuclear Information System (INIS)

Deigan, N.; Pandey, N.K.; Kamachi Mudali, U.; Joshi, J.B.

2016-01-01

Understanding the dissolution behaviour of sintered UO 2 pellet in nitric acid is very important in designing an industrial scale dissolution system for the plutonium rich fast reactor MOX fuel. In the current article we have established the role of nitrous acid on the dissolution kinetics of UO 2 pellets in nitric acid. Under the chemical conditions that prevail in a typical Purex process, NO and NO 2 gases gets generated in the process streams. These gases produce nitrous acid in nitric acid medium. In addition, during the dissolution of UO 2 in nitric acid medium, nitrous acid is further produced in-situ at the pellet solution interface. As uranium dissolves oxidatively in nitric acid medium wherein it goes from U(IV) in solid to U(VI) in liquid, presence of nitrous acid (a good oxidizing agent) accelerates the reaction rate. Hence for determining the reaction mechanism of UO 2 dissolution in nitric acid medium, knowing the nitrous acid concentration profile during the course of dissolution is important. The current work involves the measurement of nitrous acid concentration during the course of dissolution of sintered UO 2 pellets in 8M starting nitric acid concentration as a function of mixing intensity from unstirred condition to 1500 RPM

Cation interdiffusion in the UO2 - (U, Pu)O2 and UO2 - PuO2 systems

International Nuclear Information System (INIS)

Leme, D.G.

1985-01-01

The interdiffusion of U and Pu ions in UO sub(2 +- x) - (U sub(0,83) Pu sub(0,17))O sub(2 + - x) and UO sub(2 + - x) -PuO sub(2 - x) sintered pellets and UO sub(2 +- x) -(U sub(0,82) Pu sub(0,18))O sub(2 + - x) single crystals has been studied as a function of the oxygen potential ΔG sup(-) (O 2 ) or the stoichiometric ratio O/M. The diffusion profiles of UO 2 /(U,Pu)O 2 and UO 2 /PuO 2 couples of different O/M ratios have been measured using high resolution α-spectrometer and microprobe. Thermal annealing of the specimens was performed in controlled atmospheres using either CO-CO 2 gas mixtures for constant O/M ratios or purified argon. The interdiffusion profiles have been analysed by means of the Boltzmann-Matano and Hall methods. The interdiffusion coefficient D sus(approx.) increases with increasing Pu content in sintered pellets (up to 17 wt. %PuO 2 ) showing a strong dependence of D sup(approx.) on the O/M ratio. The micropobe results show that the interdiffusion along grain boundaries is the main diffusion mechanism in the pellets. Experiments have also been carried out in single cristals to measure just the bulk-interdiffusion and avoiding effects due to grain boundaries. A marked dependence of D sup(approx.) on O/M ratio or on oxygen potential ΔG sup(-) (O 2 ), similar to the dependence already reported for self diffusion by means of radioactive tracers, has also been observed. (Author) [pt

Metallurgical structure modification of UO{sub 2} pellet during sintering - experience at NFC, Hyderabad, India

Energy Technology Data Exchange (ETDEWEB)

Santra, N.; Sinha, T.K.; Singh, A.K.; Sairam, S.; Sheela, S.; Saibaba, N., E-mail: santra@nfc.gov.in [Nuclear Fuel Complex, Dept. of Atomic Energy, Hyderabad (India)

2013-07-01

Nuclear Fuel Complex (NFC), Department of Atomic Energy (DAE) produces UO{sub 2} fuel pellets by powder compaction, high temperature sintering followed by centreless wet grinding method from the stabilized UO{sub 2} powder generated through ADU-route. Enhancement of fuel burn up of the Indian PHWRs becomes very important in order to effectively utilize the fuel to the maximum extent inside the reactor. Burn up is mainly limited by increased fission gas release from the fuel during reactor operation. Without introducing much change in the design, rate of release of fission gas can be reduced through enlargement of UO{sub 2} grain size. In Powder Metallurgical (PM) route of fuel fabrication, trials were taken by doping various oxide powder additives like TiO{sub 2}, Al{sub 2}O{sub 3}, SiO{sub 2}, Nb{sub 2}O{sub 5} and Cr{sub 2}O{sub 3}. The dopant normally goes into the solid solution of parent matrix during sintering at 1700 {sup o}C and thus enhance the rate of diffusion. Aliovalant dopant can alter the defect chemistry of the parent material either by creating vacancy or interstitial. It is apparently understood that the combination of above mechanisms are responsible for structural modification of UO{sub 2}. Hence selection of dopant remains largely empirical. It has been observed at NFC Hyderabad that the Cr{sub 2}O{sub 3} is the most suitable for achieving average UO{sub 2} grain size of about 70 micron and 98%TD of the sintered pellet. The paper discusses about the various experimental trials, sintered densities, metallographic examination, effect of different quantities, analysis and result obtained thereof. (author)

The burn-up credit physics and the 40. Minerve anniversary; La physique du credit Burn-Up et le 40. anniversaire de Minerve

Energy Technology Data Exchange (ETDEWEB)

Santamarina, A [CEA/Cadarache, Departement d' Etudes des Reacteurs, DER/SPRC, 13 - Saint-Paul-lez-Durance (France); Toubon, H [Cogema, 78 - Velizy Villacoublay (France); Trakas, C [FRAMATOME, 92 - Paris La Defense (France); and others

2000-03-21

The technical meeting organized by the SFEN on the burn-up credit (CBU) physics, took place the 23 november 1999 at Cadarache. the first presentation dealt with the economic interest and the neutronic problems of the CBU. Then two papers presented how taking into account the CBU in the industry in matter of transport, storage in pool, reprocessing and criticality calculation (MCNP4/Apollo2-F benchmark). An experimental method for the reactivity measurement through oscillations in the Minerve reactor, has been presented with an analysis of the possible errors. The future research program OSMOSE, taking into account the minor actinides in the CBU, was also developed. The last paper presented the national and international research programs in the CBU domain, in particular experiments realized in CEA/Valduc and the OECD Burn-up Criticality Benchmark Group activities. (A.L.B.)

Investigations of the trend followed in heat capacity of Re_6UO_1_2 (s) along lanthanide series

International Nuclear Information System (INIS)

Sahu, Manjulata; Saxena, M.K.; Rawat, Deepak; Dash, Smruti

2017-01-01

The compound RE_6UO_1_2 (s) (RE = Ho, Er, Tm, Yb and Lu) was synthesized by complex polymerisation method and characterised using X-ray diffraction (XRD). Heat capacity measurements of RE_6UO_1_2 (s) were performed with heat flux-type differential scanning calorimeter in the temperature range of 300-870 K. The trend in heat capacity along the rare earth series was proposed for RE_6UO_1_2 (s) and thermodynamic functions were generated. (author)

Benchmark testing of CENDL-2 for U-fuel thermal reactors

International Nuclear Information System (INIS)

Zhang Baocheng; Liu Guisheng; Liu Ping

1995-01-01

Based on CENDL-2, NJOY-WIMS code system was used to generate 69-group constants, and do benchmark testing for TRX-1,2; BAPL-UO-2-1,2,3; ZEEP-1,2,3. All the results proved that CENDL-2 is reliable for thermal reactor calculations. (3 tabs.)

Thermal expansion of UO2 and simulated DUPIC fuel

International Nuclear Information System (INIS)

Ho Kang, Kweon; Jin Ryu, Ho; Chan Song, Kee; Seung Yang, Myung

2002-01-01

The lattice parameters of simulated DUPIC fuel and UO 2 were measured from room temperature to 1273 K using neutron diffraction to investigate the thermal expansion and density variation with temperature. The lattice parameter of simulated DUPIC fuel is lower than that of UO 2 , and the linear thermal expansion of simulated DUPIC fuel is higher than that of UO 2 . For the temperature range from 298 to 1273 K, the average linear thermal expansion coefficients for UO 2 and simulated DUPIC fuel are 10.471x10 -6 and 10.751x10 -6 K -1 , respectively

Fabrication and Testing of Prototype APM-Clad UO{sub 2} Fuel Elements; Fabrication et essai de prototypes de cartouches de combustible en bioxyde d'uranium gaine d'aluminium (APM); Izgotovlenie i ispytanie prototipa toplivnykh ehlementov na osnove UO{sub 2} s obolochkoj iz alyuminiya metodom poroshkovoj metallurgii; Elaboracion y ensayo de elementos combustibles prototipo de UO{sub 2} con revestimiento de aluminio sinterizado

Energy Technology Data Exchange (ETDEWEB)

Ballif, III, J. L.; Friske, W. H.; Gordon, R. B. [Atomics International, Canoga Park, California (United States)

1963-11-15

In support of the 50-MW(e) Prototype Organic Power Reactor Programme (POPR), extensive development work has been performed on aluminium powder metallurgy (ARM) products, toward their use as cladding for UO{sub 2} fuel. As part of this development work, eutectic bonding, flash butt welding, and cold-pressure welding were investigated as methods for making end closures in die fuel element cladding. Vibratory packing was studied as a means of filling APM tubes with UO{sub 2}. Out-of-pile tests were conducted to obtain information on APM-UO{sub 2} compatibility. This work revealed that, under present conditions, eutectic bonding was the most suitable method for making end closures; vibratory packing produced fuel densities in the range of 80 to 88% of theoretical density; and no APM-UO{sub 2} reaction took place in the range of POPR operating temperatures (850{sup o}F maximum fuel-cladding interface temperature). As a result o f this development work, five APM-clad UO{sub 2} prototype fuel elements have been fabricated for testing in the Organic Moderated Reactor Experiment (OMRE). Each element consisted of 24 or 25 APM-clad fuel rods, arranged in a 5 x 5 array in a nickel-plated steel or an APM fuel box. To increase surface area, the extruded APM cladding had eight fins which were spiralled to a pitch of 45 or 90e/ ft to further improve heat transfer. The fuel rod end closures were made by eutectic bonding of silver-plated aluminium end plugs to the APM tubing. The elements were instrumented to: (1) Measure cladding surface and coolant temperatures, (2) Detect fuel rod failure, (3) Change coolant velocity (means of achieving peak cladding surface temperature of 850{sup o}F), (4) Measure coolant velocity, and (5) Measure fission gas build-up. These elements have been installed in the OMRE with target fuel burn-ups of 25000 to 30000 MWd/t of uranium. As of 1 April 1963, they had achieved accumulated burn-ups ranging from 7700 to 12 000 MWd/t of uranium. Two of the

Microbes make average 2 nanometer diameter crystalline UO2 particles.

Science.gov (United States)

Suzuki, Y.; Kelly, S. D.; Kemner, K. M.; Banfield, J. F.

2001-12-01

It is well known that phylogenetically diverse groups of microorganisms are capable of catalyzing the reduction of highly soluble U(VI) to highly insoluble U(IV), which rapidly precipitates as uraninite (UO2). Because biological uraninite is highly insoluble, microbial uranyl reduction is being intensively studied as the basis for a cost-effective in-situ bioremediation strategy. Previous studies have described UO2 biomineralization products as amorphous or poorly crystalline. The objective of this study is to characterize the nanocrystalline uraninite in detail in order to determine the particle size, crystallinity, and size-related structural characteristics, and to examine the implications of these for reoxidation and transport. In this study, we obtained U-contaminated sediment and water from an inactive U mine and incubated them anaerobically with nutrients to stimulate reductive precipitation of UO2 by indigenous anaerobic bacteria, mainly Gram-positive spore-forming Desulfosporosinus and Clostridium spp. as revealed by RNA-based phylogenetic analysis. Desulfosporosinus sp. was isolated from the sediment and UO2 was precipitated by this isolate from a simple solution that contains only U and electron donors. We characterized UO2 formed in both of the experiments by high resolution-TEM (HRTEM) and X-ray absorption fine structure analysis (XAFS). The results from HRTEM showed that both the pure and the mixed cultures of microorganisms precipitated around 1.5 - 3 nm crystalline UO2 particles. Some particles as small as around 1 nm could be imaged. Rare particles around 10 nm in diameter were also present. Particles adhere to cells and form colloidal aggregates with low fractal dimension. In some cases, coarsening by oriented attachment on \\{111\\} is evident. Our preliminary results from XAFS for the incubated U-contaminated sample also indicated an average diameter of UO2 of 2 nm. In nanoparticles, the U-U distance obtained by XAFS was 0.373 nm, 0.012 nm

Formation of ternary CaUO2(CO3)3(2-) and Ca2UO2(CO3)3(aq) complexes under neutral to weakly alkaline conditions.

Science.gov (United States)

Lee, Jun-Yeop; Yun, Jong-Il

2013-07-21

The chemical behavior of ternary Ca-UO2-CO3 complexes was investigated by using time-resolved laser fluorescence spectroscopy (TRLFS) in combination with EDTA complexation at pH 7-9. A novel TRLFS revealed two distinct fluorescence lifetimes of 12.7 ± 0.2 ns and 29.2 ± 0.4 ns for uranyl complexes which were formed increasingly dependent upon the calcium ion concentration, even though nearly indistinguishable fluorescence peak shapes and positions were measured for both Ca-UO2-CO3 complexes. For identifying the stoichiometric number of complexed calcium ions, slope analysis in terms of relative fluorescence intensity versus calcium concentration was employed in a combination with the complexation reaction of CaEDTA(2-) by adding EDTA. The formation of CaUO2(CO3)3(2-) and Ca2UO2(CO3)3(aq) was identified under given conditions and their formation constants were determined at I = 0.1 M Na/HClO4 medium, and extrapolated to infinitely dilute solution using specific ion interaction theory (SIT). As a result, the formation constants for CaUO2(CO3)3(2-) and Ca2UO2(CO3)3(aq) were found to be log β113(0) = 27.27 ± 0.14 and log β213(0) = 29.81 ± 0.19, respectively, providing that the ternary Ca-UO2-CO3 complexes were predominant uranium(vi) species at neutral to weakly alkaline pH in the presence of Ca(2+) and CO3(2-) ions.

Characterization of UO2 by infrared spectroscopy

International Nuclear Information System (INIS)

Faeda, Kelly C.M.; Machado, Geraldo C.; Lameiras, Fernando S.

2011-01-01

The characterization of nuclear fuel is of great importance to minimize the effects related to burnup and temperature and to achieve stability during in-core operation. The understanding the U-O system and its thermodynamic properties has fundamental importance in nuclear industry. Many physical properties of UO 2±x depend on the ratio O / U, such as the electrical conductivity and thermal properties, as well as the diffusivities of its constituents and solutes. The U-O system presents various oxides such as UO 2±x , U 4 O 9 , U 3 O 8 , and UO 3 . The control of the O/U relation is critical to the manufacturing process of UO 2 . In this work, the infrared spectroscopy was used to identify the presence of phases in UO 2 powder samples that cannot be identified by thermogravimetry and X-ray diffraction. (author)

Formation of (Cr, Al)UO{sub 4} from doped UO{sub 2} and its influence on partition of soluble fission products

Energy Technology Data Exchange (ETDEWEB)

Cooper, M.W.D. [Department of Materials, Imperial College London, London (United Kingdom); Gregg, D.J.; Zhang, Y.; Thorogood, G.J.; Lumpkin, G.R. [Institute of Materials Engineering, Australian Nuclear Science and Technology Organisation, Lucas Heights, New South Wales (Australia); Grimes, R.W. [Department of Materials, Imperial College London, London (United Kingdom); Middleburgh, S.C., E-mail: simm@ansto.gov.au [Institute of Materials Engineering, Australian Nuclear Science and Technology Organisation, Lucas Heights, New South Wales (Australia)

2013-11-15

CrUO{sub 4} and (Cr, Al)UO{sub 4} have been fabricated by a sol–gel method, studied using diffraction techniques and modelled using empirical pair potentials. Cr{sub 2}O{sub 3} was predicted to preferentially form CrUO{sub 4} over entering solution into hyper-stoichiometric UO{sub 2+x} by atomic scale simulation. Further, it was predicted that the formation of CrUO{sub 4} can proceed by removing excess oxygen from the UO{sub 2} lattice. Attempts to synthesise AlUO{sub 4} failed, instead forming U{sub 3}O{sub 8} and Al{sub 2}O{sub 3}. X-ray diffraction confirmed the structure of CrUO{sub 4} and identifies the existence of a (Cr, Al)UO{sub 4} phase for the first time (with a maximum Al to Cr mole ratio of 1:3). Simulation was subsequently used to predict the partition energies for the removal of fission products or fuel additives from hyper-stoichiometric UO{sub 2+x} and their incorporation into the secondary phase. The partition energies are consistent only with smaller cations (e.g. Zr{sup 4+}, Mo{sup 4+} and Fe{sup 3+}) residing in CrUO{sub 4}, while all divalent cations are predicted to remain in UO{sub 2+x}. Additions of Al had little effect on partition behaviour. The reduction of UO{sub 2+x} due to the formation of CrUO{sub 4} has important implications for the solution limits of other fission products as many species are less soluble in UO{sub 2} than UO{sub 2+x}.

Migration behavior of palladium in UO2, (3)

International Nuclear Information System (INIS)

Yoneyama, Mitsuru; Sato, Seichi; Ohashi, Hiroshi; Ogawa, Toru; Ito, Akinori; Fukuda, Kousaku.

1992-08-01

Palladium (Pd) is easily released from UO 2 kernels in HTGR coated fuel particles, and reacts with SiC coating layer. In addition, Pd is one of metallic fission products in irradiation UO 2 , which constitutes in dissoluble residue in reprocessing of LWR fuels. In the present investigation, the migration of palladium in UO 2 was examined by heating diffusion pairs sandwiched Pd foil between UO 2 wafers at 1300 ∼ 1800degC. Experiments were also carried out on affinity of Pd to UP 2 and a formation of U-Pd alloy. Pd was found mainly in the pores of UO 2 . The maximum depth intruded by Pd in fairly large amount was more than 100 μm for UO 2 with 90%TD and 50μm for UO 2 with 95%TD, while the maximum length of open pores was 330 μm for UO 2 with 90%TD, and 50 m for that with 95%TD. Fused Pd wetted UO 2 very much. Pd intruded deeply into UO 2 , especially in the edge of Pd droplet. Furthermore, U was detected either in precipitates or the Pd source with α-Pd phase of U-Pd alloy containing Pd at about 10at%. This fact indicates that Pd highly reacts with UO 2 . From the above results, the transport of Pd in UO 2 was explained by the model of gaseous diffusion through pores in UO 2 , which is retarded by formation of U-Pd alloy. It is also indicated that UPd 3 forms even at the oxygen potential condition of O/U ratio, which is a little higher than 2.00 on the basis of thermodynamic calculation. (author)

Development of UO2-30 WT per cent PuO2 fuel for FBTR

International Nuclear Information System (INIS)

Majumdar, S.; Kumar, Arun; Kamath, H.S.; Ramachandran, R.; Purushotham, D.S.C.; Roy, P.R.

1983-01-01

The specifications on Fast Breeder Reactor (FBTR) fuel pellets have two apparently contradictory requirements viz. (1) formation of homogeneous solid between UO 2 and PuO 2 which can only be achieved by high temperature sintering and (2) density of sintered pellets in the range of 92 ± 1 per cent T.D. which is normally achieved by low temperature sintering. Deactivation of starting powders under CO 2 or addition of volatile pore formers to the powders are the two methods which have been developed for lowering the denity of the pellets without reducing the sintering temperature. Two alternative fabrication routes utilizing these processes for manufacturing of FBTR pellets are described in this report. (author)

Application of pulsed electron beam vaporization to studies of UO2

International Nuclear Information System (INIS)

Benson, D.A.

1977-06-01

A method for determining the pressure versus internal energy coordinates of the liquid-vapor saturation curve is applied to the study of UO 2 . The experimental details and results of an initial series of tests are described. A comparison of the measurement results to models of the UO 2 equation of state illustrates the role of the heat capacity in describing the P--E characteristics of the state surface. A discussion of the available heat capacity information suggests that additional modeling and measurements of the heat capacity may be needed to give a complete temperature and energy dependent state surface description. Because of these modeling uncertainties, a method of thermodynamically describing the P(V, E) state surface entirely through the use of dynamic vapor measurements is given. Such a model satisfies transient thermomechanical analysis requirements. Next the effect of the state surface on one type of core disruptive reactor analysis is examined. And finally, the property determinations and models for UO 2 are reviewed with requirements for future work being outlined

Cracking and healing behavior of UO2 as related to pellet-cladding mechanical interaction. Interim report, July 1976

International Nuclear Information System (INIS)

Kennedy, C.R.; Yaggee, F.L.; Voglewede, J.C.; Kupperman, D.S.; Wrona, B.J.; Ellingson, W.A.; Johanson, E.; Evans, A.G.

1976-10-01

A direct-electrical-heating apparatus has been designed and fabricated to investigate those nuclear-fuel-related phenomena involved in the gap closure-bridging annulus formation mechanism that can be reproduced in an out-of-reactor environment. Prototypic light-water-reactor UO 2 fuel-pellet temperature profiles have been generated utilizing high flow rates (approximately 700 liters/min) of helium coolant gas, and a recirculating system has been fabricated to permit tests of up to 1000 h. Simulated light-water-reactor single- and multiple-thermal-cycle experiments will be conducted on both unclad and ceramic (fused silica) clad UO 2 pellet stacks. A laser dilatometer with a resolution of 1.27 x 10 -2 mm (5 x 10 -4 in.) is used to measure pellet dimensional increase continuously during thermal cycling. Acoustic emissions from thermal-gradient cracking have been detected and correlated with crack length and crack area. The acoustic emissions are monitored continuously to provide instantaneous information about thermal-gradient cracking. Posttest metallography and fracture-mechanics measurements are utilized to characterize cracking and crack healing

Summary of IEA-R1 research a reactor licensing related to its power increase from 2 to 10 MW

International Nuclear Information System (INIS)

1989-04-01

This work is a summary of IEA-R1 research reactor licensing related to its power increase from 2 to 10 MW. It reports also safety requirements, fuel elements, and reactor control modifications inherent to power increase. (A.C.A.S.)

Structural effects in UO{sub 2} thin films irradiated with U ions

Energy Technology Data Exchange (ETDEWEB)

Popel, A.J., E-mail: apopel@cantab.net [Department of Earth Sciences, University of Cambridge, Downing Street, Cambridge CB2 3EQ (United Kingdom); Adamska, A.M.; Martin, P.G.; Payton, O.D. [Interface Analysis Centre, School of Physics, University of Bristol, Bristol BS8 1TL (United Kingdom); Lampronti, G.I. [Department of Earth Sciences, University of Cambridge, Downing Street, Cambridge CB2 3EQ (United Kingdom); Picco, L.; Payne, L.; Springell, R.; Scott, T.B. [Interface Analysis Centre, School of Physics, University of Bristol, Bristol BS8 1TL (United Kingdom); Monnet, I.; Grygiel, C. [CIMAP, CEA-CNRS-ENSICAEN-Université de Caen, BP 5133, 14070 Caen Cedex5 (France); Farnan, I. [Department of Earth Sciences, University of Cambridge, Downing Street, Cambridge CB2 3EQ (United Kingdom)

2016-11-01

Highlights: • Quantitative characterisation of radiation damage by kernel average misorientation. • UO{sub 2} (1 1 1) plane showed higher irradiation tolerance than (1 1 0) plane. • UO{sub 2} film-YSZ substrate interface is stable under low fluence irradiation. • (0 0 1), (1 1 0), (1 1 1) single crystal UO{sub 2} thin films on YSZ substrates are expected. - Abstract: This work presents the results of a detailed structural characterisation of irradiated and unirradiated single crystal thin films of UO{sub 2}. Thin films of UO{sub 2} were produced by reactive magnetron sputtering onto (0 0 1), (1 1 0) and (1 1 1) single crystal yttria-stabilised zirconia (YSZ) substrates. Half of the samples were irradiated with 110 MeV {sup 238}U{sup 31+} ions to fluences of 5 × 10{sup 10}, 5 × 10{sup 11} and 5 × 10{sup 12} ions/cm{sup 2} to induce radiation damage, with the remainder kept for reference measurements. It was observed that as-produced UO{sub 2} films adopted the crystallographic orientation of their YSZ substrates. The irradiation fluences used in this study however, were not sufficient to cause any permanent change in the crystalline nature of UO{sub 2}. It has been demonstrated that the effect of epitaxial re-crystallisation of the induced radiation damage can be quantified in terms of kernel average misorientation (KAM) and different crystallographic orientations of UO{sub 2} respond differently to ion irradiation.

OSMOSE experiment representativity studies.

Energy Technology Data Exchange (ETDEWEB)

Aliberti, G.; Klann, R.; Nuclear Engineering Division

2007-10-10

The OSMOSE program aims at improving the neutronic predictions of advanced nuclear fuels through measurements in the MINERVE facility at the CEA-Cadarache (France) on samples containing the following separated actinides: Th-232, U-233, U-234, U-235, U-236, U-238, Np-237, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, Am-241, Am-243, Cm-244 and Cm-245. The goal of the experimental measurements is to produce a database of reactivity-worth measurements in different neutron spectra for the separated heavy nuclides. This database can then be used as a benchmark for integral reactivity-worth measurements to verify and validate reactor analysis codes and integral cross-section values for the isotopes tested. In particular, the OSMOSE experimental program will produce very accurate sample reactivity-worth measurements for a series of actinides in various spectra, from very thermalized to very fast. The objective of the analytical program is to make use of the experimental data to establish deficiencies in the basic nuclear data libraries, identify their origins, and provide guidelines for nuclear data improvements in coordination with international programs. To achieve the proposed goals, seven different neutron spectra can be created in the MINERVE facility: UO2 dissolved in water (representative of over-moderated LWR systems), UO2 matrix in water (representative of LWRs), a mixed oxide fuel matrix, two thermal spectra containing large epithermal components (representative of under-moderated reactors), a moderated fast spectrum (representative of fast reactors which have some slowing down in moderators such as lead-bismuth or sodium), and a very hard spectrum (representative of fast reactors with little moderation from reactor coolant). The different spectra are achieved by changing the experimental lattice within the MINERVE reactor. The experimental lattice is the replaceable central part of MINERVE, which establishes the spectrum at the sample location. This configuration

Studies of the role of molten materials in interactions with UO2 and graphite

International Nuclear Information System (INIS)

Fink, J.K.; Heiberger, J.J.; Leibowitz, L.

1979-01-01

Graphite, which is being considered as a lower reactor shield in gas-cooled fast reactors, would be contacted by core debris during a core disruptive accident. Information on the interaction of graphite, UO 2 , and stainless steel is needed in assessing the safety of the GCFR. In an ongoing study of the interaction of graphite, UO 2 , and stainless steel, the effects of the steel components have been investigated by electron microprobe scans, x-ray diffraction, and reaction-rate measurements. Experiments to study the role of the reaction product, FeUC 2 , in the interaction suggested that FeUC 2 promotes the interaction by acting as a carrier to bring graphite to the reaction site. Additional experiments using pyrolytic graphite show that while the reaction rate is decreased at 2400 K, at higher temperatures the rate is similar to that using other grades of graphite

The Surface Reactions of Ethanol over UO2(100) Thin Film

KAUST Repository

Senanayake, Sanjaya D.

2015-10-08

The study of the reactions of oxygenates on well-defined oxide surfaces is important for the fundamental understanding of heterogeneous chemical pathways that are influenced by atomic geometry, electronic structure and chemical composition. In this work, an ordered uranium oxide thin film surface terminated in the (100) orientation is prepared on a LaAlO3 substrate and studied for its reactivity with a C-2 oxygenate, ethanol (CH3CH2OH). With the use of synchrotron X-ray photoelectron spectroscopy (XPS), we have probed the adsorption and desorption processes observed in the valence band, C1s, O1s and U4f to investigate the bonding mode, surface composition, electronic structure and probable chemical changes to the stoichiometric-UO2(100) [smooth-UO2(100)] and Ar+-sputtered UO2(100) [rough-UO2(100)] surfaces. Unlike UO2(111) single crystal and UO2 thin film, Ar-ion sputtering of this UO2(100) did not result in noticeable reduction of U cations. The ethanol molecule has C-C, C-H, C-O and O-H bonds, and readily donates the hydroxyl H while interacting strongly with the UO2 surfaces. Upon ethanol adsorption (saturation occurred at 0.5 ML), only ethoxy (CH3CH2O-) species is formed on smooth-UO2(100) whereas initially formed ethoxy species are partially oxidized to surface acetate (CH3COO-) on the Ar+-sputtered UO2(100) surface. All ethoxy and acetate species are removed from the surface between 600 and 700 K.

The Surface Reactions of Ethanol over UO2(100) Thin Film

KAUST Repository

Senanayake, Sanjaya D.; Mudiyanselage, Kumudu; Burrell, Anthony K; Sadowski, Jerzy T.; Idriss, Hicham

2015-01-01

The study of the reactions of oxygenates on well-defined oxide surfaces is important for the fundamental understanding of heterogeneous chemical pathways that are influenced by atomic geometry, electronic structure and chemical composition. In this work, an ordered uranium oxide thin film surface terminated in the (100) orientation is prepared on a LaAlO3 substrate and studied for its reactivity with a C-2 oxygenate, ethanol (CH3CH2OH). With the use of synchrotron X-ray photoelectron spectroscopy (XPS), we have probed the adsorption and desorption processes observed in the valence band, C1s, O1s and U4f to investigate the bonding mode, surface composition, electronic structure and probable chemical changes to the stoichiometric-UO2(100) [smooth-UO2(100)] and Ar+-sputtered UO2(100) [rough-UO2(100)] surfaces. Unlike UO2(111) single crystal and UO2 thin film, Ar-ion sputtering of this UO2(100) did not result in noticeable reduction of U cations. The ethanol molecule has C-C, C-H, C-O and O-H bonds, and readily donates the hydroxyl H while interacting strongly with the UO2 surfaces. Upon ethanol adsorption (saturation occurred at 0.5 ML), only ethoxy (CH3CH2O-) species is formed on smooth-UO2(100) whereas initially formed ethoxy species are partially oxidized to surface acetate (CH3COO-) on the Ar+-sputtered UO2(100) surface. All ethoxy and acetate species are removed from the surface between 600 and 700 K.

A thermal hydraulic analysis in PWR reactors with UO2 or (U-Th)O2 fuel rods employing a simplified code

International Nuclear Information System (INIS)

Santos, Thiago A. dos; Maiorino, José R.; Stefanni, Giovanni L. de

2017-01-01

In order to project a nuclear reactor, the neutronic calculus must be validated, so that its thermal limits and safety parameters are respected. Considering this issue, this research aims to evaluate the APTh-100 reactor thermal limits. This PWR is a project developed in Universidade Federal do ABC (UFABC) using fuel composed of Uranium and Thorium oxide mixed (U,Th)O 2 . For this purpose, a simplified, although conservative, code was developed in a MATLAB environment named STC-MOX-Th 'Simplified Thermal-hydraulics Code-Mixed Oxide Thorium'. This code provides axial and radial temperature distribution, as well as DNBR distribution over the hottest channel of the reactor core. Moreover, it brings other hydraulic quantities, such as pressure drop over the fuel rod, considering any fuel proportion of (U,Th)O 2 .The software uses basic laws of conservation of mass, momentum and energy, it also calculates the thermal conduction equation, considering the thermal conductive coefficient as a temperature function. In order to solve this equation, the finite elements method was used. Furthermore, the proportion of 36% of UO 2 was used to evaluate the temperature over the fuel rod and DNBR minimum in three burn conditions: beginning, middle and ending. The program has proven to be efficient in every condition and the results evidenced that the APTh-1000 reactor, in an initial analysis, has its thermal limits within the recommended security parameters. (author)

Development of ceramics based fuel, Phase I, Kinetics of UO2 sintering by vibration compacting of UO2 powder (Introductory report)

International Nuclear Information System (INIS)

Ristic, M.M.

1962-10-01

After completing the Phase I of the task related to development of ceramics nuclear fuel the following reports are presented: Kinetics of UO 2 sintering; Vibrational compacting and sintering of UO 2 ; Characterisation of of UO 2 powder by DDK and TGA methods; Separation of UO 2 powder

A calorimetric and thermodynamic investigation of A2[(UO2)2(MoO4)O2] compounds with A = K and Rb and calculated phase relations in the system (K2MoO4 + UO3 + H2O)

International Nuclear Information System (INIS)

Lelet, Maxim I.; Suleimanov, Evgeny V.; Golubev, Aleksey V.; Geiger, Charles A.; Bosbach, Dirk; Alekseev, Evgeny V.

2015-01-01

Highlights: • We determined the low temperature heat capacity of A 2 [(UO 2 ) 2 (MoO 4 )O 2 ] compounds with A = K and Rb. • We determined enthalpy of formation of K 2 [(UO 2 ) 2 (MoO 4 )O 2 ] by HF solution calorimetry. • We calculated Δ f G° (T = 298 K) of all phases from studied series. • Using obtained data we performed a thermodynamic modelling in the system (K 2 MoO 4 + UO 3 + H 2 O). - Abstract: A calorimetric and thermodynamic investigation of two alkali-metal uranyl molybdates with general composition A 2 [(UO 2 ) 2 (MoO 4 )O 2 ], where A = K and Rb, was performed. Both phases were synthesized by solid-state sintering of a mixture of potassium or rubidium nitrate, molybdenum (VI) oxide and gamma-uranium (VI) oxide at high temperatures. The synthetic products were characterised by X-ray powder diffraction and X-ray fluorescence methods. The enthalpy of formation of K 2 [(UO 2 ) 2 (MoO 4 )O 2 ] was determined using HF-solution calorimetry giving Δ f H° (T = 298 K, K 2 [(UO 2 ) 2 (MoO 4 )O 2 ], cr) = −(4018 ± 8) kJ · mol −1 . The low-temperature heat capacity, C p °, was measured using adiabatic calorimetry from T = (7 to 335) K for K 2 [(UO 2 ) 2 (MoO 4 )O 2 ] and from T = (7 to 326) K for Rb 2 [(UO 2 ) 2 (MoO 4 )O 2 ]. Using these C p ° values, the third law entropy at T = 298.15 K, S°, is calculated as (374 ± 1) J · K −1 · mol −1 for K 2 [(UO 2 ) 2 (MoO 4 )O 2 ] and (390 ± 1) J · K −1 · mol −1 for Rb 2 [(UO 2 ) 2 (MoO 4 )O 2 ]. These new experimental results, together with literature data, are used to calculate the Gibbs energy of formation, Δ f G°, for both phases giving: Δ f G° (T = 298 K, K 2 [(UO 2 ) 2 (MoO 4 )O 2 ], cr) = (−3747 ± 8) kJ · mol −1 and Δ f G° (T = 298 K, Rb 2 [(UO 2 ) 2 (MoO 4 )], cr) = −3736 ± 5 kJ · mol −1 . Smoothed C p °(T) values between 0 K and 320 K are presented, along with values for S° and the functions [H°(T) − H°(0)] and [G°(T) − H°(0)], for both phases. The

Thermal expansion of UO2-Gd2O3 fuel pellets

International Nuclear Information System (INIS)

Une, Katsumi

1986-01-01

In recent years, more consideration has been given to the application of UO 2 -Gd 2 O 3 burnable poison fuel to LWRs in order to improve the core physics and to extend the burnup. It has been known that UO 2 forms a single phase cubic fluorite type solid solution with Gd 2 O 3 up to 20 - 30 wt.% above 1300 K. The addition of Gd 2 O 3 to UO 2 lattices changes the properties of the fuel pellets. The limited data on the thermal expansion of UO 2 -Gd 2 O 3 fuel exist, but those are inconsistent. UO 2 -Gd 2 O 3 fuel pellets were fabricated, and the linear thermal expansion of UO 2 and UO 2 -(5, 8 and 10 wt.%)Gd 2 O 3 fuel pellets was measured with a differential dilatometer over the temperature range of 298 - 1973 K. A sapphire rod of 6 mm diameter and 15.5 mm length was used as the reference material. After the preheating cycle, the measurement was performed in argon atmosphere. The results for UO 2 pellets showed excellent agreement with the data in literatures. The linear thermal expansion of UO 2 -Gd 2 O 3 fuel pellets showed the increase with increasing the Gd 2 O 3 content. Consideration must be given to this excessive expansion in the fuel design of UO 2 -Gd 2 O 3 pellets. The equations for the linear thermal expansion and density of UO 2 -Gd 2 O 3 fuel pellets were derived by the method of least squares. (Kako, I.)

Shutdown channels and fitted interlocks in atomic reactors

International Nuclear Information System (INIS)

Furet, J.; Landauer, C.

1968-01-01

This catalogue consists of tables (one per reactor) giving the following information: number and type of detectors, range of the shutdown channels, nature of the associated electronics, thresholds setting off the alarms, fitted interlocks. These cards have been drawn up with a view to an examination of the reactors safety by the 'Reactor Safety Sub-Commission', they take into account the latest decisions. The reactors involved in this review are: Azur, Cabri, Castor-Pollux, Cesar-Marius-2, Edf-2, EL3, EL4, Eole, G1, G2-G3, Harmonie, Isis, Masurca, Melusine, Minerve, Osiris, Pegase, Peggy, PAT, Rapsodie, SENA, Siloe, Siloette, Triton-Nereide, and Ulysse. (authors) [fr

Simulation of reactivity-initiated accident transients on UO2-M5® fuel rods with ALCYONE V1.4 fuel performance code

Directory of Open Access Journals (Sweden)

Isabelle Guénot-Delahaie

2018-03-01

Full Text Available The ALCYONE multidimensional fuel performance code codeveloped by the CEA, EDF, and AREVA NP within the PLEIADES software environment models the behavior of fuel rods during irradiation in commercial pressurized water reactors (PWRs, power ramps in experimental reactors, or accidental conditions such as loss of coolant accidents or reactivity-initiated accidents (RIAs. As regards the latter case of transient in particular, ALCYONE is intended to predictively simulate the response of a fuel rod by taking account of mechanisms in a way that models the physics as closely as possible, encompassing all possible stages of the transient as well as various fuel/cladding material types and irradiation conditions of interest. On the way to complying with these objectives, ALCYONE development and validation shall include tests on PWR-UO2 fuel rods with advanced claddings such as M5® under “low pressure–low temperature” or “high pressure–high temperature” water coolant conditions.This article first presents ALCYONE V1.4 RIA-related features and modeling. It especially focuses on recent developments dedicated on the one hand to nonsteady water heat and mass transport and on the other hand to the modeling of grain boundary cracking-induced fission gas release and swelling. This article then compares some simulations of RIA transients performed on UO2-M5® fuel rods in flowing sodium or stagnant water coolant conditions to the relevant experimental results gained from tests performed in either the French CABRI or the Japanese NSRR nuclear transient reactor facilities. It shows in particular to what extent ALCYONE—starting from base irradiation conditions it itself computes—is currently able to handle both the first stage of the transient, namely the pellet-cladding mechanical interaction phase, and the second stage of the transient, should a boiling crisis occur.Areas of improvement are finally discussed with a view to simulating and

Performance evaluation of UO2-Zr fuel in power ramp tests

International Nuclear Information System (INIS)

Knudsen, P.; Bagger, C.

1977-01-01

In power reactors using UO 2 -Zr fuel, rapid power increases may lead to failures in fuel pins that have been irradiated at steady or decreasing heat loads. This paper presents results which extend the experience with power ramp performance of high burn-up fuel pins. A test fuel element containing both pellet and vipac UO 2 -Zr fuel pins was irradiated in the HBWR at Halden for effectively 2 1/2 years to an average burn-up of 21,000 MWD/te UO 2 at gradually decreasing power levels. The subsequent non-destructive characterization revealed formation of transverse cracks in the vipac fuel columns. After the HBWR irradiation, five of the fuel pins were power ramp tested individually in the DR 3 Reactor at Riso. The ramp rates in this test series were in the range 3-60 W/cm min. The maximum local heat loads seen in the ramp tests were 20-120% above the highest levels experienced at the same axial positions during the HBWR irradiation. Three pellets and one vipac fuel pin failed, whereas another vipac pin gave no indication of clad penetration. Profilometry after the ramp testing indicated the formation of small ridges for both types of fuel pins. For vipac fuel, the ridges were less regularly distributed along the pin length than for pellet fuel. Neutron radiography revealed the formation of additional transverse and longitudinal fuel cracks during the power ramps for both types of fuel pins. The observed failures seemed to be marginal since little or no indication as to the locations of the clad penetrations could be derived from the non-destructive post-irradiation examinations. The cases have been analyzed by means of the Danish fuel performance codes. The calculations, which are in general agreement with the observations, are discussed. The results of the investigations indicate qualitative similarities in over power performance of the two fuel types

Post-irradiation examination of fifteen UO2/PuO2-fuel pins from the experiment DFR-350

International Nuclear Information System (INIS)

Geithoff, D.

1975-06-01

Within the framework of the fuel pin development for a sodium-cooled fast reactor a subassembly containing 77 fuel pins has been irradiated up to 5.65% fima in the Dounreay fast reactor. The pins were prototypes in terms of fuel and cladding material. The fuel consisted of mechanically mixed UO 2 (80%) and PuO 2 (20%) pressed into pellets whereas austenitic steels (W.-No. 1,4961 and 1,4988) were used as cladding material. Furthermore a blanket column of UO 2 pellets and a gas plenum were incorporated in the pin. For irradiation the conditions in a fast breeder were simulated by a linear rod power of 450 W/cm and a maximum cladding temperature of 630 0 C. After the successful completion of the irradiation, the subassembly was dismantled and fifteen pins were selected for a nondestructive and destructive examination. The tests included visual control, measurement of external dimensions, γ-spectroscopy, X-ray radiography, fission gas measurement, ceramography, radiochemical burn-up measurement. The results are presented. The most important results of the examinations seem to be the migration of fission product cesium and the fact that no signs of impending pin failure have been found. Thus the pin specification tested in this experiment is capable of achieving higher burnups under the irradiation conditions described above. (orig./AK) [de

Thermal hydraulic analysis of the IPR-R1 TRIGA reactor

International Nuclear Information System (INIS)

Veloso, Marcelo Antonio; Fortini, Maria Auxiliadora

2002-01-01

The subchannel approach, normally employed for the analysis of power reactor cores that work under forced convection, have been used for the thermal hydraulic evaluation of a TRIGA Mark I reactor, named IPR-R1, at 250 kW power level. This was accomplished by using the PANTERA-1P subchannel code, which has been conveniently adapted to the characteristics of natural convection of TRIGA reactors. The analysis of results indicates that the steady state operation of IPR-R1 at 250 kW do not imply risks to installations, workers and public. (author)

Technological investigation for producing UO2 powder from ADU by using rotary furnace

International Nuclear Information System (INIS)

Pham Duc Thai; Ngo Trong Hiep; Dam Van Tien; Vu Quang Chat; Nguyen Duy Lam; Ngo Xuan Hung; Ngo Quang Hien; Tran Duy Hai; Nguyen Van Sinh

2003-01-01

Uranium dioxide powder UO 2 is main material for producing UO 2 fuel ceramic pellets. The technical characteristics of UO 2 powder directly affect on mechanical and physical characteristics of UO 2 fuel ceramic pellets. Project titled 'Technological investigation for producing UO 2 powder from ADU by using rotary furnace' with the code number BO/01/03-06 for two years 2001 and 2002, on purpose to step by step perfect the technology and equipments for producing UO 2 powder, that is as nuclear fuel. This UO 2 powder may be good material for producing UO 2 fuel ceramic pellets. The results had been achieved as follows: 1. Study on the perfection of the reduction process U 3 O 8 to UO 2 in the gas mixture of 3H 2 + N 2 in inactive condition. 2. Study, design and production of active device system called rotary furnace for manufacturing UO 2 powder from ADU. 3. Study on 4 steps of technology process: drying, calcination, reduction and stabilization of UO 2 powder in the system of rotary furnace from which obtained UO 2 with technical characteristics meeting basic criteria of UO 2 fuel powder. (author)

Description and crystal structure of albrechtschraufite, MgCa{sub 4}F{sub 2}[UO{sub 2}(CO{sub 3}){sub 3}]{sub 2}.17-18H{sub 2}O

Energy Technology Data Exchange (ETDEWEB)

Mereiter, K. [Vienna Univ. of Technology (Austria). Inst. of Chemical Technologies and Analytics

2013-04-15

Albrechtschraufite, MgCa{sub 4}F{sub 2}[UO{sub 2}(CO{sub 3}){sub 3}]{sub 2}.17-18H{sub 2}O, triclinic, space group P anti 1, a = 13.569(2), b = 13.419(2), c = 11.622(2) Aa, α = 115.82(1), β = 107.61(1), γ = 92.84(1) (structural unit cell, not reduced), V = 1774.6(5) Aa{sup 3}, Z = 2, Dc = 2.69 g/cm{sup 3} (for 17.5 H{sub 2}O), is a mineral that was found in small amounts with schroeckingerite, NaCa{sub 3}F[UO{sub 2}(CO{sub 3}){sub 3}](SO{sub 4}).10H{sub 2}O, on a museum specimen of uranium ore from Joachimsthal (Jachymov), Czech Republic. The mineral forms small grain-like subhedral crystals (= 0.2 mm) that resemble in appearance liebigite, Ca{sub 2}[UO{sub 2}(CO{sub 3}){sub 3}]. ∝ 11H{sub 2}O. Colour pale yellow-green, luster vitreous, transparent, pale bluish green fluorescence under ultraviolet light. Optical data: Biaxial negative, nX = 1.511(2), nY = 1.550(2), nZ = 1.566(2), 2V = 65(1) (λ = 589 nm), r < v weak. After qualitative tests had shown the presence of Ca, U, Mg, CO{sub 2} and H{sub 2}O, the chemical formula was determined by a crystal structure analysis based on X-ray four-circle diffractometer data. The structure was later on refined with data from a CCD diffractometer to R1 = 0.0206 and wR2 = 0.0429 for 9,236 independent observed reflections. The crystal structure contains two independent [UO{sub 2}(CO{sub 3}){sub 3}]{sup 4-} anions of which one is bonded to two Mg and six Ca while the second is bonded to only one Mg and three Ca. Magnesium forms a MgF{sub 2}(O{sub carbonate}){sub 3}(H{sub 2}O) octahedron that is linked via the F atoms with three Ca atoms so as to provide each F atom with a flat pyramidal coordination by one Mg and two Ca. Calcium is 7- and 8-coordinate forming CaFO{sub 6}, CaF{sub 2}O{sub 2}(H{sub 2}O){sub 4}, CaFO{sub 3}(H{sub 2}O){sub 4} and CaO{sub 2}(H{sub 2}O){sub 6} coordination polyhedra. The crystal structure is built up from MgCa{sub 3}F{sub 2}[UO{sub 2}(CO{sub 3}){sub 3}].8H{sub 2}O layers parallel to (001) which

Experimental study of the IPR-R1 TRIGA reactor power channels responses

International Nuclear Information System (INIS)

Mesquita, Henrique F.A.; Ferreira, Andrea V.

2015-01-01

The IPR-R1 nuclear reactor installed at Centro de Desenvolvimento da Tecnologia Nuclear CDTN/CNEN, Belo Horizonte, Brazil, is a Mark I TRIGA reactor (Training, Research, Isotopes, General Atomics) and became operational on November of 1960. The reactor has four irradiation devices: a rotary specimen rack with 40 irradiation channels, the central tube, and two pneumatic transfer tubes. The nuclear reactor is operated in a power range between zero and 100 kW. The instrumentation for IPR-R1 operation is mainly composed of four neutronic channels for power measurements. The aim of this work is to investigate the responses of neutronic channels of IPR-R1, Linear, Log N and Percent Power channels, and to check their linearity. Gold foils were activated at low powers (0.125-1.000 kW), and cobalt foils were activated at high powers (10-100kW). For each sample irradiated at rotary specimen rack, another one was irradiated at the same time at the pneumatic transfer tube-2. The obtained results allowed evaluating the linearity of the neutronic channels responses. (author)

Optimization of process parameters in precipitation for consistent quality UO{sub 2} powder production

Energy Technology Data Exchange (ETDEWEB)

Tiwari, S.K.; Reddy, A.L.V.; Venkataswamy, J.; Misra, M.; Setty, D.S.; Sheela, S.; Saibaba, N., E-mail: misra@nfc.gov.in [Nuclear Fuel Complex, Hyderabad (India)

2013-07-01

Nuclear reactor grade natural uranium dioxide powder is being produced through precipitation route, which is further processed before converting into sintered pellets used in the fabrication of PHWR fuel assemblies of 220 and 540 MWe type reactors. The process of precipitating Uranyl Nitrate Pure Solution (UNPS) is an important step in the UO{sub 2} powder production line, where in soluble uranium is transformed into solid form of Ammonium Uranate (AU), which in turn reflects and decides the powder characteristics. Precipitation of UNPS with vapour ammonia is being carried out in semi batch process and process parameters like ammonia flow rate, temperature, concentration of UNPS and free acidity of UNPS are very critical and decides the UO{sub 2} powder quality. Variation in these critical parameters influences powder characteristics, which in turn influences the sinterability of UO{sub 2} powder. In order to get consistent powder quality and sinterability the critical parameter like ammonia flow rate during precipitation is studied, optimized and validated. The critical process parameters are controlled through PLC based automated on-line data acquisition systems for achieving consistent powder quality with increased recovery and production. The present paper covers optimization of process parameters and powder characteristics. (author)

Facile reductive silylation of UO{sub 2}{sup 2+} to uranium(IV) chloride

Energy Technology Data Exchange (ETDEWEB)

Kiernicki, John J.; Bart, Suzanne C. [H.C. Brown Laboratory, Department of Chemistry, Purdue University, West Lafayette, IN (United States); Zeller, Matthias [H.C. Brown Laboratory, Department of Chemistry, Purdue University, West Lafayette, IN (United States); Department of Chemistry, Youngstown State University, Youngstown, OH (United States)

2017-01-19

General reductive silylation of the UO{sub 2}{sup 2+} cation occurs readily in a one-pot, two-step stoichiometric reaction at room temperature to form uranium(IV) siloxides. Addition of two equivalents of an alkylating reagent to UO{sub 2}X{sub 2}(L){sub 2} (X=Cl, Br, I, OTf; L=triphenylphosphine oxide, 2,2'-bipyridyl) followed by two equivalents of a silyl (pseudo)halide, R{sub 3}Si-X (R=aryl, alkyl, H; X=Cl, Br, I, OTf, SPh), cleanly affords (R{sub 3}SiO){sub 2}UX{sub 2}(L){sub 2} in high yields. Support is included for the key step in the process, reduction of U{sup VI} to U{sup V}. This procedure is applicable to a wide range of commercially available uranyl salts, silyl halides, and alkylating reagents. Under this protocol, one equivalent of SiCl{sub 4} or two equivalents of Me{sub 2}SiCl{sub 2} results in direct conversion of the uranyl to uranium(IV) tetrachloride. Full spectroscopic and structural characterization of the siloxide products is reported. (copyright 2017 Wiley-VCH Verlag GmbH and Co. KGaA, Weinheim)

The burn-up credit physics and the 40. Minerve anniversary

International Nuclear Information System (INIS)

Santamarina, A.; Toubon, H.; Trakas, C.

2000-01-01

The technical meeting organized by the SFEN on the burn-up credit (CBU) physics, took place the 23 november 1999 at Cadarache. the first presentation dealt with the economic interest and the neutronic problems of the CBU. Then two papers presented how taking into account the CBU in the industry in matter of transport, storage in pool, reprocessing and criticality calculation (MCNP4/Apollo2-F benchmark). An experimental method for the reactivity measurement through oscillations in the Minerve reactor, has been presented with an analysis of the possible errors. The future research program OSMOSE, taking into account the minor actinides in the CBU, was also developed. The last paper presented the national and international research programs in the CBU domain, in particular experiments realized in CEA/Valduc and the OECD Burn-up Criticality Benchmark Group activities. (A.L.B.)

Measurements of the viscosity of sodium tetraborate (borax)-UO2 and of sodium metaborate-UO2 liquid solutions

International Nuclear Information System (INIS)

Dalle Donne, M.; Dorner, S.; Roth, A.

1983-01-01

Adding UO 2 produces an increase of viscosity of borax and sodium metaborate. For temperatures below 920 0 C the measurements with the borax-UO 2 solution show a phase separation. Contrary to borax the sodium metaborate solutions indicate a well defined melting point. At temperatures slightly below the melting point a solid phase is formed. The tested sodium-borates-UO 2 mixtures are in liquid form. (DG)

Porosity influence on UO2 pellet fracture

International Nuclear Information System (INIS)

Quadros, N.F. de; Abreu Aires, M. de; Gentile, E.F.

1976-01-01

Compression tests were made with UO 2 pellets with grain size of 0,01 mm, approximately the same for all pellets, and with different porosities. The strain rate was 5,5 X 10 -5 sec -1 at room temperature. From fractographic studies and observations made during the compression tests, it was suggested that the pores and flaws resulting from sintering at 1650 0 C, play a fundamental role on the fracture mechanism of the UO 2 pellets [pt

Compliance characteristics of cracked UO2 pellets

International Nuclear Information System (INIS)

Williford, R.E.; Mohr, C.L.; Lanning, D.D.

1981-01-01

The thermally induced cracking of UO 2 fuel pellets causes simultaneous reductions of the bulk (extrinsic) fuel thermal conductivity and elastic moduli to values significantly less than those for solid pellets. The magnitude of these bulk properly reductions was found to be primarily dependent on the amount of crack area in the transverse plane of the fuel. The model described herein uses a simple description of the crack geometry to couple the fuel rod thermal and mechanical behaviors by relating in-reactor data to Hooke's Law and a crack compliance model. Data from the NRC/PNL Halden experiment IFA-432 show that for a typical helium-filled BWR-design rod at 30 kW/m, the effective thermal conductivity and elastic moduli of the cracked fuel are 4/5 and 1/40 of that for solid pellets, respectively

Tracer surface diffusion on UO2

International Nuclear Information System (INIS)

Zhou, S.Y.; Olander, D.R.

1983-06-01

Surface diffusion on UO 2 was measured by the spreading of U-234 tracer on the surface of a duplex diffusion couple consisting of wafers of depleted and enriched UO 2 joined by a bond of uranium metal

High temperature investigation of the solid/liquid transition in the PuO2-UO2-ZrO2 system

Science.gov (United States)

Quaini, A.; Guéneau, C.; Gossé, S.; Sundman, B.; Manara, D.; Smith, A. L.; Bottomley, D.; Lajarge, P.; Ernstberger, M.; Hodaj, F.

2015-12-01

The solid/liquid transitions in the quaternary U-Pu-Zr-O system are of great interest for the analysis of core meltdown accidents in Pressurised Water Reactors (PWR) fuelled with uranium-dioxide and MOX. During a severe accident the Zr-based cladding can become completely oxidised due to the interaction with the oxide fuel and the water coolant. In this framework, the present analysis is focused on the pseudo-ternary system UO2-PuO2-ZrO2. The melting/solidification behaviour of five pseudo-ternary and one pseudo-binary ((PuO2)0.50(ZrO2)0.50) compositions have been investigated experimentally by a laser heating method under pre-set atmospheres. The effects of an oxidising or reducing atmosphere on the observed melting/freezing temperatures, as well as the amount of UO2 in the sample, have been clearly identified for the different compositions. The oxygen-to-metal ratio is a key parameter affecting the melting/freezing temperature because of incongruent vaporisation effects. In parallel, a detailed thermodynamic model for the UO2-PuO2-ZrO2 system has been developed using the CALPHAD method, and thermodynamic calculations have been performed to interpret the present laser heating results, as well as the high temperature behaviour of the cubic (Pu,U,Zr)O2±x-c mixed oxide phase. A good agreement was obtained between the calculated and experimental data points. This work enables an improved understanding of the major factors relevant to severe accident in nuclear reactors.

Formation, stability and structural characterization of ternary MgUO{sub 2}(CO{sub 3}){sub 3}{sup 2-} and Mg{sub 2}UO{sub 2}(CO{sub 3}){sub 3}(aq) complexes

Energy Technology Data Exchange (ETDEWEB)

Lee, Jun-Yeop; Yun, Jong-Il [KAIST, Daejeon (Korea, Republic of). Dept. of Nuclear and Quantum Engineering; Vespa, Marika; Gaona, Xavier; Dardenne, Kathy; Rothe, Joerg; Rabung, Thomas; Altmaier, Marcus [Karlsruhe Institute of Technology, Karlsruhe (Germany). Inst. for Nuclear Waste Disposal

2017-06-01

The formation of ternary Mg-UO{sub 2}-CO{sub 3} complexes under weakly alkaline pH conditions was investigated by time-resolved laser fluorescence spectroscopy (TRLFS) and extended X-ray absorption fine structure (EXAFS) and compared to Ca-UO{sub 2}-CO{sub 3} complexes. The presence of two different Mg-UO{sub 2}-C{sub 3} complexes was identified by means of two distinct fluorescence lifetimes of 17±2 ns and 51±2 ns derived from the multi-exponential decay of the fluorescence signal. Slope analysis in terms of fluorescence intensity coupled with fluorescence intensity factor as a function of log [Mg(II)] was conducted for the identification of the Mg-UO{sub 2}-CO{sub 3} complexes forming. For the first time, the formation of both MgUO{sub 2}(CO{sub 3}){sub 3}{sup 2-} and Mg{sub 2}UO{sub 2}(CO{sub 3}){sub 3}(aq) species was confirmed and the corresponding equilibrium constants were determined as log β {sub 113}=25.8±0.3 and β {sub 213}=27.1±0.6, respectively. Complementarily, fundamental structural information for both Ca-UO{sub 2}-CO{sub 3} and Mg-UO{sub 2}-CO{sub 3} complexes was gained by extended EXAFS revealing very similar structures between these two species, except for the clearly shorter U-Mg distance (3.83 Aa) compared with U-Ca distance (4.15 Aa). These results confirmed the inner-sphere character of the Ca/Mg-UO{sub 2}-CO{sub 3} complexes. The formation constants determined for MgUO{sub 2}(CO{sub 3}){sub 3}{sup 2-} and Mg{sub 2}UO{sub 2}(CO{sub 3}){sub 3}(aq) species indicate that ternary Mg-UO{sub 2}-CO{sub 3} complexes contribute to the relevant uranium species in carbonate saturated solutions under neutral to weakly alkaline pH conditions in the presence of Mg(II) ions, which will induce notable influences on the U(VI) chemical species under seawater conditions.

UO2 pellet and manufacturing method

International Nuclear Information System (INIS)

Komada, Kiichi; Nishinaka, Keiji; Adachi, Kazunori; Fujiwara, Shuji.

1995-01-01

The present invention concerns an uranium dioxide pellet having a large crystal grain size. The grain size of the pellet is enlarged to increase the distance of an FP gas generated in the crystal grain to reach the grain boundary and, as a result, decrease the releasing speed of the FP gas. A UO 2 powder having a specific surface area of from 5 to 50m 2 /g is used as a starting powder in a step of forming a molding product, and chlorine or a chlorine compound is added in such an amount that the chlorine content in the UO 2 pellet is from 3 to 25ppm, in one of a production step, a molding step or a sintering step for UO 2 powder. With such procedures, a UO 2 pellet having a large crystal grain size can be prepared with good reproducibility. (T.M.)

Sintering of Kernel UO2 for High Temperature Reactor Fuel

International Nuclear Information System (INIS)

Sukarsono; Dwi-Heru-Sucahyo; Hidayati; Evi-Hertiviana; Bambang-Sugeng

2000-01-01

Sintering investigation of UO 2 gel has been done. The gel was preparedthrough two ways. The first, gel was produced using PVA as additive agent.The second gel was produced using HMTA and Urea as additive agent. From thepreparation of gel, the PVA method better than the urea - HMTA method,because was not necessary the cold temperature for sol preparation and alsowas not necessary the hot temperature for gelation process. After nextprocessing, the sintered gel of gel through PVA, also better than HMTAprocess. (author)

Characterization of Compaction Process on UO2 Powder Pelletisation

International Nuclear Information System (INIS)

Rachmawati, M; Langenati, R; Saputra, T.T; Mahpudin, A; Histori; Sutarya, D; Zahedi

1998-01-01

Determination of compaction pressure of pelletization which is based on density characterization in conjunction with satisfactory green strength of the UO 2 pellet, is carried out in this experiment. Cameco UO 2 powder has been mixed up with Zn-stearate lubricant prior to compaction process. The compaction pressure is varied from the range of 2 Mp up to 6 Mp. The mechanical strength is determined using diametral compression strength with the speed of loading of 0.1 mm.min 1 . The density measurement and compression strength test are performed on each of the applied pressure. The result shows that compaction at 5 Mp gives the maximum green strength of UO 2 pellet, while the maximum density is achieved at 5.7 Mp. The maximum green strength and green density of UO 2 (+ TiO 2 ) pellets is achieved at the addition of 0.25% and 0.125% TiO 2 respectively. The compaction pressure which is showing the maximum pellet green strength but still having the required density, is chosen to be the determinant compaction pressure in condition of pelletization

NuMI Flux Predictions for NOvA and MINERvA

Science.gov (United States)

Aliaga Soplin, Leonidas; Nova Collaboration; Minerva Collaboration

2017-01-01

The determination of the neutrino flux in any conventional neutrino beam presents a challenge for the current and future short and long baseline neutrino experiments. The uncertainties associated with the production and attenuation of the hadrons in the beamline materials along with those associated with the beam optics have a big effect in the knowledge of the flux. For experiments like MINERvA and NOvA, understanding the flux is crucial since it enters directly into every neutrino-nucleus cross-section measurement. The majority of this work involves predicting the neutrino flux using dedicated hadron production measurements from hadron-nucleus collisions. The predictions at the MINERvA and NOvA near detectors are presented as well as the results of incorporating in-situ MINERvA data that can provide additional constraints. These results have been fully implemented in MINERvA and they are currently use for its cross-section analysis. The implementation for NoVA is underway. The procedure and conclusions of this work will have a big impact on future hadron production experiments and on determining the flux for the upcoming DUNE experiment.

Electrical system regulations of the IEA-R1 reactor

International Nuclear Information System (INIS)

Mello, Jose Roberto de; Madi Filho, Tufic

2013-01-01

The IEA-R1 reactor of the Nuclear and Energy Research Institute (IPEN-CNEN/SP), is a research reactor open pool type, designed and built by the U.S. firm Babcock and Wilcox, having, as coolant and moderator, deionized light water and beryllium and graphite, as reflectors. Until about 1988, the reactor safety systems received power from only one source of energy. As an example, it may be cited the control desk that was powered only by the vital electrical system 220V, which, in case the electricity fails, is powered by the generator group: no-break 220V. In the years 1989 and 1990, a reform of the electrical system upgrading to increase the reactor power and, also, to meet the technical standards of the ABNT (Associacao Brasileira de Normas Tecnicas) was carried out. This work has the objective of showing the relationship between the electric power system and the IEA-R1 reactor security. Also, it demonstrates that, should some electrical power interruption occur, during the reactor operation, this occurrence would not start an accident event. (author)

CLUMPED LIGHT WATER MODERATED UO$sub 2$ SUPERHEAT CRITICALS. PART I. EXPERIMENTS

Energy Technology Data Exchange (ETDEWEB)

Warzek, F. G.; Johnston, H. F.

1963-11-15

The following critical and subcritical measurements were made in the EVESR core: reactivity with no control rods; full core reactivity with control rods; and power distribution in the full core with control rods. The fuel was UO/ sub 2/, and the elements were of the superheating type. The reactor was light- water-cooled and -moderated. (T.F.H.)

High temperature thermal conductivity measurements of UO/sub 2/ by Direct Electrical Heating. Final report. [MANTRA-III

Energy Technology Data Exchange (ETDEWEB)

Bassett, B

1980-10-01

High temperature properties of reactor type UO/sub 2/ pellets were measured using a Direct Electrical Heating (DEH) Facility. Modifications to the experimental apparatus have been made so that successful and reproducible DEH runs may be carried out while protecting the pellets from oxidation at high temperature. X-ray diffraction measurements on the UO/sub 2/ pellets have been made before and after runs to assure that sample oxidation has not occurred. A computer code has been developed that will model the experiment using equations that describe physical properties of the material. This code allows these equations to be checked by comparing the model results to collected data. The thermal conductivity equation for UO/sub 2/ proposed by Weilbacher has been used for this analysis. By adjusting the empirical parameters in Weilbacher's equation, experimental data can be matched by the code. From the several runs analyzed, the resulting thermal conductivity equation is lambda = 1/4.79 + 0.0247T/ + 1.06 x 10/sup -3/ exp(-1.62/kT/) - 4410. exp(-3.71/kT/) where lambda is in w/cm K, k is the Boltzman constant, and T is the temperature in Kelvin.

Study of UO{sub 2}F{sub 2} - H{sub 2}O - HF compounds; Etude des composes UO{sub 2}F{sub 2} - H{sub 2}O - HF

Energy Technology Data Exchange (ETDEWEB)

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

1961-07-01

We study various compounds resulting from the interaction of UO{sub 2}F{sub 2} with H{sub 2}O and HF (gas), and various triple compounds UO{sub 2}F{sub 2} - H{sub 2}O - HF; the conditions of decomposition and the thermodynamic limits of stability are specified. (author) [French] Nous etudions divers composes formes par reaction de UO{sub 2}F{sub 2} avec H{sub 2}O et HF (gaz) et divers composes triples UO{sub 2}F{sub 2} - H{sub 2}O - HF, en essayant de preciser les decompositions et domaines d'exisfence thermodynamiques de ces corps. (auteur)

Comparative study of the different industrial manufacturing routes for UO2 pellet specifications through the wet process

International Nuclear Information System (INIS)

Palheiros, Franklin; Gonzaga, Reinaldo; Soares, Alexandre

2009-01-01

In the fuel cycle, converting UF 6 to UO 2 powder is an intermediate step for fabrication of pellets for fuel assemblies to be used in nuclear power plants. The basic proposal common to the different powder fabrication processes is to provide raw material capable of being processed into the form of pellets. The wet processes is the most often used industrially and are divided in two categories: the ADU (Ammonium Diuranate) and AUC (Ammonium Uranyl Carbonate) processes, whose names originate in the precipitate obtained in aqueous solution during the intermediate steps of UO 2 powder fabrication. It has known that the powder characteristics have a considerable influence in the UO 2 pellet manufacturing and quality characteristics. INB has used the AUC process to produce UO 2 pellets and supply fuel to Angra 1 and 2 Nuclear Power Plants. Despite of this process is characterized by the precipitation of a different intermediate precipitate compared to the ADU route (i.e., (NH 4 ) 4 UO 2 (CO 3 ) 3 , in the AUC process, and (NH 4 ) 2 U 2 O 7 in ADU process) leading to some slight differences in the final pellet microstructure, it has been considered that the models that predict the pellet behavior during irradiation in a nuclear reactor are basically the same compared to those used to predict the pellets form the ADU process. In order to evaluate how different the pellets originated from these two industrial routes are, this paper aims to compare the INB production historical data (Angra 1, Cycles 14 and 15) with the key parameters of a common product specification from the ADU process. (author)

Dissolution of UO2 in redox conditions

International Nuclear Information System (INIS)

Casas, I.; Pablo de, J.; Rovira, M.

1998-01-01

The performance assessment of the final disposal of the spent nuclear fuel in geological formations is strongly dependent on the spent fuel matrix dissolution. Unirradiated uranium (IV) dioxide has shown to be very useful for such purposes. The stability of UO 2 is very dependent on vault redox conditions. At reducing conditions, which are expected in deep groundwaters, the dissolution of the UO 2 -matrix can be explained in terms of solubility, while under oxidizing conditions, the UO 2 is thermodynamically unstable and the dissolution is kinetically controlled. In this report the parameters which affect the uranium solubility under reducing conditions, basically pH and redox potential are discussed. Under oxidizing conditions, UO 2 dissolution rate equations as a function of pH, carbonate concentration and oxidant concentration are reported. Dissolution experiments performed with spent fuel are also reviewed. The experimental equations presented in this work, have been used to model independent dissolution experiments performed with both unirradiated and irradiated UO 2 . (Author)

Molybdenum-UO2 cerment irradiation at 1145 K

Science.gov (United States)

Mcdonald, G.

1971-01-01

Two molybdenum-UO2 cermet fuel pins were fission heated in a helium-cooled loop at a temperature of 1145 K and to a total burnup of 5.3 % of the U-235. After irradiation the fuel pins were measured to check dimensional stability, punctured at the plenums to determine fission gas release, and examined metallographically to determine the effect of irradiation. Burnup was determined in several sections of the fuel pin. The results of the postirradiation examination indicated: (1) There was no visible change in the fuel pins on irradiation under the above conditions. (2) The maximum swelling of the fuel pins was less than 1%. (3) There was no migration of UO2 and no visible interaction between the molybdenum and the UO2. (4) Approximately 12% of the fission gas formed was released from the cermet cone into the gas plenum.

Modernization of control instrumentation and security of reactor IAN - R1

International Nuclear Information System (INIS)

Gonzalez, J. M.

1993-01-01

The program to modernize IAN-R1 research reactor control and safety instrumentation has been carried out considering two main aspects: updating safety philosophy requirements and acquiring the newest reactor control instrumentation controlled by computer, following the present criteria internationally recognized, for safety and reliable reactor operations and the latest developments of nuclear electronic technology. The new IAN-R1 reactor instrumentation consist of two wide range neutron monitoring channels, commanded by microprocessor a data acquisition system and reactor control, (controlled by computers). The reactor control desk is providing through two displays; all safety and control signals to the reactor operators; furthermore some signals like reactor power, safety and period signals are also showed on digital bar graphics, which are hard wired directly from the neutron monitoring channels

Contribution to the identification and the evaluation of a doped UO2 fuel with controlled oxygen potential

International Nuclear Information System (INIS)

Pennisi, Vanessa

2015-01-01

Temperature and oxygen partial pressure (PO 2 ) of nuclear oxide fuels are the main parameters governing both their thermochemical evolution in reactor and the speciation of volatile fission products such as Cs, I or Te. An innovative way to limit the risk of cladding rupture by corrosion under irradiation consists in buffering the oxygen partial pressure of the fuel under operation in a PO 2 domain where the fission gas are harmless towards Zr clad, by using solid redox buffers as additives. Niobium, with its NbO 2 /NbO and Nb 2 O 5 /NbO 2 redox couples has been found to be a promising candidate to this end. A manufacturing process of a buffered UO 2 fuel, doped with niobium has been optimized, in order to fulfill usual specifications (density, microstructure). The experimental study of the UO 2 -NbO x system has shown the existence of a liquid phase between UO 2 and NbO x at 810 C, which was not reported in the literature. The characterization of Nb containing phases present in UO 2 both in solid solution and as precipitates has lead us to propose a solubility thermodynamic model of niobium in UO 2 at 1700 C. An extensive study of the niobium precipitates shows the co-existence in the fuel of NbO 2 and NbO as major phases, together with small amounts of metallic Nb. The coexistence of niobium under two oxidation states inside the fuel is a key element of demonstration of a possible in-situ buffering effect, which is likely to impact some properties of the material that are dependent upon PO 2 , such as densification. These results confirm the promising potential of oxygen buffered fuels as regard to their performance in reactor. (author) [fr

Synthesis and crystal structure of Na6[(UO2)3O(OH)3(SeO4)2]2·10H2O

International Nuclear Information System (INIS)

Baeva, E.Eh.; Serezhkina, L.B.; Virovets, A.V.; Peresypkina, E.V.

2006-01-01

The complex Na 6 [(UO 2 ) 3 O(OH) 3 (SeO 4 ) 2 ] 2 ·10H 2 O (I) is synthesized and studied by monocrystal X-ray diffraction. The compound crystallizes in the orthorhombic crystal system with the unit cell parameters: a=14.2225(7) A, b=18.3601(7) A, c=16.5406(6) A, V=4319.2(3) A 3, Z=4, space group Cmcm, R 1 =0.0406. Compound I is found to be a representative of the crystal-chemical group A 3 M 3 M 3 2 T 2 3 (A=UO 2 2+ , M 3 =O 2- , M 2 =OH - , T 3 =SeO 4 2- ) of the uranyl complexes; it contains layer uranium-containing groups [(UO 2 ) 3 O(OH) 3 (SeO 4 ) 2 ] 3- . These layers are linked to form a three-dimensional cage through bonds formed by the sodium atoms with the oxygen atoms of the uranyl ions and SeO 4 groups that belong to different layers [ru

Photochemical synthesis of UO2 nanoparticles

International Nuclear Information System (INIS)

Rath, M.C.; Keny, Sangeeta; Naik, D.B.

2014-01-01

UO 2 nanoparticles have been recently synthesized by us from aqueous solutions of uranyl nitrate through radiolytic method on high-energy electron beam irradiation. In this study, the synthesis of UO 2 nanoparticles through photochemical method is reported which is a complementary route to radiation chemical method

Development on UO3-K2O and MoO3-K2O binary systems and study of UO2MoO4-MoO3 domain within UO3-MoO3-K2O ternary system

International Nuclear Information System (INIS)

Dion, C.; Noel, A.

1983-01-01

This paper confirms the previous study on the MoO 3 -K 2 O system, and constitutes a clarity of the UO 3 -K 2 O system. Four distinct uranates VI with alkaline metal/uranium ratio's 2, 1, 0,5 and 0,285 exist. Preparation conditions and powder diffraction spectra of these compounds are given. Additional informations relative to K 2 MoO 4 allotropic transformations are provided. Study of UO 2 MoO 4 -K 2 MoO 4 diagram has brought three new phases into prominence: (B) K 6 UMo 4 O 18 incongruently melting point, (E) K 2 UMo 2 O 10 congruently melting and (F) K 2 U 3 Mo 4 O 22 incongruently melting point. Within MoO 3 -K 2 MoO 4 -UO 2 MoO 4 ternary system, no new phase is found. The general appearance of ternary liquidus and crystallization fields of several compounds are given. These three new compounds become identified with these of UO 2 MoO 4 -Na 2 MoO 4 binary system [fr

Thermal hydraulic analysis of the IPR-R1 TRIGA reactor; Analise termo-hidraulica do reator TRIGA IPR-R1

Energy Technology Data Exchange (ETDEWEB)

Veloso, Marcelo Antonio [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), Belo Horizonte, MG (Brazil); Fortini, Maria Auxiliadora [Minas Gerais Univ., Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear

2002-07-01

The subchannel approach, normally employed for the analysis of power reactor cores that work under forced convection, have been used for the thermal hydraulic evaluation of a TRIGA Mark I reactor, named IPR-R1, at 250 kW power level. This was accomplished by using the PANTERA-1P subchannel code, which has been conveniently adapted to the characteristics of natural convection of TRIGA reactors. The analysis of results indicates that the steady state operation of IPR-R1 at 250 kW do not imply risks to installations, workers and public. (author)

High temperature drop calorimetric studies on La6UO12 and Nd6UO12

International Nuclear Information System (INIS)

Babu, R.; Senapati, A.; Rao, G.J.; Venkata Krishnan, R.; Ananthasivan, K.; Nagarajan, K.

2014-01-01

Rare earth elements produced in the reactor during irradiation can interact with the fuel. Under transient conditions, compounds of formula, RE 6 UO 12 with rhombohedral crystal structure are expected to be formed. Hence, thermodynamic properties of these compounds are useful in interpreting the behaviour of fuels during irradiation. Thermal expansion and heat capacities by DSC have been reported for La 6 UO 12 and Nd 6 UO 12 . There are no experimentally measured values of enthalpy. Hence, measurements on enthalpy increments of La 6 UO 12 and Nd 6 UO 12 were carried out for the first time by inverse drop calorimetry in the temperature range 534-1738 K and computed the thermodynamic functions

Structural and thermodynamic characterization of the perovskite-related BA{sub 1+y}UO{sub 3+x} and (BA,SR){sub 1+y}UO{sub 3+x} phases

Energy Technology Data Exchange (ETDEWEB)

Cordfunke, E.H.P. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Booij, A.S. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Smit-Groen, V. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Vlaanderen, P. van [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); IJdo, D.J.W. [Rijksuniversiteit Leiden (Netherlands). Gorlaeus Labs.

1996-11-01

The perovskite-type BaUO{sub 3} structure has been investigated by X-ray and neutron diffraction. The Ba/U ratio, the (Ba, Sr)/U ratio, and the oxygen stoichiometry in Ba{sub 1+y}UO{sub 3+x} were varied, and the integral enthalpies of formation determined by solution calorimetry. In addition, equilibrium oxygen partial pressures were measured using a reversible EMF cell. The chemical defect mechanism is discussed, and it is shown that a continuous series BaUO{sub 3}-Ba{sub 1+y}UO{sub 3+x}-Ba{sub 3}UO{sub 6} exist in which uranium vacancies are gradually filled up with barium ions, whereas uranium is oxidized via the pentavalent to the hexavalent state in Ba{sub 3}UO{sub 6}(=Ba{sub 2}(Ba, U)O{sub 6}). (orig.).

Present status of reactor physics in the United States and Japan-IV. 2. Micro-Reactor Physics of MOX-Fueled Core

International Nuclear Information System (INIS)

Takeda, Toshikazu

2001-01-01

Recently, fuel assemblies of light water reactors have become complicated because of the extension of fuel burnup and the use of high-enriched Gd and mixed-oxide (MOX) fuel, etc. In conventional assembly calculations, the detailed flux distribution, spectrum distribution, and space dependence of self-shielding within a fuel pellet are not directly taken into account. The experimental and theoretical study of investigating these microscopic properties is named micro-reactor physics. The purpose of this work is to show the importance of micro-reactor physics in the analysis of MOX fuel assemblies. Several authors have done related studies; however, their studies are limited to fuel pin cells, and they are never mentioned with regard to burnup effect, which is important for actual core design. We used the subgroup method to treat the space dependence of the self-shielding effect of heavy nuclides, and we used the characteristics method to treat the angular dependence of neutron flux in a fuel pellet. Figure 1 compares the power distributions in MOX and UO 2 fuel cells at the beginning of burnup. The power is calculated with and without considering the space dependence of the self-shielding effect of the cross sections. For the MOX cell, the power distribution has a peak at the cell edge because of large Pu absorption especially when considering the spatial self-shielding effect. When a MOX rod is adjacent to UO 2 fuel rods, the flux distribution has an azimuthal dependence in addition to the radial dependence within a rod. For example, consider a 2x2 fuel assembly composed of three UO 2 rods and one MOX rod, with the mirror reflection boundary condition. A burnup calculation was done with the condition; the radius of the MOX pellet is divided into two regions, and the azimuthal angle is divided into eight. The number density of 239 Pu at 44 000 MWd/t for the MOX rod shows azimuthal dependence by 20%. The maximum burnup occurs in the direction of the UO 2 rods. This is

A thermal hydraulic analysis in PWR reactors with UO{sub 2} or (U-Th)O{sub 2} fuel rods employing a simplified code

Energy Technology Data Exchange (ETDEWEB)

Santos, Thiago A. dos; Maiorino, José R., E-mail: thiago.santos@ufabc.edu.br, E-mail: joserubens.maiorino@ufabc.edu.br [Universidade Federal do ABC (UFABC), Santo André, SP (Brazil); Stefanni, Giovanni L. de, E-mail: giovanni.stefanni@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil)

2017-07-01

In order to project a nuclear reactor, the neutronic calculus must be validated, so that its thermal limits and safety parameters are respected. Considering this issue, this research aims to evaluate the APTh-100 reactor thermal limits. This PWR is a project developed in Universidade Federal do ABC (UFABC) using fuel composed of Uranium and Thorium oxide mixed (U,Th)O{sub 2}. For this purpose, a simplified, although conservative, code was developed in a MATLAB environment named STC-MOX-Th 'Simplified Thermal-hydraulics Code-Mixed Oxide Thorium'. This code provides axial and radial temperature distribution, as well as DNBR distribution over the hottest channel of the reactor core. Moreover, it brings other hydraulic quantities, such as pressure drop over the fuel rod, considering any fuel proportion of (U,Th)O{sub 2}.The software uses basic laws of conservation of mass, momentum and energy, it also calculates the thermal conduction equation, considering the thermal conductive coefficient as a temperature function. In order to solve this equation, the finite elements method was used. Furthermore, the proportion of 36% of UO{sub 2} was used to evaluate the temperature over the fuel rod and DNBR minimum in three burn conditions: beginning, middle and ending. The program has proven to be efficient in every condition and the results evidenced that the APTh-1000 reactor, in an initial analysis, has its thermal limits within the recommended security parameters. (author)

Separation of UO2 powder

International Nuclear Information System (INIS)

Ristic, M.M.

1962-01-01

This report deals with theoretical approach to separation process and describes the constructed separator with liquid medium. The separator was calibrated and tested with Al 3 O 3 and UO 2 . it has been concluded that it can be used for separation of powders with sufficient accuracy if the separation is performed for a longer period of time. The separated fractions were characterised by microscopic method and the UO 2 fraction additionally by sedimentation method

Oxidation kinetic changes of UO2 by additive addition and irradiation

International Nuclear Information System (INIS)

You, Gil-Sung; Kim, Keon-Sik; Min, Duck-Kee; Ro, Seung-Gy

2000-01-01

The kinetic changes of air-oxidation of UO 2 by additive addition and irradiation were investigated. Several kinds of specimens, such as unirradiated-UO 2 , simulated-UO 2 for spent PWR fuel (SIMFUEL), unirradiated-Gd-doped UO 2 , irradiated-UO 2 and -Gd-doped UO 2 , were used for these experiments. The oxidation results represented that the kinetic patterns among those samples are remarkably different. It was also revealed that the oxidation kinetics of irradiated-UO 2 seems to be more similar to that of unirradiated-Gd-doped UO 2 than that of SIMFUEL

BR2 Reactor: Irradiation of fuels

International Nuclear Information System (INIS)

Verwimp, A.

2005-01-01

Safe, reliable and economical operation of reactor fuels, both UO 2 and MOX types, requires in-pile testing and qualification up to high target burn-up levels. In-pile testing of advanced fuels for improved performance is also mandatory. The objectives of research performed at SCK-CEN are to perform Neutron irradiation of LWR (Light Water Reactor) fuels in the BR2 reactor under relevant operating and monitoring conditions, as specified by the experimenter's requirements and to improve the on-line measurements on the fuel rods themselves

Crystal structure of [UO2(NH35]NO3·NH3

Directory of Open Access Journals (Sweden)

Patrick Woidy

2016-12-01

Full Text Available Pentaammine dioxide uranium(V nitrate ammonia (1/1, [UO2(NH35]NO3·NH3, was obtained in the form of yellow crystals from the reaction of caesium uranyl nitrate, Cs[UO2(NO33], and uranium tetrafluoride, UF4, in dry liquid ammonia. The [UO2]+ cation is coordinated by five ammine ligands. The resulting [UO2(NH35] coordination polyhedron is best described as a pentagonal bipyramid with the O atoms forming the apices. In the crystal, numerous N—H...N and N—H...O hydrogen bonds are present between the cation, anion and solvent molecules, leading to a three-dimensional network.

A new UO2 sintering technology for the recycling of defective fuel pellets

International Nuclear Information System (INIS)

Song, K. W.; Kim, K. S.; Jeong, Y. H.

1998-01-01

A new UO 2 sintering technology to recycle defective UO 2 pellets has been developed. The defective UO 2 pellets were oxidized in an air to produce U 3 O 8 powder, and the U 3 O 8 powder was mixed with fresh AUC-UO 2 powder in the range of 10 to 100 wt%. Nb 2 O 5 and TiO 2 are added to the mixed powder. The mixed powder was pressed and sintered at 1680 deg C for 4 hours in hydrogen. The density of UO 2 pellets without sintering agents decreased linearly with the U 3 O 8 content at the rate of 0.2 %TD per 1 wt% U 3 O 8 , and the density was below 93.5 %TD at the U 3 O 8 contents above 10 wt%. However, the mixed UO 2 and U 3 O 8 powder containing Nb 2 O 5 (≥0.3 wt%) and TiO 2 (≥0.1 wt%) yielded a sintered density above 94 %TD in all ranges of U 3 O 8 contents. It was found that higher mixing ratios of U 3 O 8 to UO 2 powder did not affect the grain size of UO 2 pellets under the addition of Nb 2 O 5 , but decreased the grain size of UO 2 pellets under the addition of TiO 2 . The doped UO 2 pellets have grain sizes larger than 20 μm, and have small density gain after re-sintering test, owing to large pores. Therefore, the sintering agents such as Nb 2 O 5 and TiO 2 can make highly densified UO 2 pellets from the powder comprising a large amount of U 3 O 8 powder

Possible effects of oxidation on the transient release of fission gas from UO2

International Nuclear Information System (INIS)

Stoner, H.C.; Matthews, J.R.; Wood, M.H.

1981-01-01

The effect of varying the fuel composition from UO 2 to UOsub(2.3), on the transient behaviour of fission gas is simulated on the assumption that surface diffusion behaves in a similar manner to volume diffusion. The results may help in the understanding of fuel behaviour after pin failure in accident conditions in thermal reactor systems. (author)

Nitrate conversion and supercritical fluid extraction of UO2-CeO2 solid solution prepared by an electrolytic reduction-coprecipitation method

International Nuclear Information System (INIS)

Zhu, L.Y.; Duan, W.H.; Wen, M.F.; Xu, J.M.; Zhu, Y.J.

2014-01-01

A low-waste technology for the reprocessing of spent nuclear fuel (SNF) has been developed recently, which involves the conversion of actinide and lanthanide oxides with liquid N 2 O 4 into their nitrates followed by supercritical fluid extraction of the nitrates. The possibility of the reprocessing of SNF from high-temperature gas-cooled reactors (HTGRs) with nitrate conversion and supercritical fluid extraction is a current area of research in China. Here, a UO 2 -CeO 2 solid solution was prepared as a surrogate for a UO 2 -PuO 2 solid solution, and the recovery of U and Ce from the UO 2 -CeO 2 solid solution with liquid N 2 O 4 and supercritical CO 2 containing tri-n-butyl phosphate (TBP) was investigated. The UO 2 -CeO 2 solid solution prepared by electrolytic reduction-coprecipitation method had square plate microstructures. The solid solution after heat treatment was completely converted into nitrates with liquid N 2 O 4 . The XRD pattern of the nitrates was similar to that of UO 2 (NO 3 ) 2 . 3H 2 O. After 120 min of online extraction at 25 MPa and 50 , 99.98% of the U and 98.74% of the Ce were recovered from the nitrates with supercritical CO 2 containing TBP. The results suggest a promising potential technology for the reprocessing of SNF from HTGRs. (orig.)

Arachne - A web-based event viewer for MINERvA

International Nuclear Information System (INIS)

Tagg, N.; Brangham, J.; Chvojka, J.; Clairemont, M.; Day, M.; Eberly, B.; Felix, J.; Fields, L.; Gago, A.M.; Gran, R.; Harris, D.A.

2011-01-01

Neutrino interaction events in the MINERvA detector are visually represented with a web-based tool called Arachne. Data are retrieved from a central server via AJAX, and client-side JavaScript draws images into the user's browser window using the draft HTML 5 standard. These technologies allow neutrino interactions to be viewed by anyone with a web browser, allowing for easy hand-scanning of particle interactions. Arachne has been used in MINERvA to evaluate neutrino data in a prototype detector, to tune reconstruction algorithms, and for public outreach and education.

Arachne - A web-based event viewer for MINERvA

Energy Technology Data Exchange (ETDEWEB)

Tagg, N.; /Otterbein Coll.; Brangham, J.; /Otterbein Coll.; Chvojka, J.; /Rochester U.; Clairemont, M.; /Otterbein Coll.; Day, M.; /Rochester U.; Eberly, B.; /Pittsburgh U.; Felix, J.; /Guanajuato U.; Fields, L.; /Northwestern U.; Gago, A.M.; /Lima, Pont. U. Catolica; Gran, R.; /Maryland U.; Harris, D.A.; /Fermilab /William-Mary Coll.

2011-11-01

Neutrino interaction events in the MINERvA detector are visually represented with a web-based tool called Arachne. Data are retrieved from a central server via AJAX, and client-side JavaScript draws images into the user's browser window using the draft HTML 5 standard. These technologies allow neutrino interactions to be viewed by anyone with a web browser, allowing for easy hand-scanning of particle interactions. Arachne has been used in MINERvA to evaluate neutrino data in a prototype detector, to tune reconstruction algorithms, and for public outreach and education.

The compaction and sintering of UO_2-Zr cermet pellets

International Nuclear Information System (INIS)

Tri Yulianto; Meniek Rachmawati; Etty Mutiara

2013-01-01

An innovative fuel pellet of UO_2-Zr cermet has been developed to improve thermal conductivity of UO_2 pellet by adding small amount Zr metal in to UO_2 matrix below 10 % weight. Zirconium powder will serve for the creation of bridges or web structure during compaction and will effectively reduce contact between of UO_2 particles. Based on the theory of phase equilibrium of metals-metal oxides-ceramic, this fabrication technique may produce UO_2 pellets containing continuous metal channel on the grain boundary of UO_2 through sintering in a reduction atmosphere. The fabrication was done by varying process parameters of mixing and compaction. Characterisation of UO_2-Zr cermet pellet involved visual test, dimensional and density measurement, and ceramography test. This advanced cermet fabrication technology may address common issue with cermet fuels such as microstructure with continuous metal channel structure in the UO_2 matrix, which is more effectively than the commonly accepted microstructure involving fraction of UO_2 pellet by standard fabrication route. (author)

Fabrication of nano-structured UO2 fuel pellets

International Nuclear Information System (INIS)

Yang, Jae Ho; Kang, Ki Won; Rhee, Young Woo; Kim, Dong Joo; Kim, Jong Heon; Kim, Keon Sik; Song, Kun Woo

2007-01-01

Nano-structured materials have received much attention for their possibility for various functional materials. Ceramics with a nano-structured grain have some special properties such as super plasticity and a low sintering temperature. To reduce the fuel cycle costs and the total mass of spent LWR fuels, it is necessary to extend the fuel discharged burn-up. In order to increase the fuel burn-up, it is important to understand the fuel property of a highly irradiated fuel pellet. Especially, research has focused on the formation of a porous and small grained microstructure in the rim area of the fuel, called High Burn-up Structure (HBS). The average grain size of HBS is about 300nm. This paper deals with the feasibility study on the fabrication of nano-structured UO 2 pellets. The nano sized UO 2 particles are prepared by a combined process of a oxidation-reducing and a mechanical milling of UO 2 powder. Nano-structured UO 2 pellets (∼300nm) with a density of ∼93%TD can be obtained by sintering nano-sized UO 2 compacts. The SEM study reveals that the microstructure of the fabricated nano-structure UO 2 pellet is similar to that of HBS. Therefore, this bulk nano-structured UO 2 pellet can be used as a reference pellet for a measurement of the physical properties of HBS

Decommissioning and decontrolling the R1-reactor

International Nuclear Information System (INIS)

Bergman, C.; Holmberg, B.T.

1985-01-01

Sweden's first nuclear reactor - the research reactor R1 - situated in bedrock under the Royal Technical Institute of Stockholm, has in the period 1981-1983 been subject to a complete decommissioning. The National Institute for Radiation Protection has followed the work in detail, and has after the completion of the decommissioning performed measurements of radioactivity on site. The report gives an account of the work the Institute has done in preparation for- and during decommissioning and specifically report on the measurements for classification of the local as free for non-nuclear use. (aa)

Perovskite phases in the systems AO-SE/sub 2/O/sub 3/-UO/sub 2,x/ with A=alkaline earth metal and SE=rare earths, La, and Y. VII. The systems Ba/sub 2/CaUO/sub 6/-Ba/sub 2/Gd/sub 0. 67/UO/sub 6/ and Ba/sub 2/CaUO/sub 6/-Ba/sub 2/Y/sub 0. 67/UO/sub 6/

Energy Technology Data Exchange (ETDEWEB)

Kemmler-Sack, S; Seemann, I; Schittenhelm, H J [Tuebingen Univ. (F.R. Germany). Institut fuer Anorganische Chemie

1976-05-01

The ordered perovskite Ba/sub 2/CaUO/sub 6/ forms a solid solution series with Ba/sub 2/Gdsub(0.67)UO/sub 6/ and Ba/sub 2/Ysub(0.67)UO/sub 6/, respectively. The deviations from the ideal behaviour are studied by X-ray, diffuse reflectance and vibrational methods.

Perovskite phases in the systems AO-SE/sub 2/O/sub 3/-UO/sub 2,x/ with A=alkaline earth metal and SE=rare earths, La, and Y. IX. The systems Ba/sub 2/SrUO/sub 6/-Ba/sub 2/Gd/sub 0. 67/UO/sub 6/ and Ba/sub 2/SrUO/sub 6/-Ba/sub 2/Y/sub 0. 67/UO/sub 6/

Energy Technology Data Exchange (ETDEWEB)

Kemmler-Sack, S; Seemann, I [Tuebingen Univ. (F.R. Germany). Inst. fuer Anorganische Chemie I

1976-07-01

The ordered perovskite Ba/sub 2/SrUO/sub 6/ forms a solid solution series with Ba/sub 2/Gdsub(0.67)UO/sub 6/ and Ba/sub 2/Ysub(0.67)UO/sub 6/ respectively. The deviations from the ideal behaviour are studied by X-ray, diffuse reflectance and vibrational methods.

Cracking and relocation of UO2 fuel during nuclear operation

International Nuclear Information System (INIS)

Appelhans, A.D.; Dagbjartsson, S.J.

1981-01-01

Cracking and relocation of light water reactor (LWR) fuel pellets affect the axial gas flow path within nuclear reactor fuel rods and the thermal performance of the fuel. As part of the Nuclear Regulatory Commission's Water Reactor Safety Research Fuel Behavior Program, the Thermal Fuels Behavior Program of EG and G Idaho, Inc., is conducting fuel rod behavior studies in the Heavy Boiling Water Reactor in Halden, Norway. The Instrumental Fuel Assembly-430 (IFA-430) operated in that facility is a multipurpose assembly designed to provide information on fuel cracking and relocation, the long-term thermal response of LWR fuel rods subjected to various internal pressures and gas compositions, and the release of fission gases. This report presents the results of an analysis of fuel cracking and relocation phenomena as deduced from fuel rod axial gas flow and fuel temperature data from the first 6.5 GWd/tUO 2 burnup of the IFA-430

New instrumentation for the IPR-R1 reactor of CDTN

International Nuclear Information System (INIS)

Carvalho, P.V.R. de.

1992-01-01

The Nuclear Engineering Institute reactor instrumentation area has developed systems and equipment for reactor operation and safety. In such way, the new I and C for IEN Argonauta reactor and the nuclear instrumentation for IPEN critical facility were built. This paper describes our real work, the new I and C systems for IPR-R1, a Triga type reactor, located at CDTN (Belo Horizonte - MG). (author)

High density UO2 powder preparation for HWR fuel

International Nuclear Information System (INIS)

Hwang, S. T.; Chang, I. S.; Choi, Y. D.; Cho, B. R.; Kwon, S. W.; Kim, B. H.; Moon, B. H.; Kim, S. D.; Phyu, K. M.; Lee, K. A.

1992-01-01

The objective of this project is to study on the preparation of method high density UO 2 powder for HWR Fuel. Accordingly, it is necessary to character ize the AUC processed UO 2 powder and to search method for the preparation of high density UO 2 powder for HWR Fuel. Therefore, it is expected that the results of this study can effect the producing of AUC processed UO 2 powder having sinterability. (Author)

Kinetics of UO2 sintering

International Nuclear Information System (INIS)

Ristic, M.M.

1962-01-01

Detailed conclusions related to the UO 2 sintering can be drawn from investigating the kinetics of the sintering process. This report gives an thorough analysis of the the data concerned with sintering available in the literature taking into account the Jander and Arrhenius laws. This analysis completes the study of influence of the O/U ratio and the atmosphere on the sintering. Results presented are fundamentals of future theoretical and experimental work related to characterisation of the UO 2 sintering process

Production of Depleted UO2Kernels for the Advanced Gas-Cooled Reactor Program for Use in TRISO Coating Development

International Nuclear Information System (INIS)

Collins, J.L.

2004-01-01

The main objective of the Depleted UO 2 Kernels Production Task at Oak Ridge National Laboratory (ORNL) was to conduct two small-scale production campaigns to produce 2 kg of UO 2 kernels with diameters of 500 ± 20 (micro)m and 3.5 kg of UO 2 kernels with diameters of 350 ± 10 (micro)m for the U.S. Department of Energy Advanced Fuel Cycle Initiative Program. The final acceptance requirements for the UO 2 kernels are provided in the first section of this report. The kernels were prepared for use by the ORNL Metals and Ceramics Division in a development study to perfect the triisotropic (TRISO) coating process. It was important that the kernels be strong and near theoretical density, with excellent sphericity, minimal surface roughness, and no cracking. This report gives a detailed description of the production efforts and results as well as an in-depth description of the internal gelation process and its chemistry. It describes the laboratory-scale gel-forming apparatus, optimum broth formulation and operating conditions, preparation of the acid-deficient uranyl nitrate stock solution, the system used to provide uniform broth droplet formation and control, and the process of calcining and sintering UO 3 · 2H 2 O microspheres to form dense UO 2 kernels. The report also describes improvements and best past practices for uranium kernel formation via the internal gelation process, which utilizes hexamethylenetetramine and urea. Improvements were made in broth formulation and broth droplet formation and control that made it possible in many of the runs in the campaign to produce the desired 350 ± 10-(micro)m-diameter kernels, and to obtain very high yields

Irradiation routine in the IPR-R1 Triga reactor

International Nuclear Information System (INIS)

Maretti Junior, F.

1980-01-01

Information about irradiations in the IPR-R1 TRIGA reactor and procedures necessary for radioisotope solicitation are presented All procedures necessary for asking irradiation in the reactor, shielding types, norms of terrestrial and aerial expeditions, payment conditions, and catalogue of disposable isotopes with their respective saturation activities are described. (M.C.K.)

Development of irradiated UO2 thermal conductivity model

International Nuclear Information System (INIS)

Lee, Chan Bock; Bang Je-Geon; Kim Dae Ho; Jung Youn Ho

2001-01-01

Thermal conductivity model of the irradiated UO 2 pellet was developed, based upon the thermal diffusivity data of the irradiated UO 2 pellet measured during thermal cycling. The model predicts the thermal conductivity by multiplying such separate correction factors as solid fission products, gaseous fission products, radiation damage and porosity. The developed model was validated by comparison with the variation of the measured thermal diffusivity data during thermal cycling and prediction of other UO 2 thermal conductivity models. Since the developed model considers the effect of gaseous fission products as a separate factor, it can predict variation of thermal conductivity in the rim region of high burnup UO 2 pellet where the fission gases in the matrix are precipitated into bubbles, indicating that decrease of thermal conductivity by bubble precipitation in rim region would be significantly compensated by the enhancing effect of fission gas depletion in the UO 2 matrix. (author)

Microstructure study of AUC and UO2

International Nuclear Information System (INIS)

Pan Ying; Gao Dihua; Lu Huaichang

1992-01-01

The microstructures of AUC, UO 2 powder and pellets were investigated with metallo-scope, SEM, TEM, XRD, and image analyzer. The influence of the reduction conditions of AUC on the microstructures of UO 2 powder and pellet were studied

AGR-2 Irradiation Test Final As-Run Report, Rev 2

Energy Technology Data Exchange (ETDEWEB)

Collin, Blaise P. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2014-08-01

This document presents the as-run analysis of the AGR-2 irradiation experiment. AGR-2 is the second of the planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the U.S. Department of Energy as part of the Very High Temperature Reactor (VHTR) Technical Development Office (TDO) program. The objectives of the AGR-2 experiment are to: (a) Irradiate UCO (uranium oxycarbide) and UO₂ (uranium dioxide) fuel produced in a large coater. Fuel attributes are based on results obtained from the AGR-1 test and other project activities. (b) Provide irradiated fuel samples for post-irradiation experiment (PIE) and safety testing. (c) Support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. The primary objective of the test was to irradiate both UCO and UO₂ TRISO (tri-structural isotropic) fuel produced from prototypic scale equipment to obtain normal operation and accident condition fuel performance data. The UCO compacts were subjected to a range of burnups and temperatures typical of anticipated prismatic reactor service conditions in three capsules. The test train also includes compacts containing UO₂ particles produced independently by the United States, South Africa, and France in three separate capsules. The range of burnups and temperatures in these capsules were typical of anticipated pebble bed reactor service conditions. The results discussed in this report pertain only to U.S. produced fuel. In order to achieve the test objectives, the AGR-2 experiment was irradiated in the B-12 position of the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) for a total irradiation duration of 559.2 effective full power days (EFPD). Irradiation began on June 22, 2010, and ended on October 16, 2013, spanning 12 ATR power cycles and approximately three and a

Neutronics experimental validation of the Jules Horowitz reactor fuel by interpretation of the VALMONT experimental program-transposition of the uncertainties on the reactivity of JHR with JEF2.2 and JEFF3.1.1

International Nuclear Information System (INIS)

Leray, O.; Hudelot, J.P.; Doederlein, C.; Vaglio-Gaudard, C.; Antony, M.; Santamarina, A.; Bernard, D.

2012-01-01

The new European material testing Jules Horowitz Reactor (JHR), currently under construction in Cadarache center (CEA France), will use LEU (20% enrichment in 235 U) fuels (U 3 Si 2 for the start up and UMoAl in the future) which are quite different from the industrial oxide fuel, for which an extensive neutronics experimental validation database has been established. The HORUS3D/N neutronics calculation scheme, used for the design and safety studies of the JHR, is being developed within the framework of a rigorous verification-numerical validation-experimental validation methodology. In this framework, the experimental VALMONT (Validation of Aluminium Molybdenum uranium fuel for Neutronics) program has been performed in the MINERVE facility of CEA Cadarache (France), in order to qualify the capability of HORUS3D/N to accurately calculate the reactivity of the JHR reactor. The MINERVE facility using the oscillation technique provides accurate measurements of reactivity effect of samples. The VALMONT program includes oscillations of samples of UAl ∞ /Al and UMo/Al with enrichments ranging from 0.2% to 20% and Uranium densities from 2.2 to 8 g/cm 3 . The geometry of the samples and the pitch of the experimental lattice ensure maximum representativeness with the neutron spectrum expected for JHR. By comparing the effect of the sample with the one of a known fuel specimen, the reactivity effect can be measured in absolute terms and be compared to computational results. Special attention was paid to the rigorous determination and reduction of the experimental uncertainties. The calculational analysis of the VALMONT results was performed with the French deterministic code APOLLO2. A comparison of the impact of the different calculation methods, data libraries and energy meshes that were tested is presented. The interpretation of the VALMONT experimental program allowed the experimental validation of JHR fuel UMoAl8 (with an enrichment of 19.75% 235 U) by the Minerve

Nuclear Physics with MINERvA

International Nuclear Information System (INIS)

Tice, Brian G.

2011-01-01

MINERvA is a precision neutrino experiment designed to improve our understanding of the neutrino-nucleus interaction. The experiment uses a fully active scintillation detector to allow full event reconstruction and includes nuclear targets helium, water, carbon, iron and lead. Here we describe the first steps in measuring lead to iron to carbon cross section ratios.

Solution of a benchmark set problems for BWR and PWR reactors with UO2 and MOX fuels using CASMO-4

International Nuclear Information System (INIS)

Martinez F, M.A.; Valle G, E. del; Alonso V, G.

2007-01-01

In this work some of the results for a group of benchmark problems of light water reactors that allow to study the physics of the fuels of these reactors are presented. These benchmark problems were proposed by Akio Yamamoto and collaborators in 2002 and they include two fuel types; uranium dioxide (UO 2 ) and mixed oxides (MOX). The range of problems that its cover embraces three different configurations: unitary cell for a fuel bar, fuel assemble of PWR and fuel assemble of BWR what allows to carry out an understanding analysis of the problems related with the fuel performance of new generation in light water reactors with high burnt. Also these benchmark problems help to understand the fuel administration in core of a BWR like of a PWR. The calculations were carried out with CMS (of their initials in English Core Management Software), particularly with CASMO-4 that is a code designed to carry out analysis of fuels burnt of fuel bars cells as well as fuel assemblies as much for PWR as for BWR and that it is part in turn of the CMS code. (Author)

Design, calibration, and performance of the MINERvA detector

Energy Technology Data Exchange (ETDEWEB)

Aliaga, L. [Department of Physics, College of William and Mary, Williamsburg, VA 23187 (United States); Sección Física, Departamento de Ciencias, Pontificia Universidad Católica del Perú, Apartado 1761, Lima, Perú (Peru); Bagby, L.; Baldin, B. [Fermi National Accelerator Laboratory, Batavia, IL 60510 (United States); Baumbaugh, A. [Sección Física, Departamento de Ciencias, Pontificia Universidad Católica del Perú, Apartado 1761, Lima, Perú (Peru); Bodek, A.; Bradford, R. [University of Rochester, Rochester, NY 14610 (United States); Brooks, W.K. [Departamento de Física, Universidad Técnica Federico Santa María, Avda. España 1680, Casilla 110-V, Valparaíso (Chile); Boehnlein, D. [Fermi National Accelerator Laboratory, Batavia, IL 60510 (United States); Boyd, S. [Department of Physics and Astronomy, University of Pittsburgh, Pittsburgh, PA 15260 (United States); Budd, H. [University of Rochester, Rochester, NY 14610 (United States); Butkevich, A. [Institute for Nuclear Research of the Russian Academy of Sciences, 117312 Moscow (Russian Federation); Martinez Caicedo, D.A.; Castromonte, C.M. [Hampton University, Department of Physics, Hampton, VA 23668 (United States); Christy, M.E. [Department of Physics, University of Minnesota – Duluth, Duluth, MN 55812 (United States); Chvojka, J. [University of Rochester, Rochester, NY 14610 (United States); Motta, H. da [Centro Brasileiro de Pesquisas Físicas, Rua Dr. Xavier Sigaud 150, Urca, Rio de Janeiro, RJ 22290-180 (Brazil); and others

2014-04-11

The MINERvA experiment is designed to perform precision studies of neutrino-nucleus scattering using ν{sub μ} and ν{sup ¯}{sub μ} neutrinos incident at 1–20 GeV in the NuMI beam at Fermilab. This article presents a detailed description of the MINERvA detector and describes the ex situ and in situ techniques employed to characterize the detector and monitor its performance. The detector is composed of a finely segmented scintillator-based inner tracking region surrounded by electromagnetic and hadronic sampling calorimetry. The upstream portion of the detector includes planes of graphite, iron and lead interleaved between tracking planes to facilitate the study of nuclear effects in neutrino interactions. Observations concerning the detector response over sustained periods of running are reported. The detector design and methods of operation have relevance to future neutrino experiments in which segmented scintillator tracking is utilized.

Measurement of the in-pile core temperature of an EL-4 pencil element, first charge (can of type-347 stainless steel, 0.4 mm thick, UO{sub 2} fuel, 11 mm diameter). Determination of the apparent thermal conductivity integral of in-pile UO{sub 2}; Mesure de la temperature a coeur en pile d'un crayon EL-4 1er jeu (gaine acier inoxydable, nuance 347 - epaisseur 0,4 mm - combustible UO{sub 2} - diametre 11 mm). Determination de l'integrale de conductibilite thermique apparente de l'UO{sub 2} en pile

Energy Technology Data Exchange (ETDEWEB)

Lavaud, B; Ringot, C; Vignesoult, N [Commissariat a l' Energie Atomique, Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France)

1966-11-01

The core temperature of a pencil fuel element depends on the thermal conductivity of the UO{sub 2}, and on the UO{sub 2}-can contact. This temperature may be known accurately only if in-pile tests using the actual geometry are carried out. The test described concerns the measurement of the core- temperature of an EL-4 fuel element, first charge, having a stainless steel can. This temperature is measured at the center of the in-pile pencil element using a high-temperature thermocouple (W-Re with Ta sheath). The element is subjected to operating conditions similar to those of EL-4, both for the specific power and the can temperature and for the pressure acting on the can. The specific power is obtained in the EL-3 reactor using a slightly higher enrichment for the UO{sub 2} than that planned for EL-4. The required can temperature and pressure are obtained using a Zircaloy-2 irradiation container filled with NaK, adapted for use in the EL-3 reactor. The core temperatures of the UO{sub 2}, and that of the can surface are measured. The power is calculated from the heat exchanges in the container calibrated in the laboratory. The temperature drop at the UO{sub 2}-can interface is deduced from laboratory measurements carried out under comparable heat flux conditions, and in a gas atmosphere corresponding to the beginning of the life-time of the fuel element. It is possible to draw an integral conductivity curve. It is also possible to check the temperature distribution in the oxide, as deduced from the thermal conductivity integral, by micro-graphic examination of the oxide structure. (authors) [French] La temperature a coeur d'un crayon combustible est fonction de la conductibilite thermique de l'UO{sub 2}, mais aussi du contact UO{sub 2}-gaine. Les essais de mesure en geometrie reelle en pile sont les seuls qui permettent d'avoir une connaissance exacte de cette valeur. L'essai dont il est question dans ce rapport a trait a la mesure de la temperature a coeur d

Measurement of β/Λ ratio in IEA-R1 reactor using noise technique

International Nuclear Information System (INIS)

Moreira, J.M.L.; Kassar, E.

1986-01-01

The ratio β/Λ for the IEA-R1 reactor is obtained experimentally through the noise analysis technique. This technique is based on the determination of the power spectral density of the reactor neutron population, with the reactor in a subcritical state driven by a 'white' neutron source. A ratio β/Λ of 43,5 s -1 is estimated from the break frequency of the measured transfer function of the IEA-R1 reactor. (Author) [pt

Optimization of Additive-Powder Characteristics for Metallic Micro-Cell UO{sub 2} Fuel Pellet Fabrication

Energy Technology Data Exchange (ETDEWEB)

Kim, Dong-Joo; Kim, Keon Sik; Rhee, Young Woo; Kim, Jong Hun; Oh, Jang Soo; Yang, Jae Ho; Koo, Yang-Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-05-15

The improvement in the thermal conductivity of the UO{sub 2} fuel pellet can enhance the fuel performance in various aspects. The mobility of the fission gases is reduced by the lower temperature gradient in the UO{sub 2} fuel pellet. That is to say, the capability of the fission gas retention of the fuel pellet can increase. In addition, the lower centerline temperature of the fuel pellet affects the accident tolerance for nuclear fuel as well as the enhancement of fuel safety and fuel pellet integrity under normal operation conditions. The nuclear reactor power can be uprated owing to the higher safety margin. Thus, many researches on enhancing the thermal conductivity of a nuclear fuel pellet for LWRs have been performed. Typically, an enhancement of the thermal conductivity of the UO{sub 2} fuel pellet can be obtained by the addition of a higher thermal conductive material in the fuel pellet. To maximize the effect of the thermal conductivity enhancement, a continuous and uniform channel of the thermal conductive material in the UO{sub 2} matrix must be formed. To enhance the thermal conductivity of a UO{sub 2} fuel pellet, the development of fabrication process of a Cr metallic micro-cell UO{sub 2} pellet with a continuous and uniform channel of the Cr metallic phase was carried out. The formation of the Cr-oxide phases was prevented and the uniformity of the Cr-metal phase distribution was enhanced simultaneously, through the optimization of the additive-powder characteristics. In the results, the Cr metallic micro-cell pellet with continuous and uniform Cr metallic channel could be obtained.

Determination of U{sub 3}O{sub 8} in UO{sub 2} by infrared spectroscopy

Energy Technology Data Exchange (ETDEWEB)

Silva, Liliane Aparecida; Lameiras, Fernando Soares; Santos, Ana Maria Matildes dos; Ferraz, Wilmar Barbosa; Barbosa, Joao Batista Santos, E-mail: lasfisica@gmail.com, E-mail: sl@cdtn.br, E-mail: amms@cdtn.br, E-mail: ferrazw@cdtn.br, E-mail: jbsb@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN), Belo Horizonte, MG (Brazil)

2017-01-15

The oxygen-uranium (O-U) system has various oxides, such as UO{sub 2}, U{sub 4}O{sub 9}, U{sub 3}O{sub 8}, and UO{sub 3}. Uranium dioxide is the most important one because it is used as nuclear fuel in nuclear power plants. UO{sub 2} can have a wide stoichiometric variation due to excess or deficiency of oxygen in its crystal lattice, which can cause significant modifications of its proprieties. O/U relation determination by gravimetry cannot differentiate a stoichiometric deviation from contents of other uranium oxides in UO{sub 2}. The presence of other oxides in the manufacturing of UO{sub 2} powder or sintered pellets is a critical factor. Fourier Transform Infrared Spectroscopy (FTIR) was used to identify U{sub 3}O{sub 8} in samples of UO{sub 2} powder. UO{sub 2} can be identified by bands at 340 cm{sup -1} and 470 cm{sup -1}, and U{sub 3}O{sub 8} and UO{sub 3} by bands at 735 cm{sup -1}, 910 cm{sup -1}, respectively. The methodology for sample preparation for FTIR spectra acquisition is presented, as well as the calibration for quantitative measurement of U{sub 3}O{sub 8} in UO{sub 2}. The content of U{sub 3}O{sub 8} in partially calcined samples of UO{sub 2} powder was measured by FTIR with good agreement with X-rays diffractometry (XRD). (author)

Recent Results From MINERvA

Energy Technology Data Exchange (ETDEWEB)

Patrick, Cheryl [Northwestern U.

2015-05-13

The MINERvA detector is situated in Fermilab's NuMI beam, which provides neutrinos and antineutrinos in the 1-20 GeV range. It is designed to make precision cross-section measurements for scattering processes on various nuclei. These proceedings summarize the differential cross-section distributions measured for several different processes. Comparison of these with various models hints at additional nuclear effects not included in common simulations. These results will help constrain generators' nuclear models and reduce systematic uncertainties on their predictions. An accurate cross-section model, with minimal uncertainties, is vital to oscillation experiments.

Contribution to the thermodynamic study of the non-stoichiometric oxides UO{sub 2+x} et FeO{sub 1+x}; Contribution a l'etude thermodynamique des oxydes non stoechiometriques UO{sub 2+x} et FeO{sub 1+x}

Energy Technology Data Exchange (ETDEWEB)

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

1964-04-15

This thermodynamic study has provided new results concerning the oxide UO{sub 2+X} and FeO{sub 1+x}. For the oxides UO{sub 2+X} correct values have been obtained for {mu}{sub O{sub 2}}{sup M} at 900, 1000 and 1100 deg. C using an improved method based on physico-chemical equilibria. For the oxides FeO{sub 1+x} the use of an E. Calvet high temperature calorimeter has made it possible to measure for the first time the values of h{sub O{sub 2}}{sup M} at 800 deg. C over the whole iron monoxide range. The method of oxygen transfer between oxides, usually used to determine the phase limits, has been improved by using a thermo-balance; this has made it possible to draw up simple rules which have to be respected in order to detect the phenomena under study. The theory due to J.S. Anderson has been applied to the oxides UO{sub 2+X} and a new method is given for improving the representation of non-stoichiometric oxides by models. (author) [French] Cette etude thermodynamique presente des resultats nouveaux en ce qui concerne les oxydes UO{sub 2+X} et FeO{sub 1+x}. Pour les oxydes UO{sub 2+X} les valeurs correctes de {mu}{sub O{sub 2}}{sup M} a 900, 1000 et 1100 deg. C ont pu etre obtenues, grace a la methode des equilibres physico-chimiques qui a ete amelioree. Pour les oxydes FeO{sub 1+x} l'emploi du microcalorimetre a haute temperature de Ed. CALVET a permis de mesurer pour la premiere fois les valeurs de h{sub O{sub 2}}{sup M} a 800 deg. C dans toute l'etendue du domaine du protoxyde de fer. La metode de transfert d'oxygene entre oxydes, habituellement utilisee pour determiner les limites de phase a ete perfectionnee par l'emploi d'une thermo-balance ce qui a permis d'enoncer les regles simples auxquelles il est indispensable de se conformer pour obtenir les limites cherchees. La theorie de J.S. Anderson a ete appliquee aux oxydes UO{sub 2+X} et une nouvelle voie est indiquee qui peut permettre de perfectionner la representation des oxydes non-stoechiometriques par des

Experimental characterization and modelling of UO2 mechanical behaviour at high temperatures and high strain rates

International Nuclear Information System (INIS)

Salvo, Maxime

2014-01-01

The aim of this work is to characterize and model the mechanical behavior of uranium dioxide (UO 2 ) during a Reactivity Initiated Accident (RIA). The fuel loading during a RIA is characterized by high strain rates (up to 1/s) and high temperatures (1000 C - 2500 C). Two types of UO 2 pellets (commercial and high density) were therefore tested in compression with prescribed displacement rates (0.1 to 100 mm/min corresponding to strain rates of 10 -4 - 10 -1 /s) and temperatures (1100 C - 1350 C - 1550 C et 1700 C). Experimental results (geometry, yield stress and microstructure) allowed us to define a hyperbolic sine creep law and a Drucker-Prager criterion with associated plasticity, in order to model grain boundaries fragmentation at the macroscopic scale. Finite Element Simulations of these tests and of more than 200 creep tests were used to assess the model response to a wide range of temperatures (1100 C - 1700 C) and strain rates (10 -9 /s - 10 -1 /s). Finally, a constitutive law called L3F was developed for UO 2 by adding to the previous model irradiation creep and tensile macroscopic cracking. The L3F law was then introduced in the 1.5D scheme of the fuel performance code ALCYONE-RIA to simulate the REP-Na tests performed in the experimental reactor CABRI. Simulation results are in good agreement with post tests examinations. (author) [fr

Shadow corrosion evaluation in the Studsvik R2 reactor

International Nuclear Information System (INIS)

Sanders, Ch.; Lysell, G.

2000-01-01

Post-irradiation examination has shown that increased corrosion occurs when zirconium alloys are in contact with or in proximity to other metallic objects. The observations indicate an influence of irradiation from the adjacent component as the enhanced corrosion occurs as a 'shadow' of the metallic object on the zirconium surface. This phenomenon could ultimately limit the lifetime of certain zirconium alloy components in the reactor. The Studsvik R2 materials test reactor has an In-Core Autoclave (INCA) test facility especially designed for water chemistry and materials research. The INCA facility has been evaluated and found suitable for shadow corrosion studies. The R2 reactor core containing the INCA facility was modeled with the Monte Carlo N-Particle (MCNP) code in order to evaluate the electron deposition in various materials and to develop a hypothesis of the shadow corrosion mechanism. (authors)

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

International Nuclear Information System (INIS)

Lucchini, J.F.

2001-01-01

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

Structure changes of irradiated UO2

International Nuclear Information System (INIS)

Komatsu, Junji; Yokouchi, Yoji; Kajiyama, Takashi; Terunuma, Toshihiro; Koizumi, Masumichi

1973-01-01

The structural change of UO 2 irradiated in GETR reactor was analyzed on void distribution, fuel cracking, and gap conductance between fuel and cladding. Metallographic analysis was carried out on 21 sections of irradiated fuel pins. Radial void distribution was measured by the linear analysis technique based on the equivalence between the volume fraction of voids and the intercepted length of lines between void boundaries. Fuel cracks were classified into two types, namely radial cracks and circumferential cracks. The radial position, length, angle and number of each fuel clad were measured on metallographic section and autoradiography. The gap conductance between fuel and cladding was calculated from the equation h = q/(T sub(s) - T sub(i)) where h is gap conductance, T sub(i) is inside clad temperature and T sub(s) is outside clad temperature. In void distribution, as the result of studying the effect of linear heat rating on the radial void fraction of UO 2 fuel irradiated with the similar level of burnup, one specimen showed that the void fraction of columnar grain growth region was comparable to that of fabricated region, and two specimens showed higher void fraction at fabricated region than the calculated one. In fuel cladding, no significant effect of burnup on fuel cracking was observed, and the number of fuel cracking increased with shutdown or scram numbers, but the radial position of circumferential cracks was not much changed. In gap conductance, it was influenced by the estimation of temperature of columnar grain growth. (Iwakiri, K.)

The preparation of UO2 ceramic microspheres with an advanced process (TGU)

International Nuclear Information System (INIS)

Xu Zhichang; Tang Yaping; Zhang Fuhong

1994-04-01

The UO 2 ceramic microspheres are the most important materials in the spherical fuel elements for high temperature reactor (HTR). An advanced process for preparation of UO 2 ceramic microspheres has been developed at Institute of Nuclear Energy Technology, Tsinghua University. This process known as total gelation process of uranium (TGU), is based on the traditional sol-gel process, external gelation process and internal gelation process of uranium (EGU and IGU), and has been selected as the production process. The result of batch test is described. Accordance with the requirements of quality control (QC) and quality assurance (QA), the stabilization of operating parameters and product quality is tested., The results on batch test have shown that as well as all of the operated parameters are fixed, then the product quality can be stable as well as the product specification can be met. When the colloidal flow rate and the vibration frequency of nozzle are fixed, the kernel's size is also fixed. When the sintering temperature and time are fixed, the product density is also fixed. When the hydrogen atmosphere is used, the O/U ratio is very near to stoichiometry. The performance and structure of UO 2 ceramic microspheres are also given

Reaction kinetics aspect of U3O8 kernel with gas H2 on the characteristics of activation energy, reaction rate constant and O/U ratio of UO2 kernel

International Nuclear Information System (INIS)

Damunir

2007-01-01

The reaction kinetics aspect of U 3 O 8 kernel with gas H 2 on the characteristics of activation energy, reaction rate constant and O/U ratio of UO 2 kernel had been studied. U 3 O 8 kernel was reacted with gas H 2 in a reduction furnace at varied reaction time and temperature. The reaction temperature was varied at 600, 700, 750 and 850 °C with a pressure of 50 mmHg for 3 hours in gas N 2 atmosphere. The reation time was varied at 1, 2, 3 and 4 hours at a temperature of 750 °C using similar conditions. The reaction product was UO 2 kernel. The reaction kinetic aspect between U 3 O 8 and gas H 2 comprised the minimum activation energy (ΔE), the reaction rate constant and the O/U ratio of UO 2 kernel. The minimum activation energy was determined from a straight line slope of equation ln [{D b . R o {(1 - (1 - X b ) ⅓ } / (b.t.Cg)] = -3.9406 x 10 3 / T + 4.044. By multiplying with the straight line slope -3.9406 x 10 3 , the ideal gas constant (R) 1.985 cal/mol and the molarity difference of reaction coefficient 2, a minimum activation energy of 15.644 kcal/mol was obtained. The reaction rate constant was determined from first-order chemical reaction control and Arrhenius equation. The O/U ratio of UO 2 kernel was obtained using gravimetric method. The analysis result of reaction rate constant with chemical reaction control equation yielded reaction rate constants of 0.745 - 1.671 s -1 and the Arrhenius equation at temperatures of 650 - 850 °C yielded reaction rate constants of 0.637 - 2.914 s -1 . The O/U ratios of UO 2 kernel at the respective reaction rate constants were 2.013 - 2.014 and the O/U ratios at reaction time 1 - 4 hours were 2.04 - 2.011. The experiment results indicated that the minimum activation energy influenced the rate constant of first-order reaction and the O/U ratio of UO 2 kernel. The optimum condition was obtained at reaction rate constant of 1.43 s -1 , O/U ratio of UO 2 kernel of 2.01 at temperature of 750 °C and reaction time of 3

A new technique to measure fission-product diffusion coefficients in UO2 fuel

International Nuclear Information System (INIS)

Hocking, W.H.; Verrall, R.A.; Bushby, S.J.

1999-01-01

This paper describes a new out-reactor technique for the measurement of fission-product diffusion rates in UO 2 . The technique accurately simulates in-reactor fission-fragment effects: a thermal diffusion that is due to localized mixing in the fission track, radiation-enhanced diffusion that is due to point-defect creation by fission fragments, and bubble resolution. The technique utilizes heavy-ion accelerators - low energy (40 keV to 1 MeV) for fission-product implantation, high energy (72 MeV) to create fission-fragment damage effects, and secondary ion mass spectrometry (SIMS) for measuring the depth profile of the implanted species. Preliminary results are presented from annealing tests (not in the 72 MeV ion flux) at 1465 deg. C and 1650 deg. C at low and high concentrations of fission products. (author)

Study of physical properties of UO2 quality improvement result

International Nuclear Information System (INIS)

Rachmat-Pratomo; Hidayati; Didiek Herhady, R; Busron-Masduki

1996-01-01

Activation of uranium dioxide (UO 2 ) by reoxidation to U 3 O 8 and reduction to uranium dioxide (UO 2 ) by temperature reduction variation of 850 o C and 900 o C for 3 hours has been studied. The physical properties before and after treatment are compared. It proved that the oxidation-reduction cycle increased the physical properties. It can be concluded that the reoxidation of UO 2 to U 3 O 8 on fourth cycle and reduction at 900 o C for 3 hours result in a density of 1.32 gram/ml a tap density of 1.60 gram/ml, true density of 9.08 gram/ml and O/U ratio : 2.04. Reduction at 850 o C, for 3 hours result in the bulk density of 1.30 gram/ml, tap density of 1.58 gram/ml, true density of 9.04 gram/ml and O/U ratio 2.09

Contribution to the study of UO2 pellet fabrication

International Nuclear Information System (INIS)

Fogaca Filho, N.; Gentile, E.F.; Mourao, M.B.; Souza Santos, T.D. de; Haydt, H.M.

1977-01-01

The establishment of a set of parametric comparisons related to UO 2 powders of two different origins as the ammonium diuranate and the ammonium uranyl carbonate is presented. It is emphasized the importance due to the pressing capability of the powders and the requirement for homogeneous microstructure for both, the pore distribution and the grain size. In order to establish the parameters of comparison, all the required normal tests for the in-process control of fabrication of fuel elements for nuclear power reactors were performed, particularly to the re-sintering test, in view of the evaluation of dimensional stability of the pellets [pt

Dissolution kinetics of UO2: Flow-through tests on UO2.00 pellets and polycrystalline schoepite samples in oxygenated, carbonate/bicarbonate buffer solutions at 25 degree C

International Nuclear Information System (INIS)

Nguyen, S.N.; Weed, H.C.; Leider, H.R.; Stout, R.B.

1991-10-01

The modelling of radionuclide release from waste forms is an important part of the performance assessment of a potential, high-level radioactive waste repository. Since spent fuel consists of UO 2 containing actinide elements and other fission products, it is necessary to determine the principal parameters affecting UO 2 dissolution and quantify their effects on the dissolution rate before any prediction of long term release rates of radionuclides from the spent fuel can be made. As part of a complex matrix to determine the dissolution kinetics of UO 2 as a function of time, pH, carbonate/bicarbonate concentration and oxygen activity, we have measured the dissolution rates at 25 degrees C of: (1) UO 2 pellets; (2) UO 2.00 powder and (3) synthetic dehydrated schoepite, UO 3 .H 2 O using a single-pass flow through system in an argon-atmosphere glove box. Carbonate, carbonate/bicarbonate, and bicarbonate buffers with concentrations ranging from 0.0002 M to 0.02 M and pH values form 8 to 11 have been used. Argon gas mixtures containing oxygen (from 0.002 to 0.2 atm) and carbon dioxide (from 0 to 0.011 atm) were bubbled through the buffers to stabilize their pH values. 12 refs., 2 tabs

Oxidation of UO2 at 150 to 3500C

International Nuclear Information System (INIS)

White, G.D.; Knox, C.A.; Gilbert, E.R.; Johnson, A.B. Jr.

1983-07-01

Oxidation of UO 2 through breached LWR spent fuel rods during interim storage in air atmospheres is a potential mechanism for degradation of cladding integrity. The temperature-time range of published data are inadequate to establish long term behavior under dry storage conditions. Consequently, tests are being conducted in the temperature range of 150 to 350 0 C on unirradiated pellets to evaluate fuel oxidation behavior. The tests have revealed significant-to-minor oxidation at temperatures down to 200 0 C and no measurable oxidation at 150 0 C for times up to 3000 hours. Oxidation at 200 0 C for 2000 hours led to formation of low density particulate U 3 O 8 which destroys pellet integrity. Oxidation of UO 2 pellets at 215 and 250 0 C was signifcantly accelerated by the presence of 1 volume percent NO 2 in the air. NO 2 is a potential constituent of the air, forming by radiolysis in the gamma radiation field associated with spent fuel assemblies. NO 2 reaction with UO 2 pellets leads to accelerated formation of UO 3 and pellet disintegration. 11 references, 15 figures

Assessment of structures and stabilities of defect clusters and surface energies predicted by nine interatomic potentials for UO{sub 2}

Energy Technology Data Exchange (ETDEWEB)

Taller, Stephen A. [School of Nuclear Engineering, Purdue University, West Lafayette, IN 47907 (United States); Bai, Xian-Ming, E-mail: xianming.bai@inl.gov [Fuels Modeling and Simulation Department, Idaho National Laboratory, Idaho Falls, ID 83415 (United States)

2013-11-15

The irradiation in nuclear reactors creates many point defects and defect clusters in uranium dioxide (UO{sub 2}) and their evolution severely degrades the thermal and mechanical properties of the nuclear fuels. Previously many empirical interatomic potentials have been developed for modeling defect production and evolution in UO{sub 2}. However, the properties of defect clusters and extended defects are usually not fitted into these potentials. In this work nine interatomic potentials for UO{sub 2} are examined by using molecular statics and molecular dynamics to assess their applicability in predicting the properties of various types of defect clusters in UO{sub 2}. The binding energies and structures for these defect clusters have been evaluated for each potential. In addition, the surface energies of voids of different radii and (1 1 0) flat surfaces predicted by these potentials are also evaluated. It is found that both good agreement and significant discrepancies exist for these potentials in predicting these properties. For oxygen interstitial clusters, these potentials predict significantly different defect cluster structures and stabilities; For defect clusters consisting of both uranium and oxygen defects, the prediction is in better agreement; The surface energies predicted by these potentials have significant discrepancies, and some of them are much higher than the experimentally measured values. The results from this work can provide insight on interpreting the outcome of atomistic modeling of defect production using these potentials and may provide guidelines for choosing appropriate potential models to study problems of interest in UO{sub 2}.

Irradiation experience of IPEN fuel at IEA-R1 research reactor

International Nuclear Information System (INIS)

Perrotta, Jose A.; Neto, Adolfo; Durazzo, Michelangelo; Souza, Jose A.B. de; Frajndlich, Roberto

1998-01-01

IPEN/CNEN-SP produces, for its IEA-R1 Research Reactor, MTR fuel assemblies based on U 3 O 8 -Al dispersion fuel type. Since 1985 a qualification program on these fuel assemblies has been performed. Average 235 U burnup of 30% and peak burnup of 50% was already achieved by these fuel assemblies. This paper presents some results acquire, by these fuel assemblies, under irradiation at IEA-R1 Research Reactor. (author)

Measurement of partonic nuclear effects in deep-inelastic neutrino scattering using MINERvA

Science.gov (United States)

Mousseau, J.; Wospakrik, M.; Aliaga, L.; Altinok, O.; Bellantoni, L.; Bercellie, A.; Betancourt, M.; Bodek, A.; Bravar, A.; Budd, H.; Cai, T.; Carneiro, M. F.; Christy, M. E.; Chvojka, J.; da Motta, H.; Devan, J.; Dytman, S. A.; Díaz, G. A.; Eberly, B.; Felix, J.; Fields, L.; Fine, R.; Gago, A. M.; Galindo, R.; Gallagher, H.; Ghosh, A.; Golan, T.; Gran, R.; Harris, D. A.; Higuera, A.; Hurtado, K.; Kiveni, M.; Kleykamp, J.; Kordosky, M.; Le, T.; Maher, E.; Manly, S.; Mann, W. A.; Marshall, C. M.; Martinez Caicedo, D. A.; McFarland, K. S.; McGivern, C. L.; McGowan, A. M.; Messerly, B.; Miller, J.; Mislivec, A.; Morfín, J. G.; Naples, D.; Nelson, J. K.; Norrick, A.; Nuruzzaman; Osta, J.; Paolone, V.; Park, J.; Patrick, C. E.; Perdue, G. N.; Rakotondravohitra, L.; Ramirez, M. A.; Ransome, R. D.; Ray, H.; Ren, L.; Rimal, D.; Rodrigues, P. A.; Ruterbories, D.; Schellman, H.; Schmitz, D. W.; Solano Salinas, C. J.; Tagg, N.; Tice, B. G.; Valencia, E.; Walton, T.; Wolcott, J.; Zavala, G.; Zhang, D.; Minerν A Collaboration

2016-04-01

The MINERvA Collaboration reports a novel study of neutrino-nucleus charged-current deep inelastic scattering (DIS) using the same neutrino beam incident on targets of polystyrene, graphite, iron, and lead. Results are presented as ratios of C, Fe, and Pb to CH. The ratios of total DIS cross sections as a function of neutrino energy and flux-integrated differential cross sections as a function of the Bjorken scaling variable x are presented in the neutrino-energy range of 5-50 GeV. Based on the predictions of charged-lepton scattering ratios, good agreement is found between the data and prediction at medium x and low neutrino energy. However, the ratios appear to be below predictions in the vicinity of the nuclear shadowing region, x <0.1 . This apparent deficit, reflected in the DIS cross-section ratio at high Eν, is consistent with previous MINERvA observations [B. Tice et al. (MINERvA Collaboration), Phys. Rev. Lett. 112, 231801 (2014).] and with the predicted onset of nuclear shadowing with the axial-vector current in neutrino scattering.

Irradiation of defected SAP clad UO2 fuel in the X-7 organic loop

International Nuclear Information System (INIS)

Robertson, R.F.S.; Cracknell, A.G.; MacDonald, R.D.

1961-10-01

This report describes an experiment designed to test the behaviour under irradiation of a UO 2 fuel specimen clad in a defected SAP sheath and cooled by recirculating organic liquid. The specimen containing the defect was irradiated in the X-7 loop in the NRX reactor from the 25th of November until the 13th of December 1960. Up to the 13th of December the behaviour was analogous to that seen with defected UO 2 specimens clad in zircaloy which were irradiated in water loops. Reactor power transients resulted in peaking of gamma ray activities in the loop, but on steady operation these activities tended to fall to a steady state level, Over this period the pressure drop across the fuel increased by a factor of two, the increases occurring after reactor shut downs and start ups. On 13th December the pressure drop increased rapidly, after a reactor shut down and start up, to over five times its original value and the activities in the loop rose to a high level. The specimen was removed and examination showed that the sheath was very badly split and that the volume between the fuel and the sheath was filled with a hard black organic substance. This report gives full details of the irradiation and of the post -irradiation examination. Correlation of the observed phenomenon is attempted and a preliminary assessment of the problems which would be associated with defect fuel in an organic reactor is given. (author)

Correlations between different methods of UO2 pellet density measurement

International Nuclear Information System (INIS)

Yanagisawa, Kazuaki

1977-07-01

Density of UO 2 pellets was measured by three different methods, i.e., geometrical, water-immersed and meta-xylene immersed and treated statistically, to find out the correlations between UO 2 pellets are of six kinds but with same specifications. The correlations are linear 1 : 1 for pellets of 95% theoretical densities and above, but such do not exist below the level and variated statistically due to interaction between open and close pores. (auth.)

Low Temperature Two-Steps Sintering (LTTSS) - an innovative method for consolidating porous UO2 pellets

International Nuclear Information System (INIS)

Sanjay Kumar, D.; Ananthasivan, K.; Senapati, Abhiram; Venkata Krishnan, R.

2015-01-01

Metallic uranium and its alloys are an important fuel for fast reactors. Presently, metallic uranium is being prepared using expensive fluoro-metallothermic process. Recent reports suggest that metal oxide could be reduced to metal using a novel electrochemical de-oxidation method and this could serve as attractive alternate for expensive metallothermic process. In view of which, a research program is being pursued in our Centre to develop an optimum process parameter for the scaled up preparation of metallic uranium efficiently. One of the important process parameter is the size, nature and distribution of porosity in the urania pellet. Essentially the ceramic form of the urania should encompass interconnected porosity that would allow percolation of melts into the UO 2 . However, the matrix density of the pellet should be high to ensure that it possesses good handling strength and is electrically conducting. Hence preparation of high dense porous UO 2 pellets was required. In this study, we report the preparation of porous UO 2 pellets possessing a very high matrix density by using the citrate gel-combustion method. The 'as-prepared' powders were consolidated at various compaction pressures as such and these pellets were sintered in 8 mol %Ar+H 2 gas with a flow rate of 250 mL/min at 1073 K for 30 min followed by soaking at 1473 K for 4 h with heating rate of 5 K min -1 in a molybdenum furnace. X-ray diffraction studies revealed that these pellets contained UO 2 . The morphological analysis sintered pellets was carried out by using Scanning Electron Microscope (M/s. Philips model XL 30, Netherlands). All these pellets were gold coated

Thermal and Mechanical Properties of UO2 and PuO2

International Nuclear Information System (INIS)

Kato, M.; Matsumoto, T.

2015-01-01

It is important to evaluate basic properties of UO 2 and PuO 2 as fundamental aspects of MA-bearing MOX fuel development. In this work, mechanical properties of UO 2 and PuO 2 were investigated by an ultrasound pulse-echo method. Longitudinal and transversal wave velocities were measured in UO 2 and PuO 2 pellets, and Young's modulus and shear modulus were evaluated, which were 219 MPa and 89 MPa for PuO 2 , and 249 MPa and 95 MPa for UO 2 , respectively. Poisson's ratio was 0.32 in both materials. The relationship between mechanical and thermal properties was described by using thermal expansion data which had been reported previously, and the heat capacity and thermal conductivity were analysed. (authors)

Reinvestigation of the crystal structure of kasolite, Pb[(UO{sub 2})(SiO{sub 4})](H{sub 2}O), an important alteration product of uraninite, UO{sub 2+x}

Energy Technology Data Exchange (ETDEWEB)

Fejfarová, Karla; Dušek, Michal [Institute of Physics ASCR, v.v.i., Na Slovance 2, 18221 Praha (Czech Republic); Plášil, Jakub, E-mail: jakub_plasil@nm.cz [Department of Mineralogy and Petrology, National Museum, Václavské nám. 68, Prague 1, 115 79-CZ (Czech Republic); Institute of Geological Science, Faculty of Science, Masaryk University, Kotlářská 2, CZ-611 37, Brno (Czech Republic); Čejka, Jiří; Sejkora, Jiří [Department of Mineralogy and Petrology, National Museum, Václavské nám. 68, Prague 1, 115 79-CZ (Czech Republic); Škoda, Radek [Institute of Geological Science, Faculty of Science, Masaryk University, Kotlářská 2, CZ-611 37, Brno (Czech Republic)

2013-03-15

The crystal structure of kasolite, Pb[(UO{sub 2})(SiO{sub 4})](H{sub 2}O), Z = 4, monoclinic, with a = 6.7050(3), b = 6.9257(2), c = 13.2857(5) Å, β = 105.064(4)°, V = 595.74(3) Å{sup 3}, the space group P2{sub 1}/c, has been solved by charge-flipping method and refined by the full-matrix least-squares techniques to an agreement factor (R{sub obs}) of 2.2% and, a goodness-of-fit (GOF) of 1.26 using 1243 unique observed diffraction maxima (I{sub obs} > 3σ(I)) collected with MoKα X-radiation and a 4 K CCD area detector. The crystal structure of kasolite contains 1 unique U{sup 6+} position that is part of a nearly linear uranyl ion (UO{sub 2}){sup 2+}, coordinated in the equatorial plane by five O ligands, forming pentagonal bipyramid. The uranyl pentagonal bipyramids share edges to form chains parallel to [0 1 0]. The additional edge of uranyl polyhedra is shared by silicate tetrahedra to form sheets parallel to (1 0 0). There is one unique position of Pb{sup 2+} in the interlayer. O ligands and 1 (H{sub 2}O) non-transformer group coordinate Pb{sup 2+} exhibiting [2 + 6] coordination. A network of H-bonds provides an additional linkage of an interlayer to the sheets besides Pb–O bonds. Chemical composition of the studied crystals, obtained by the electron microprobe, is reported and is in agreement with the crystal structure refinement.

Tellurites of hexavalent uranium: first observation of polymerized (UO{sub 4}){sup 2-} tetraoxido cores

Energy Technology Data Exchange (ETDEWEB)

Zadoya, Anastasiya I.; Siidra, Oleg I.; Nazarchuk, Evgeny V.; Bocharov, Sergey N. [Department of Crystallography, St. Petersburg State University (Russian Federation); Bubnova, Rimma S. [Department of Crystallography, St. Petersburg State University (Russian Federation); Institute of Silicate Chemistry, Russian Academy of Sciences, St. Petersburg (Russian Federation)

2016-09-15

Two novel Ca{sub 2}(UO{sub 3})(TeO{sub 3}){sub 2} (1) and K{sub 2}(UO{sub 2}){sub 2}O{sub 2}(TeO{sub 3}) (2) uranyl tellurites were obtained from telluric acid, used as a starting reagent for both compounds. In 1, the tetraoxido core is coordinated by TeO{sub 3} groups and UO{sub 4} squares polymerize into [UO{sub 3}] chains. The tetraoxido core coordination modes in compound 1 are unique. New layered {sub ∞}{sup 2}[(UO{sub 2}){sub 2}(TeO{sub 3})O{sub 2}]{sup 2-} topology is observed for 2. Both of the compounds were studied by the means of high-temperature X-ray diffraction. The thermal decomposition of 1 and 2 is different and leads to formation of uranate compounds. (copyright 2016 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

Fission gas release from UO2 pellet fuel at high burn-up

International Nuclear Information System (INIS)

Vitanza, C.; Kolstad, E.; Graziani, U.

1979-01-01

Analysis of in-reactor measurements of fuel center temperature and rod internal pressure at the OECD Halden Reactor Project has led to the development of an empirical fission gas release model, which is described. The model originally derived from data obtained in the low and intermediate burn-up range, appears to give good predictions for rods irradiated to high exposures as well. PIE puncturing data from seven fuel rods, operated at relatively constant powers and peak center temperatures between 1900 and 2000 0 C up to approx. 40,000 MWd/t UO 2 , did not exhibit any burn-up enhancement on the fission gas release rate

Grain growth behavior of Cr dispersed UO{sub 2} pellets according to change of oxygen potential during the isothermal sintering

Energy Technology Data Exchange (ETDEWEB)

Oh, Jang Soo; Yang, Jae Ho; Kim, Dong Joo; Kim, Jong Hun; Nam, Ik Hui; Rhee, Young Woo; Kim, Keon Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2012-10-15

Recent development of advanced UO{sub 2} pellet materials for commercial reactors is mainly focused on the large grain pellet which can deform easily at an elevated temperature. Cr{sub 2}O{sub 3}-doped UO{sub 2} pellet is one of the promising candidates. To increase the grain size effectively, it is important to control the additive content and sintering atmosphere. Relevant research on the Cr{sub 2}O{sub 3} doped UO{sub 2} system revealed that the doped Cr{sub 2}O{sub 3} formed a liquid phase under optimized oxygen potential, and those liquid phases promoted the grain growth. Recent work also showed that step-wise variation of sintering atmosphere during the isothermal annealing step significantly increased the grain size of Cr{sub 2}O{sub 3} doped UO{sub 2} pellet. In this paper, we investigated effect of oxygen potential change at the beginning of isothermal sintering stage on the grain growth in metallic Cr dispersed UO{sub 2} pellets. The study on the milling effect of powder mixture on the grain growth is also a part of this work.

Simulation of channel blockage for the IEA-R1 research reactor using RELAP/MOD 3

International Nuclear Information System (INIS)

Oliveira, Eduardo C.F. de; Castrillo, Lazara Silveira

2015-01-01

Research reactors have great importance in the area of nuclear technology, such as radioisotope production, research in nuclear physics, development of new technologies and staff training for reactor operation. The IEA-R1 is a Brazilian research reactor type pool, located at the IPEN (Instituto de Pesquisas Energeticas e Nucleares). In this work is simulated with computer code RELAP5 / MOD 3.3.2 gamma, the effect caused by partial and complete blockage of a channel in MTR fuel element of the IEA-R1 core, in order to analyzed the thermal hydraulic parameters on adjacent channels. (author)

Simulation of channel blockage for the IEA-R1 research reactor using RELAP/MOD 3

Energy Technology Data Exchange (ETDEWEB)

Oliveira, Eduardo C.F. de; Castrillo, Lazara Silveira, E-mail: ecfoliveira@hotmail.com, E-mail: lazara.castrillo@upe.br [Universidade de Pernambuco (UPE), Recife, PE (Brazil). Escola Politecnica de Pernambuco

2015-07-01

Research reactors have great importance in the area of nuclear technology, such as radioisotope production, research in nuclear physics, development of new technologies and staff training for reactor operation. The IEA-R1 is a Brazilian research reactor type pool, located at the IPEN (Instituto de Pesquisas Energeticas e Nucleares). In this work is simulated with computer code RELAP5 / MOD 3.3.2 gamma, the effect caused by partial and complete blockage of a channel in MTR fuel element of the IEA-R1 core, in order to analyzed the thermal hydraulic parameters on adjacent channels. (author)

Physical characteristics of Gd2O3-UO2 fuel in LWR

International Nuclear Information System (INIS)

Matsuura, Shojiro; Kobayashi, Iwao; Furuta, Toshiro; Toba, Masao; Tsuda, Katsuhiro.

1981-12-01

A series of critical experiments in light water lattice were carried out on five kinds of Gadolinia-Uranium dioxide (Gd 2 O 3 -UO 2 ) test fuel rods containing 0.0, 0.05, 0.25, 1.50, 3.00 weight % of Gd 2 O 3 in Gd 2 O 3 -UO 2 . Reactivity effect, power distribution, neutron flux distribution, and temperature coefficient were measured for three types of lattices which were in shapes of annular, rectangular parallele-piped, and JPDR mockup core. The theoretical values corresponding to the measured ones were obtained by means of the design method for the FTA which is the test fuel assembly with Gd 2 O 3 -UO 2 rods for JPDR, and the accuracy was checked. In general, the calculated values were in good agreement with the measured ones. Besides, the following characteristics of Gd 2 O 3 -UO 2 rods are recognized both in measurement and calculation, i.e. (1) the effect due to gadolinia on reactivity, power distribution, and thermal neutron flux distribution are steeply saturating; the gadolinia content of only 1.50 weight % is enough to reach the almost saturated condition, (2) the relative power becomes 20% to that of normal fuel under the saturated condition, (3) the relation between the negative reactivity and the power depression effect due to gadolinia is almost linear, and (4) the interference on power depression between the adjacent gadolinia loaded rods is almost negligible, and that on reactivity effect is 15% at most. (author)

New digital control system for the operation of the Colombian research reactor IAN-R1; Nuevo sistema de control digital para la operacion del reactor de investigacion Colombiano IAN-R1

Energy Technology Data Exchange (ETDEWEB)

Celis del A, L.; Rivero, T.; Bucio, F.; Ramirez, R.; Segovia, A.; Palacios, J., E-mail: lina.celis@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2015-09-15

En 2011, Mexico won the Colombian international tender for the renewal of instrumentation and control of the IAN-R1 Reactor, to Argentina and the United States. This paper presents the design criteria and the development made for the new digital control system installed in the Colombian nuclear reactor IAN-R1, which is based on a redundant and diverse architecture, which provides increased availability, reliability and safety in the reactor operation. This control system and associated instrumentation met all national export requirements, with the safety requirements established by the IAEA as well as the requirements demanded by the Colombian Regulatory Body in nuclear matter. On August 20, 2012, the Colombian IAN-R1 reactor reached its first criticality controlled with the new system developed at Instituto Nacional de Investigaciones Nucleares (ININ). On September 14, 2012, the new control system of the Colombian IAN-R1 reactor was officially handed over to the Colombian authorities, this being the first time that Mexico exported nuclear technology through the ININ. Currently the reactor is operating successfully with the new control system, and has an operating license for 5 years. (Author)

Thermal neutron capture cross section of gadolinium by pile-oscillation measurements in MINERVE

International Nuclear Information System (INIS)

Leconte, P.; Di-Salvo, J.; Antony, M.; Pepino, A.; Hentati, A.

2012-01-01

Natural gadolinium is used as a burnable poison in most LWR to account for the excess of reactivity of fresh fuels. For an accurate prediction of the cycle length, its nuclear data and especially its neutron capture cross section needs to be known with a high precision. Recent microscopic measurements at Rensselaer Polytechnic Inst. (RPI) suggest a 11% smaller value for the thermal capture cross section of 157 Gd, compared with most of evaluated nuclear data libraries. To solve this inconsistency, we have analyzed several pile-oscillation experiments, performed in the MINERVE reactor. They consist in the measurement of the reactivity variation involved by the introduction in the reactor of small-samples, containing different mass amounts of natural gadolinium. The analysis of these experiments is done through the exact perturbation theory, using the PIMS calculation tool, in order to link the reactivity effect to the thermal capture cross section. The measurement of reactivity effects is used to deduce the 2200 m.s-1 capture cross section of nat Gd which is (49360 ± 790) b. This result is in good agreement with the JEFF3.1.1 value (48630 b), within 1.6% uncertainty at 1σ, but is strongly inconsistent with the microscopic measurements at RPI which give (44200 ± 500) b. (authors)

Thermal neutron capture cross section of gadolinium by pile-oscillation measurements in MINERVE

Energy Technology Data Exchange (ETDEWEB)

Leconte, P.; Di-Salvo, J.; Antony, M.; Pepino, A. [CEA, DEN, DER, Cadarache, F-13108 Saint-Paul-Lez-Durance (France); Hentati, A. [International School in Nuclear Engineering, Cadarache, F-13108 Saint-Paul-Lez-Durance (France)

2012-07-01

Natural gadolinium is used as a burnable poison in most LWR to account for the excess of reactivity of fresh fuels. For an accurate prediction of the cycle length, its nuclear data and especially its neutron capture cross section needs to be known with a high precision. Recent microscopic measurements at Rensselaer Polytechnic Inst. (RPI) suggest a 11% smaller value for the thermal capture cross section of {sup 157}Gd, compared with most of evaluated nuclear data libraries. To solve this inconsistency, we have analyzed several pile-oscillation experiments, performed in the MINERVE reactor. They consist in the measurement of the reactivity variation involved by the introduction in the reactor of small-samples, containing different mass amounts of natural gadolinium. The analysis of these experiments is done through the exact perturbation theory, using the PIMS calculation tool, in order to link the reactivity effect to the thermal capture cross section. The measurement of reactivity effects is used to deduce the 2200 m.s-1 capture cross section of {sup nat}Gd which is (49360 {+-} 790) b. This result is in good agreement with the JEFF3.1.1 value (48630 b), within 1.6% uncertainty at 1{sigma}, but is strongly inconsistent with the microscopic measurements at RPI which give (44200 {+-} 500) b. (authors)

Spent UO{sub 2} TRISO coated particles. Instant release fraction and microstructure evolution

Energy Technology Data Exchange (ETDEWEB)

Curtius, Hildegard; Kaiser, Gabriele; Lieck, Norman; Guengoer, Murat; Klinkenberg, Martina; Bosbach, Dirk [Research Center Juelich (Germany). Inst. of Energy and Climate Research IEK-6: Nuclear Waste Management and Reactor Safety

2015-09-01

The impact of burn-up on the instant release fraction (IRF) from spent fuel was studied using very high burn-up UO{sub 2} fuel (∝ 100 GWd/t) from a prototype high temperature reactor (HTR). TRISO (TRi-structural-ISO-tropic) particles from the spherical fuel elements contain UO{sub 2} fuel kernels (500 μm diameter) which are coated by three tight layers ensuring the encapsulation of fission products during reactor operation. After cracking of the tight coatings {sup 85}Kr and {sup 14}C as {sup 14}CO{sub 2} were detected in the gas fraction. Xe was not detected in the gas fraction, although ESEM (Environmental Scanning Electron Microscope) investigations revealed an accumulation in the buffer. UO{sub 2} fuel kernels were exposed to synthetic groundwater under oxic and anoxic/reducing conditions. U concentration in the leachate was below the detection limit, indicating an extremely low matrix dissolution. Within the leach period of 276 d {sup 90}Sr and {sup 134/137}Cs fractions located at grain boundaries were released and contribution to IRF up to max. 0.2% respectively 8%. Depending on the environmental conditions, different release functions were observed. Second relevant release steps occurred in air after ∝ 120 d, indicating the formation of new accessible leaching sites. ESEM investigations were performed to study the impact of leaching on the microstructure. In oxic environment, numerous intragranular open pores acting as new accessible leaching sites were formed and white spherical spots containing Mo and Zr were identified. Under anoxic/reducing conditions numerous metallic precipitates (Mo, Tc and Ru) filling the intragranular pores and white spherical spots containing Mo and Zr, were detected. In conclusion, leaching in different geochemical environments influenced the speciation of radionuclides and in consequence the stability of neoformed phases, which has an impact on IRF.

Qualification of power determination and in-pile measurements in the UO{sub 2} Gd{sub 2} 0{sub 2} fuel irradiation test IFA 636

Energy Technology Data Exchange (ETDEWEB)

Tverberg, T.; Volkov, B.; Kim, J-C.

2004-04-15

showed no substantial PCMI in the Gd-doped fuel rods whereas for the UO{sub 2} rod some PCMI effects were seen to occur during reactor operation at different burnups. (Author)

The effect of fuel chemistry on UO{sub 2} dissolution

Energy Technology Data Exchange (ETDEWEB)

Casella, Amanda, E-mail: amanda.casella@pnnl.gov [Pacific Northwest National Laboratory, PO Box 999, MSIN P7-25, Richland, WA 99352 (United States); Hanson, Brady, E-mail: brady.hanson@pnnl.gov [Pacific Northwest National Laboratory, PO Box 999, MSIN P7-27, Richland, WA 99352 (United States); Miller, William [University of Missouri Research Reactor, 1513 Research Park Drive, Columbia, MO 65211 (United States)

2016-08-01

The dissolution rate of both unirradiated UO{sub 2} and used nuclear fuel has been studied by numerous countries as part of the performance assessment of proposed geologic repositories. In the scenario of waste package failure and groundwater contact with the fuel, the effects of variables such as temperature, dissolved oxygen, and water and fuel chemistry on the dissolution rates of the fuel are necessary to provide a quantitative estimate of the potential release over geologic time frames. The primary objective of this research was to determine the influence these parameters, with primary focus on the fuel chemistry, have on the dissolution rate of unirradiated UO{sub 2} under oxidizing repository conditions and compare them to the rates predicted by current dissolution models. Both unirradiated UO{sub 2} and UO{sub 2} doped with varying concentrations of Gd{sub 2}O{sub 3}, to simulate used fuel composition after long time periods when radiolysis has minor contributions to dissolution, were examined. In general, a rise in temperature increased the dissolution rate of UO{sub 2} and had a larger effect on pure UO{sub 2} than on those doped with Gd{sub 2}O{sub 3}. Oxygen dependence was observed in the UO{sub 2} samples with no dopant and increased as the temperature rose; in the doped fuels less dependence was observed. The addition of gadolinia into the UO{sub 2} matrix resulted in a significant decrease in the dissolution rate. The matrix stabilization effect resulting from the dopant proved even more beneficial in lowering the dissolution rate at higher temperatures and dissolved O{sub 2} concentrations in the leachate where the rates would typically be elevated. - Highlights: • UO{sub 2} dissolution rates were measured for a matrix of repository relevant conditions. • Dopants in the UO{sub 2} matrix lowered the dissolution rate. • Reduction in rates by dopants were increased at elevated temperature and O{sub 2} levels. • UO{sub 2} may be overly

Neutron radiography in the IEA-R1 reactor

International Nuclear Information System (INIS)

Pugliesi, R.; Moraes, A.P.V. de; Yamazaki, I.M.; Freitas Acosta, C. de.

1988-08-01

Neutronradiography of several materials have been obtained at the IEA-R1 Nuclear Research Reactor (IPEN-CNEN/SP), by means of two conversion techniques: a) (n, α) at the beam-hole n 0 3 where a collimated thermal neutron beam, exposure area 4 cm x 8cm and flux at the sample 10 5 n/s cm 2 is obtained. The film used was the CN-85 cellulose nitrate coated with lithium tetraborate (conversor). The time irradiation of the film was 15 minutes and in following was eteched during 30 minutes in a NaOH(10%) aqueous solution at a constant temperature of 60 0 C.; b) (n,γ) by using an experimental arrangement installed in the botton of the pool of the reactor. The flux of the collimated neutron beam is 10 5 n/s/cm 2 at the sample and the conversion is made by means of a dysprozium sheet. The film used was Kodak T-5. The irradiation and the transfering time was 2 hours and 20 hours respectively. (author) [pt

A microstructure-dependent model for fission product gas release and swelling in UO2 fuel

International Nuclear Information System (INIS)

Notley, M.J.F.; Hastings, I.J.

1979-06-01

A model for the release of fission gas from irradiated UO2 fuel is presented. It incorporates fission gas diffusion bubble and grain boundary movement,intergranular bubble formation and interlinkage. In addition, the model allows estimates of the extent of structural change and fuel swelling. In the latter, contributions of thermal expansion, densification, solid fission products, and gas bubbles are considered. When included in the ELESIM fuel performance code, the model yields predictions which are in good agreement with data from UO2 fuel elements irradiated over a range of water-cooled reactor conditions: linear power outputs between 40 and 120 kW/m, burnups between 10 and 300 MW.h/kg U and power histories including constant, high-to-low and low-to-high power periods. The predictions of the model are shown to be most sensitive to fuel power (temperature), the selection of diffusion coefficient for fission gas in UO2 and burnup. The predictions are less sensitive to variables such as fuel restraint, initial grain size and the rate of grain growth. (author)

UO2 dissolution rates: A review

International Nuclear Information System (INIS)

McKenzie, W.F.

1992-09-01

This report reviews literature data on UO 2 dissolution kinetics and provides a framework for guiding future experimental studies as well as theoretical modeling studies. Under oxidizing conditions, UO 2 dissolution involves formation of an oxidized surface layer which is then dissolved by formation of aqueous complexes. Higher oxygen pressures or other oxidants are required at higher temperatures to have dissolution rates independent of oxygen pressure. At high oxygen pressures (1-5 atm, 25-70 C), the dissolution rate has a one-half order dependence on oxygen pressure, whereas at oxygen pressures below 0.2 atm, Grandstaff (1976), but nobody else, observed a first-order dependence on dissolution rate. Most people found a first-order dependence on carbonate concentration; Posey-Dowty (1987) found independence of carbonate at pH 7 to 8.2. Dissolution rates increase with temperature except in experiments involving granitic groundwater. Dissolution rates were generally greater under acid or basic conditions than near neutral pH

Studsvik's R2 reactor - Review of activities

Energy Technology Data Exchange (ETDEWEB)

Grounes, Mikael; Tomani, Hans; Graeslund, Christian; Rundquist, Hans; Skoeld, Kurt [Studsvik Nuclear AB, Nykoeping (Sweden)

1993-07-01

A general description of the R2 reactor, its associated facilities and its history is given. The facilities and range of work are described for the following types of activities: fuel testing, materials testing, neutron transmutation doping of silicon, activation analysis, radioisotope production and basic research including thermal neutron scattering, nuclear chemistry and neutron capture radiography. (author)

Core calculations for the upgrading of the IEA-R1 research reactor

International Nuclear Information System (INIS)

Santos, Adimir dos; Perrotta, Jose A.; Bastos, Jose Luis F.; Yamaguchi, Mitsuo; Umbehaun, Pedro E.

1998-01-01

The IEA-R1 Research Reactor is a multipurpose reactor. It has been used for basic and applied research in the nuclear area, training and radioisotopes production since 1957. In 1995, the Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP) took the decision to modernize and upgrade the power from 2 to 5 MW and increase the operational cycle. This work presents the design requirements and the calculations effectuated to reach this goal. (author)

Dissolution of unirradiated UO{sub 2} fuel in synthetic groundwater. Final report (1996-1998)

Energy Technology Data Exchange (ETDEWEB)

Ollila, K. [VTT Chemical Technology, Espoo (Finland)

1999-05-01

This study was a part of the EU R and D programme 1994-1998: Nuclear Fission Safety, entitled `Source term for performance assessment of spent fuel as a waste form`. The research carried out at VTT Chemical Technology was focused on the effects of granitic groundwater composition and redox conditions on UO{sub 2} solubility and dissolution mechanisms. The synthetic groundwater compositions simulated deep granitic fresh and saline groundwaters, and the effects of the near-field material, bentonite, on very saline groundwater. Additionally, the Spanish granite/bentonite water was used. The redox conditions (Eh), which are obviously the most important factors that influence on UO{sub 2} solubility under the disposal conditions of spent fuel, varied from strongly oxidising (air-saturated), anaerobic (N{sub 2}, O{sub 2} < l ppm) to reducing (N{sub 2}, low Eh). The objective of the air-saturated dissolution experiments was to yield the maximum solution concentrations of U, and information on the formation of secondary phases that control the concentrations, with different groundwater compositions. The static batch solubility experiments of long duration (up to 1-2 years) were performed using unirradiated UO{sub 2} pellets and powder. Under anaerobic and reducing conditions, the solubilities were also approached from oversaturation. The results of the oxic, air-saturated dissolution experiments with UO{sub 2} powder showed that the increase in the salinity (< 1.7 M) had a minor effect on the measured steady-state concentrations of U. The concentrations, (1.2 ...2.5) x 10{sup -5} M, were at the level of the theoretical solubility of schoepite or another uranyl oxide hydrate, e.g. becquerelite (possibly Na-polyuranate). The higher alkalinity of the fresh (Allard) composition increased the aqueous U concentration. Only some kind of oxidised U-phase (U{sub 3}O{sub 8}-UO{sub 3}) was identified with XRD when studying possible secondary phases after the contact time of one year

A charge-optimized many-body potential for the U-UO2-O2 system

Science.gov (United States)

Li, Yangzhong; Liang, Tao; Sinnott, Susan B.; Phillpot, Simon R.

2013-12-01

Building on previous charge-optimized many-body (COMB) potentials for metallic α-U and gaseous O2, we have developed a new potential for UO2, which also allows the simulation of U-UO2-O2 systems. The UO2 lattice parameter, elastic constants and formation energies of stoichiometric and non-stoichiometric intrinsic defects are well reproduced. Moreover, this is the first rigid-ion potential that produces the correct deviation of the Cauchy relation, as well as the first classical interatomic potential that is able to determine the defect energies of non-stoichiometric intrinsic point defects in UO2 with an appropriate reference state. The oxygen molecule interstitial in the α-U structure is shown to decompose, with some U-O bonds approaching the natural bond length of perfect UO2. Finally, we demonstrate the capability of this COMB potential to simulate a complex system by performing a simulation of the α-U + O2 → UO2 phase transformation. We also identify a possible mechanism for uranium oxidation and the orientation of the resulting fluorite UO2 structure relative to the coordinate system of orthorhombic α-U.