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Sample records for burnup uo2 pellet

  1. Non-instrumented capsule design of HANARO irradiation test for the high burn-up large grain UO2 pellets

    International Nuclear Information System (INIS)

    Kim, D. H.; Lee, C. B.; Oh, D. S.

    2001-01-01

    Non-instrumented capsule was designed to irradiate the large grain UO 2 pellet developed for the high burn-up LWR fuel in the HANARO in-pile capsule. UO 2 pelletes will be irradiated up to the burn-up higher than 70 MWD/kgU in HANARO. To irradiate the UO 2 pellets up to the burn-up 70 MWD/kgU, need the time about 60 months and ensure the integrity of non-instrumented capsule for 30 months until replace the new capsule. In addition, to satisfy the safety criteria of HANARO such as prevention of ONB(Onset of Nucleate Boiling), fuel melting and wear damage of the capsule during the long term irradiation, design of the non-instrumented capsule was optimized

  2. 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

  3. 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

  4. 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

  5. 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

  6. 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.)

  7. Behaviour of fission gas in the rim region of high burn-up UO2 fuel pellets with particular reference to results from an XRF investigation

    International Nuclear Information System (INIS)

    Mogensen, M.; Walker, C.T.

    1999-01-01

    XRF and EPMA results for retained xenon from Battelle's high burn-up effects program are re-evaluated. The data reviewed are from commercial low enriched BWR fuel with burn-ups of 44.8-54.9 GWd/tU and high enriched PWR fuel with burn-ups from 62.5 to 83.1 GWd/tU. It is found that the high burn-up structure penetrated much deeper than initially reported. The local burn-up threshold for the formation of the high burn-up structure in those fuels with grain sizes in the normal range lay between 60 and 75 GWd/tU. The high burn-up structure was not detected by EPMA in a fuel that had a grain size of 78 μm although the local burn-up at the pellet rim had exceeded 80 GWd/tU. It is concluded that fission gas had been released from the high burn-up structure in three PWR fuel sections with burn-ups of 70.4, 72.2 and 83.1 GWd/tU. In the rim region of the last two sections at the locations where XRF indicated gas release the local burn-up was higher than 75 GWd/tU. (orig.)

  8. Method of manufacturing UO2 pellet

    International Nuclear Information System (INIS)

    Harada, Yuhei; Asami, Yasuji.

    1989-01-01

    The present invention concerns a method of manufacturing UO 2 pellets with less FP gas release and having fine structure for moderating PCMI. At first, oxide nuclear fuel pellets are placed in a sintering furnance and preliminarily sintered in a H 2 gas atmosphere at 1400 - 1600 degC. In this step, sintering is progressed to about 90 % TD, by which closed cells are formed substantially completely. Then, when sintering is further advanced at an identical temperature in a CO 2 gas atmosphere, growth of the crystal grains is advanced at the central portion of the pellets. Then, reductive heat treatment is applied at the identical temperature in a H 2 gas atmosphere. As a result, pellets having a fine double structure with the larger grain size region being in the central portion and smaller grain size region in the outer periphery can be obtained. (I.J.)

  9. Fabrication of ThO2, UO2, and PuO2-UO2 pellets

    International Nuclear Information System (INIS)

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

    1978-01-01

    Fabrication of ThO pellets for EBR-II irradiation testing and fabrication of UO 2 and PuO 2 -UO 2 pellets for United Kingdom Prototype Fast Reactor (PFR) irradiation testing is discussed. Effect of process parameters on density and microstructure of pellets fabricated by the cold press and sinter technique is reviewed

  10. 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

  11. Results of REIMEP '89 UO2 pellet

    International Nuclear Information System (INIS)

    Mayer, K.; Alonso, A.; Bievre, P. de; Lycke, W.; Bolle, W. de; Gallet, M.; Hendrickx, F.

    1991-01-01

    The interest in the safeguards of fissile material focuses on a limited number of compounds which play key roles in the nuclear fuel cycle. Amongst these materials Uranium Dioxide pellets are of considerable importance as they enter the reactors in order to generate energy. In LWR's pellets with an initial 235 U content of about 3 mass % are used, whereas natural or depleted material is applied for the breeding zone in FBR's. The 89/90 round o REIMEP covered Uranium materials with 235 U abundances in the range of natural or depleted material. UO 2 pellets were distributed to 21 laboratories for analysis. The participating laboratories were asked to determine the Uranium content and the isotopic composition of the material. The results reported by the participants are presented as graphs thus giving a picture of the state-of-the-practice

  12. 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

  13. 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

  14. Microstructural change and its influence on fission gas release in high burnup UO 2 fuel

    Science.gov (United States)

    Une, K.; Nogita, K.; Kashibe, S.; Imamura, M.

    1992-06-01

    The microstructural change of UO 2 fuel pellets (burnup: 6-83 GWd/t), base irradiated under LWR conditions, has been studied by detailed postirradiation examinations. The lattice parameter near the fuel rim in the irradiated UO 2 increased with burnup and appeared to become constant beyond about 50 GWd/t. This lattice dilation was mainly due to the accumulation of radiation induced point defects. Moreover, the dislocation density in the UO 2 matrix developed progressively with burnup, and eventually the tangled dislocations organized many sub-grain boundaries in the highest burnup fuel of 83 GWd/t. This sub-grain structure induced by accumulated radiation damage was compatible in appearance with SEM fractography results which revealed sub-divided grains of sub-micron size in as-fabricated grains. The influence of burnup on 85Kr release from the UO 2 fuels has been examined by means of a postirradiation annealing technique. The higher fractional release of high burnup fuels was mainly due to the burnup dependence of the fractional burst release evolved on temperature ramp. The fractional burst release was represented in terms of the square root of burnup from 6 to 83 GWd/t.

  15. Fission gas and iodine release measured up to 15 GWd/t UO2 burnup

    International Nuclear Information System (INIS)

    Appelhans, A.D.

    1983-01-01

    A summary is presented of the measured release of xenon, krypton and iodine up to 15 GWd/t UO 2 burnup for fuel centerline temperatures ranging from 950 to 1800 K, at average linear heat ratings of 15 to 35 kW/m. The IFA-430 is composed of four 1.28-m-long fuel rods containing 10% enriched UO 2 pellet fuel. Two of the fuel rods are connected, top and bottom, to a gas flow system that permits the fission gases released from the fuel pellets to be swept out of the rods during irradiation and measured via gamma spectrometry. The release/burnup increased significantly between 10 and 15 GWd/t burnup. Fuel temperature did not change. Increased releases were due to physical changes in the fuel-surface area. Changes appeared to be due to higher power operation and burnup

  16. 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.)

  17. 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

  18. 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)

  19. 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

  20. 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.)

  1. Factors Affecting the Sintering of UO2 Pellets

    International Nuclear Information System (INIS)

    El-Hakim, E.; Afifi, Y.K.

    1999-01-01

    Sintering of UO 2 pellets is affected by many parameters such as; UO 2 powder parameters, the conditions followed for preparing the green UO 2 pellets and the sintering scheme(heating and cooling rate, soaking time and temperature). The aim of this work is to study the effect of some these parameters on the characteristics of the sintered UO 2 pellets were qualified according to the technical specifications of Candu fuel. Pressed green pellets at different pressing force (15 to 50 k N) were sintered at 1650 ±20 degree for two hours to study the effect of pressing force on the sintered pellets characteristics; visual inspection, pellet dimensions, density and shrinkage ratio. Compacted green pellets at a pressing force of 48 k N were sintered at different sintering temperature (1600± 20 degree, 1650 ±20 degree, 1700± 20 degree) for two hours to study the effect of sintering temperature on the sintered pellets characteristics. The effect of the heating rate (200,300 and 400 degree per hour) on the sintered pellets characteristics was also investigated. It was found that the pressing force used to compact the green pellets had an effect on the density of the sintered pellets. Pellets pressed at 15 k N have a density of 10.3 g/cm 3 while, those pressed at 50 k N have a density of 10.6 g/cm 3. It was observed that increasing the heating rate to 400 degree /h lead to cracked pellets

  2. Modelling the high burnup UO2 structure in LWR fuel

    International Nuclear Information System (INIS)

    Lassmann, K.; Walker, C.T.; Laar, J. van de; Lindstroem, F.

    1995-01-01

    The concept of a burnup threshold for the formation of the high burnup UO 2 structure (HBS) is supported by experimental data, which also reveal that a transition zone exists between the normal UO 2 structure and the fully developed HBS. From the analysis of radial xenon profiles measured by EPMA a threshold burnup is obtained in the range 60-75 GW d/t U. The lower value is considered to be the threshold for the onset of the HBS and the higher value the threshold for the fully developed HBS. Xenon depletion in the transition zone and the fully developed HBS can be described by a simple model. At local burnups above 120 GW d/t U the xenon generated is in equilibrium with the xenon lost to the fission gas pores and the concentration does not fall below 0.25 wt%. The TRANSURANUS burnup model TUBRNP predicts reasonably well the penetration of the HBS and the associated xenon depletion up to a cross section average burnup of approximately 70 GW d/t U. (orig.)

  3. The Width of High Burnup Structure in LWR UO2 Fuel

    International Nuclear Information System (INIS)

    Koo, Yang-Hyun; Lee, Byung-Ho; Oh, Jae-Yong; Sohn, Dong-Seong

    2007-01-01

    The measured data available in the open literature on the width of high burnup structure (HBS) in LWR UO 2 fuel were analyzed in terms of pellet average burnup, enrichment, and grain size. Dependence of the HBS width on pellet average burnup was shown to be divided into three regions; while the HBS width is governed by accumulation of fission damage (i.e., burnup) for burnup below 60 GWd/tU, it seems to be restricted to some limiting value of around 1.5 mm for burnup above 75 GWd/tU due to high temperature which might have caused extensive annealing of irradiation damage. As for intermediate burnup between 60 and 75 GWd/tU, although temperature would not have been so high as to induce extensive annealing, the microstructural damage could have been partly annealed, resulting in the reduction of the HBS width. It was found that both enrichment and grain size also affects the HBS width. However, as long as the pellet average burnup is lower than about 75 GWd/tU, the effect does not appear to be significant for the enrichment and grain size that are typically used in current LWR fuel. (authors)

  4. Homogeneity Study of UO2 Pellet Density for Quality Control

    International Nuclear Information System (INIS)

    Moon, Je Seon; Park, Chang Je; Kang, Kwon Ho; Moon, Heung Soo; Song, Kee Chan

    2005-01-01

    A homogeneity study has been performed with various densities of UO 2 pellets as the work of a quality control. The densities of the UO 2 pellets are distributed randomly due to several factors such as the milling conditions and sintering environments, etc. After sintering, total fourteen bottles were chosen for UO 2 density and each bottle had three samples. With these bottles, the between-bottle and within-bottle homogeneity were investigated via the analysis of the variance (ANOVA). From the results of ANOVA, the calculated F-value is used to determine whether the distribution is accepted or rejected from the view of a homogeneity under a certain confidence level. All the homogeneity checks followed the International Standard Guide 35

  5. Thermal performance prediction of UO2 pellet partly containing 9%w tungsten network

    International Nuclear Information System (INIS)

    Suwardi

    2008-01-01

    Sintered UO 2 exhibits very stable in reactor core compared to UC, UN, U metal and its alloys. However, its thermal conductivity is very low (2.about.5 W/m K), that limits its performance. UO 2 pellet containing Tungsten network invented by Song improves considerably its conductivity. The paper reports an analysis of thermal performance for UO 2 pellet that contains partly or wholly with 9% b. of Tungsten. The tungsten network having a high melting point and excellent thermal conductivity is continuously formed around UO 2 grains. Since the presence of network decreases the amount of fissile material and the burn up of fissile material is higher in the near surface zone of pellet but high temperature zone that releases low conductivity fission gas to the gap located in inner part of pellet, the analysis has been done for different outer radial-portion of tungsten-free pellet. The analysis takes into account the correction factor for pellet conductivity related to both pore and temperature distribution and high burn up effect. The gap conductance has been considered invariable since decrease caused by wider gap size related to lower pellet expansion is compensated by increase caused by fewer of refractory fission gas released. The results (47 kw/m, 40% burnup) show temperature decrease in all of pellet position containing W network. Pellet containing 9%b. tungsten network lower consecutively its center line temperature from 1578 to 1406, 1292, 1231, 1192, 1111, and 1038 deg C for 0, 50, 67, 75, 80, 90, and 100 % portion of network. An 80 to 90 % portion of inner pellet containing tungsten network can be considered a best fuel design. This preliminary analysis is prospective and more realistic one is recommended. (author)

  6. 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

  7. Preparation of Fluidization Feed of UO2 Pellets by Oxidation

    International Nuclear Information System (INIS)

    Rachmat-Pratomo; H, Didiek; Suwondo, B; Sigit

    2000-01-01

    The investigation of oxidation of uranium dioxide (UO 2 ) pellets to thetri uranium octoxide (U 3 O 8 ) powder had been carried. Several factor suchtemperature, time of oxidation and the concentration of air are important.The oxidation of UO 2 pellet are carried out on electric furnace atatmosphere as media. The oxidation temperature started at 300 o C, 400 o C,500 o C, and 600 o C along 1 hour. The time oxidation removed to 2 hours and3 hours. The efficiency of oxidation are the ratio of the weight of thepowder product are the uranium content, true density, and specific surfacearea. Result the optimum temperature are 500 o C along 3 hours, uraniumcontent : 84.78%, true density: 8.8293 g/cm 3 and specific surface area :0.389071 m 2 /g. (author)

  8. Effect of continuous change of sintering atmosphere on the grain growth of Cr-doped UO2 pellets

    International Nuclear Information System (INIS)

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

    2010-01-01

    Cr-doped UO 2 pellet is one of the promising candidates for the high burn-up fuel in commercial LWRs. Major nuclear fuel vendors of such as AREVA or Westinghouse initiated the development of Cr-doped or Cr-containing additives doped UO 2 pellets since at the mid of 90's. Now, qualification programs are on-going to provide these pellets commercially. The main characteristics of the Cr-doped pellets are large-grain and visco-plasticity. Large grain pellet can reduce the corrosive fission gas release at high burn up. Viscoplastic soft pellets can lower the pressure to a cladding caused by a thermal expansion of a pellet at an elevated temperature during transient operations. Those advantages can provide room for additional power uprates and high burnup limits. Especially, PCI resistance improvement can be achieved by enlarging the pellet grain size and enhancing the fuel deformation at an elevated temperature. In this paper, to study the effect of oxygen partial pressure on grain growth in Cr-doped UO 2 pellets, Cr- doped UO 2 samples have been sintered with and without a step-wise change of sintering atmospheres. An introduction of a step-wise variation of oxygen partial pressure during the sintering enhances the grain growth of UO 2 pellets greatly. This step-wise sintering effect has been explained in terms of a continuous increase of Cr concentration along the grain boundary. The observed grain growth behavior under step-wisely changed sintering atmospheres demonstrates the possibility of reducing the amount of Cr 2 O 3 to minimum via control of oxygen partial pressure while keeping the large grain size

  9. 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

  10. 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

  11. quality assurance calculation in UO2 pellet manufacturing process

    International Nuclear Information System (INIS)

    Can, S.; Acarkan, S.; Guereli, L. and others

    1997-01-01

    A process qualification plan is prepared for preparation of quality assurance documentation in accordance with ISO-9000 series of standards, for sintered UO 2 pellets manufactured in the Nuclear Fuel Technology Department. The objectives of this plan are to determine quantitatively and statistically process capability of the pellet production, to check product properties (are) in conformance with specifications at the pre-( ) confidence levels, to prepare necessary documents and to assess the results. The product properties taking into account are chemical composition, cracks, density, microstructure and grain size. The statistical parameters used for qualification element of quality assurance are calculated.Statistical values for sintered pellets are: LENGTH/WEIGHT/DIAMETER/DENSITY/%TD: MEAN:13,395/16,808/12,293/10,679/97,400 STD:0,1651/ 0,252/0,0212/0,015/0,140. It was seen that sintered pellets manufactured in the Nuclear Fuel Technology Department meet the criteria within 95% confidence level. In this paper specifications, criteria and calculations will be explained in detail

  12. Measurement of gamma attenuation coefficients in UO2 and zirconium for self-absorption corrections of burn-up determination

    International Nuclear Information System (INIS)

    Podest, M.; Klima, J.; Stecher, P.; Stecherova, E.

    1978-01-01

    UO 2 pellets from ALUOX fuel elements were used in measuring the absorption coefficient of gamma radiation in UO 2 . The results of measurements of the energy dependence of the linear absorption coefficient (within 622 to 796 keV) and of the dependence on pellet density showed that in the given density interval the absorption coefficient was almost constant. The density interval was chosen to be typical for pellet fuel used in water cooled and water moderated power reactors. The results are also shown of the dependence of the mass absorption coefficient of gamma radiation in Zr on radiation energy and compared with the mass absorption coefficient of Mo; these also showed the independence of the absorption coefficient on density. The linear and mass absorption coefficients of UO 2 are considerably high and correspond approximately to the absorption coefficient of lead. For the measured energy range the variation of absorption coefficient is about 40%, which causes errors in burnup determination. The efficiency was also determined of Ge(Li) detectors for the energy range 0.5 to 1.2 MeV. The determination of the above coefficients was used for improving the gamma fuel scanning technique in determining the activity and burnup of spent fuel elements. (J.P.)

  13. Review of the effects of burnup on the thermal conductivity of UO2

    International Nuclear Information System (INIS)

    Lokken, R.O.; Courtright, E.L.

    1976-01-01

    The general trends which relate changes in thermal conductivity of UO 2 fuel as a function of temperature and burnup can be summarized as follows: (1) At temperatures below 500 0 C, reductions in UO 2 thermal conductivity relative to the unirradiated values can be expected up to a saturation level of approximately 10 19 fissions/cc. (2) At temperatures above 500 0 C, the thermal conductivity will undergo little change at low burnups, (less than 10 19 fissions/cc) but at higher exposures some decrease can be expected which should, in turn, diminish with increasing temperature. (3) A review of the data reported by Berman on the ThO 2 --UO 2 fuel indicates that the basic behavior is the same as for UO 2 in the temperature range of major interest. The applicability of this data to LWR UO 2 fuel is somewhat questionable because of basic physical property differences, and limited data on irradiation effects, and would not seem to support concerns that the effects of burnup on thermal conductivity for LWR fuel may be of more significance than currently believed. (4) A mathematical expression of the type proposed by Daniel and Cohen seems to provide a reasonable approximation for the behavioral trends reported in the literature which relate changes in thermal conductivity to increasing burnup in certain temperature regimes. Calculations indicate that only small incremental increases in the fuel centerline temperature might be expected if burnup effects are taken into account

  14. Fabrication, characteristics, and in-pile performance of UO2 pellets prepared from dry route powder

    International Nuclear Information System (INIS)

    Chotard, A.; Ledac, A.; Bernardin, M.

    1991-01-01

    The dry route conversion process of UF 6 to sinterable UO 2 powder has been used in France on a large scale for more than 10 years for the fabrication of PWR fuels. Thus, our fabrication and irradiation experience relates to more than 10,000 tons of fuel. As everyone knows, the dry route conversion process only involves gas-gas and gas-solid reactions which present the advantage of producing very little contaminated wastes and no liquid effluents. Powders obtained by this process are characterized by: - a very high purity, - a low specific surface area (around 2 m 2 /g), therefore a high resistance to spontaneous oxidation, - a good compressibility, - a very high sinterability (.98% T.D.), - a very high reproducibility. This powder also shows a high fineness which leads to very homogeneous blends with additives like pore former, U 3 O 8 or Gd 2 O 3 . On the other hand this fineness requires a granulation step which is actually not a disadvantage since it allows to adjust the granulate size to optimize the filling of press dies and so as to guarantee a good stability of the pellet dimensions and density. This pelletizing process leads to pellets characterized by: - a good thermal stability (0.5% T.D. after 34 hours at 1700degC), - no open porosity, - low H 2 content (0,3 ppm), - an homogeneous microstructure (grain size and porosity). Such characteristics mean that the UO 2 pellets from dry route conversion present an excellent in pile behaviour for high burnup up to 58,000 MWd/MtU in commercial plant, with: - low fission gas release, - good dimensional stability (densification, swelling), of which examples and results of PIE are described in the paper. The qualities of the dry route conversion powder and its flexibility of use make it possible to consider adjustment of the pellet characteristics, mainly: density, grain size and pore size distribution for specific uses or performance upgrade. (orig.)

  15. 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

  16. Improvement of fuel-element reliability by insertion of UO2 microspheres in the gap between pellet and clad

    International Nuclear Information System (INIS)

    Mehedinteanu, S.; Glodeanu, F.; Dobos, I.

    1979-01-01

    With the accumulation of power reactor fuel operating experience, the study of the PCI phenomenon and the development of remedies have become important items in fuel research and development everywhere. The 'power-ramp' failure has drawn attention to the problem of obtaining high reliability from high burn-up fuel rods. Considerable attention has been paid to minimizing the cladding stresses imparted by fuel pellets during the power ramp. The paper describes a new concept of pellet-clad bonding by insertion of UO 2 microspheres in the gap. It is pointed out that the main advantages of this concept are: the low friction coefficient between pellet and clad; the accomodation of cracked pellet expansion by local microyielding of irradiation-embrittled clad; the reduced ridge height by use of undished pellets or other pellet shape; that the fine-sized UO 2 microspheres infiltrate around the pellets thus permitting the use of cracked or chipped pellets and also sintered pellets without the previously required grinding step needed for accurate sizing, etc. (author)

  17. Simulation of High Burnup Structure in UO2 Using Potts Model

    International Nuclear Information System (INIS)

    Oh, Jae Yong; Koo, Yang Hyun; Lee, Byung Ho

    2009-01-01

    The evolution of a high burnup structure (HBS) in a light water reactor (LWR) UO 2 fuel was simulated using the Potts model. A simulation system for the Potts model was defined as a two-dimensional triangular lattice, for which the stored energy was calculated from both the irradiation damage of the UO 2 matrix and the formation of a grain boundary in the newly recrystallized small HBS grains. In the simulation, the evolution probability of the HBS is calculated by the system energy difference between before and after the Monte Carlo simulation step. The simulated local threshold burnup for the HBS formation was 62 MWd/kgU, consistent with the observed threshold burnup range of 60-80 MWd/kgU. The simulation revealed that the HBS was heterogeneously nucleated on the intergranular bubbles in the proximity of the threshold burnup and then additionally on the intragranular bubbles for a burnup above 86 MWd/kgU. In addition, the simulation carried out under a condition of no bubbles indicated that the bubbles played an important role in lowering the threshold burnup for the HBS formation, thereby enabling the HBS to be observed in the burnup range of conventional high burnup fuels

  18. Fission gas and iodine release measured in IFA-430 up to 15 GWd/t UO2 burnup

    International Nuclear Information System (INIS)

    Appelhans, A.D.; Turnbull, J.A.; White, R.J.

    1983-01-01

    The release of fission products from fuel pellets to the fuel-cladding gap is dependent on the fuel temperature, the power (fission rate) and the burnup (fuel structure). As part of the US Nuclear Regulatory Commission's Fuel Behavior Program, EG and G Idaho, Inc., is conducting fission product release studies in the Heavy Boiling Water Reactor in Halden, Norway. This paper presents a summary of the results up to December, 1982. The data cover fuel centerline temperatures ranging from 700 to 1500 0 C for average linear heat ratings of 16 to 35 kW/m. The measurements have been performed for the period between 4.2 and 14.8 GWd/t UO 2 of burnup of the Instrumented Fuel Assembly 430 (IFA-430). The measurement program has been directed toward quantifying the release of the short-lived radioactive noble gases and iodines

  19. 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

  20. Calculation of pellet radial power distributions with a Monte Carlo burnup code

    International Nuclear Information System (INIS)

    Suzuki, Motomu; Yamamoto, Toru; Nakata, Tetsuo

    2010-01-01

    The Japan Nuclear Energy Safety Organization (JNES) has been working on an irradiation test program of high-burnup MOX fuel at Halden Boiling Water Reactor (HBWR). MOX and UO 2 fuel rods had been irradiated up to about 64 GWd/t (rod avg.) as a Japanese utilities research program (1st phase), and using those fuel rods, in-situ measurement of fuel pellet centerline temperature was done during the 2nd phase of irradiation as the JNES test program. As part of analysis of the temperature data, power distributions in a pellet radial direction were analyzed by using a Monte Carlo burnup code MVP-BURN. In addition, the calculated results of deterministic burnup codes SRAC and PLUTON for the same problem were compared with those of MVP-BURN to evaluate their accuracy. Burnup calculations with an assembly model were performed by using MVP-BURN and those with a pin cell model by using SRAC and PLUTON. The cell pitch and, therefore, fuel to moderator ratio in the pin cell calculation was determined from the comparison of neutron energy spectra with those of MVP-BURN. The fuel pellet radial distributions of burnup and fission reaction rates at the end of the 1st phase irradiation were compared between the three codes. The MVP-BURN calculation results show a large peaking in the burnup and fission rates in the pellet outer region for the UO 2 and MOX pellets. The SRAC calculations give very close results to those of the MVP-BURN. On the other hand, the PLUTON calculations show larger burnup for the UO 2 and lower burnup for the MOX pellets in the pellet outer region than those of MVP-BURN, which lead to larger fission rates for the UO 2 and lower fission rates for the MOX pellets, respectively. (author)

  1. Kinetic Monte Carlo Potts Model for Simulating a High Burnup Structure in UO2

    International Nuclear Information System (INIS)

    Oh, Jae-Yong; Koo, Yang-Hyun; Lee, Byung-Ho

    2008-01-01

    A Potts model, based on the kinetic Monte Carlo method, was originally developed for magnetic domain evolutions, but it was also proposed as a model for a grain growth in polycrystals due to similarities between Potts domain structures and grain structures. It has modeled various microstructural phenomena such as grain growths, a recrystallization, a sintering, and so on. A high burnup structure (HBS) is observed in the periphery of a high burnup UO 2 fuel. Although its formation mechanism is not clearly understood yet, its characteristics are well recognized: The HBS microstructure consists of very small grains and large bubbles instead of original as-sintered grains. A threshold burnup for the HBS is observed at a local burnup 60-80 Gwd/tM, and the threshold temperature is 1000-1200 .deg. C. Concerning a energy stability, the HBS can be created if the system energy of the HBS is lower than that of the original structure in an irradiated UO 2 . In this paper, a Potts model was implemented for simulating the HBS by calculating system energies, and the simulation results were compared with the HBS characteristics mentioned above

  2. Burn-up credit applications for UO2 and MOX fuel assemblies in AREVA/COGEMA

    International Nuclear Information System (INIS)

    Toubon, H.; Riffard, C.; Batifol, M.; Pelletier, S.

    2003-01-01

    For the last seven years, AREVA/COGEMA has been implementing the second phase of its burn-up credit program (the incorporation of fission products). Since the early nineties, major actinides have been taken into account in criticality analyses first for reprocessing applications, then for transport and storage of fuel assemblies Next year (2004) COGEMA will take into account the six main fission products (Rh103, Cs133, Nd143, Sm149, Sm152 and Gd155) that make up 50% of the anti-reactivity of all fission products. The experimental program will soon be finished. The new burn-up credit methodology is in progress. After a brief overview of BUC R and D program and COGEMA's application of the BUC, this paper will focus on the new burn-up measurement for UO2 and MOX fuel assemblies. It details the measurement instrumentation and the measurement experiments on MOX fuels performed at La Hague in January 2003. (author)

  3. Analysis of UO2 fuel structure for low and high burn-up and its impact on fission gas release

    International Nuclear Information System (INIS)

    Szuta, M.; El-Koliel, M.S.

    1999-01-01

    During irradiation, uranium dioxide (UO 2 ) fuel undergo important restructuring mainly represented by densification and swelling, void migration, equiaxed grain growth, grain subdivision, and the formation of columnar grains. The purpose of this study is to obtain a comprehensive picture of the phenomenon of equiaxed grain growth in UO 2 ceramic material. The change of the grain size in high-density uranium dioxide as a function of temperature, initial grain size, time, and burnup is calculated. Algorithm of fission gas release from UO 2 fuel during high temperature irradiation at high burnup taking into account grain growth effect is presented. Theoretical results are compared with experimental data. (author)

  4. Effects of two types of dryer on ADU and UO2 pellet manufacture

    International Nuclear Information System (INIS)

    Wu Zhiming; He Zhengjie

    1995-05-01

    The concepts of spray drying process and pebble-bed fluidized drying process for ADU slurry is presented. And the effects of ADU powder and UO 2 powder/pellet by these processes using the statistic results from series production are discussed. It is believed that these drying methods have no influence on structure and shape of ADU particle, and thereby no difference will be made to the properties of UO 2 powder and pellet. Thus, spray drying process can really be replaced by pebble-bed fluidized drying process. (10 figs., 6 tabs.)

  5. PIE and separate effect test of high burnup UO2 fuel

    International Nuclear Information System (INIS)

    Yang, Yong Sik; Kim, S.K.; Kim, D.H.

    2005-01-01

    To investigate the performance of a high burnup UO 2 fuel, the highest burnup fuel assembly in KOREA was transported to the PIE facility in KAERI. It was a 17·17 fuel assembly irradiated at the Ulchin Unit 2 PWR. The peak fuel rod average burnup was about 57MWd/kgU and locally 65MWd/kgU. The general PIE was performed to investigate the fuel rod irradiation performance. Fission gas release, burnup, oxide thickness, hydrogen pickup, CRUD, and density change were measured by destructive of non-destructive test. Microstructure change, bubble and pore size distributions were observed by optical microscopy, SEM and EPMA. All generated and available PIE results were used to verify high burnup fuel performance code INFRA. Several rods were cut for additional separate effect test. For the high burnup fission gas release behaviour analysis, annealing apparatus were developed and installed in hot cell and preliminary test was performed. In addition to current apparatus new induction furnace will be installed in hot cell to investigate the high temperature and transient fission gas release behaviour. Ring tensile test was performed to analyze the material property degradation which caused by the oxidation and hydride, and additional mechanical tests will be performed. (Author)

  6. The effect of U3O-8 addition on the UO2 pellet

    International Nuclear Information System (INIS)

    Indrati, Y.T.; Syarif, D. G.; Handayani, A.

    1998-01-01

    The purpose of varied U 3 O 8 addition on the UO 2 pellet fabrication is to from 1-3 mu pores. The green pellets, compacted with 3 ton/cm 2 , are a mixture powder of UO 2 , TiO 2 (0.1% weight) and varied U 3 O 8 (0-12.5% weight). The green pellets were presintered by H2 atmosphere. The presintered pellets were put on the ceramic crucibles and than those were put on the SS 316 tube with argon atmosphere. The 1400 o C sintering was hold with the soaking time 3 hr and the same rate of heating and cooling 150 o C/hr. The UO 2 pellet with 5% (weight) U 3 O 8 addition has 95.17% of theoretic density and 548.4 ±6.57 VH. Based on the identification of microstructure of pellet, it is not acceptable for nuclear fuel although pellet has 10.02 mu on grain size and 1.3 mu on closed pore size. By the diffractometer X-ray, crystal structure of pellet is face centered cubic (FCC) with the O/U ratio is 2.08

  7. Corrosion behaviour of the UO2 pellet in corrosive solutions using electrochemical Technique

    International Nuclear Information System (INIS)

    Taftanzani, A.; Sucipto; Lahagu, F.; Irianto, B.

    1996-01-01

    The UO 2 electrodes has been made from the local product of UO 2 pellets. The corrosion behaviour of the UO 2 pellets is affected by solution, by pH value and by concentration of salt solution. Investigation into corrosion behaviour of UO 2 electrodes have been carried out in saturated salt solutions using electrochemical technique. The saturated solutions have been made from salts NaCl, Na 2 CO 3 , Na 2 SO 4 and Na 3 PO 4 . The pH value have been done over range 1 pH 10 and the salt concentration (C) over range 0,001 mol/l C 1,0 mol/l, Na 2 CO 3 solution produced the lowest corrosion rates of UO 2 pellets. Those rates were relative constant in the range of pH = 4 - 8. The results indicate an influence of the Na 2 CO 3 concentrations on the corrosions on the corrosion rate, and the lowest rates occur in 0,10 mol/l Na 2 CO 3 . The lowest corrosion rate was 0.3388 mil/year in 0.10 mol/l Na 2 CO 3 by pH = 4. (author)

  8. Threshold burnup for recrystallization and model for rim porosity in the high burnup UO2 fuel

    International Nuclear Information System (INIS)

    Lee, Byung Ho; Koo, Yang Hyun; Sohn, Dong Seong

    1998-01-01

    Applicability of the threshold burnup for rim formation was investigated as a function of temperature by Rest's model. The threshold burnup was the lowest in the intermediate temperature region, while on the other temperature regions the threshold burnup is higher. The rim porosity was predicted by the van der Waals equation based of the rim pore radius of 0.75μm and the overpressurization model on rim pores. The calculated centerline temperature is in good agreement with the measured temperature. However, more efforts seem to be necessary for the mechanistic model of the rim effect including rim growth with the fuel burnup

  9. Sphere-pac versus pellet UO2 fuel in de Dodewaard BWR

    International Nuclear Information System (INIS)

    Linde, A. van der.

    1989-04-01

    Comparative testing of UO 2 sphere-pac and pellet fuel rods under LWR conditions has been jointly performed by the Netherlands Utilities Research Centre (KEMA) in Arnhem, the Netherlands Energy Research Foundation (ECN) at Petten and the Netherlands Joint Nuclear Power Utility (GKN) at Dodewaard. This final report summarizes the highlights of this 1968-1988 program with strong emphasis on the fuel rods irradiated in the Dodewaard BWR. The conclusion reached is that under normal LWR conditions sphere-pac UO 2 in LWR fuel rods offers better resistance against stress corrosion cracking of the cladding, but that under fast, single step, power ramping conditions pellet UO 2 in LWR fuel rods has a better resistance against hoop stress failure of the cladding. 128 figs., 36 refs., 19 tabs

  10. Fabrication of ThO2 and ThO2-UO2 pellets for proliferation resistant fuels

    International Nuclear Information System (INIS)

    Matthews, R.B.; Davis, N.C.

    1979-10-01

    To meet this objective, batches of ThO 2 powders were compared and milling parameters, pressing and sintering conditions were established. A method for blending ThO 2 and UO 2 into homogeneous powders that press and sinter into 95% TD pellets was determined. The effect of UO 2 additions on ThO 2 -UO 2 pellet properties was determined and a process for fabricating irradiation test quality ThO 2 -20 wt% UO 2 pellets containing CaO as a dissolution aid was established

  11. Study on factors affecting sintering density of Gd2O3-UO2 pellets

    International Nuclear Information System (INIS)

    Zhu Shuming; Zou Congpei; Yang Jing; Yang Youqing; Mei Xiaohui

    1996-02-01

    The sintered density of Gd 2 O 3 -UO 2 burnable poison fuel pellets is an important quality index and is one of main QC items. Therefore, the efforts were made to investigate the factors affecting the sintered density of Gd 2 O 3 -UO 2 , that is, the influences of pre-treatment of Gd 2 O 3 powder, additives, mixing methods and time, sintering atmosphere, sintering temperature and time on the final density of Gd 2 O 3 UO 2 pellets contained 0, 3%, 7% and 10% (mass percentage) Gd 2 O 3 . The results show: the pre-treatment is useful for improving the distribution of Gd 2 O 3 ; the additive of ammonium oxalate will effectively adjust the density of pellets; 1750 degree C is the suitable sintering temperature. The proper process parameters have been obtained, and the Gd 2 O 3 -UO 2 pellets prepared for in-pile irradiation test meet the design requirements for the density (93.5%∼96.5% of T.D.), homogeneity, microstructure, etc. (8 refs., 3 figs., 8 tabs.)

  12. A study of the effectiveness of hand protection when handling UO2 fuel pellets

    International Nuclear Information System (INIS)

    Washington, R.R.; Sullivan, D.F.

    1981-01-01

    Simple tests were performed to estimate the effectiveness of various forms of hand protection in reducing skin doses when handling UO 2 fuel pellets. Household rubber gloves (rubberized cotton) appeared to be the most effective of the varieties tested. Nylon gloves and latex finger cots were least effective. (author)

  13. The dissolution of unirradiated UO2 fuel pellets under simulated disposal conditions

    International Nuclear Information System (INIS)

    Ollila, K.; Leino-Forsman, H.

    1993-03-01

    The dissolution behaviour of unirradiated UO 2 pellets was studied as a function of water composition under oxidizing and reducing conditions at 25 deg C. The waters included deionized water as the reference water, sodium bicarbonate solutions with varying bicarbonate content, and two different synthetic groundwaters. The release of uranium was measured during static batch dissolution experiments of long duration (3-4 years)

  14. 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

  15. Homogeneity and microstructure study of Gd2O3-UO2 pellets

    International Nuclear Information System (INIS)

    Pan Ying; Gao Dihua; Guo Yibai; Zhu Shuming

    1994-10-01

    The microstructure of Gd 2 O 3 -UO 2 pellets (0∼10 wt%) prepared in different conditions, the homogeneity distribution of Gd 2 O 3 in the pellets and the lattice parameter of solid solution are studied by metalloscope, WDS, EDAX, SEM-image processing system, XRD and image analyzer. The theoretical density has been calculated. The effect of size and content of Gd 2 O 3 particles, the blend process, the sintering temperature and time, and the sintering atmosphere on the microstructure of Gd 2 O 3 pellets and the homogeneity of Gd 2 O 3 in the pellets are studied. (16 refs., 10 figs., 8 tabs.)

  16. 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

  17. Analysis of burnup and isotopic compositions of BWR 9 x 9 UO2 fuel assemblies

    International Nuclear Information System (INIS)

    Suzuki, M.; Yamamoto, T.; Ando, Y.; Nakajima, T.

    2012-01-01

    In order to extend isotopic composition data focusing on fission product nuclides, measurements are progressing using facilities of JAEA for five samples taken from high burnup BWR 9 x 9 UO 2 fuel assemblies. Neutronics analysis with an infinite assembly model was applied to the preliminary measurement data using a continuous-energy Monte Carlo burnup calculation code MVP-BURN with nuclear libraries based on JENDL-3.3 and JENDL-4.0. The burnups of the samples were determined to be 28.0, 39.3, 56.6, 68.1, and 64.0 GWd/t by the Nd-148 method. They were compared with those calculated using node-average irradiation histories of power and in-channel void fractions which were taken from the plant data. The comparison results showed that the deviations of the calculated burnups from the measurements were -4 to 3%. It was confirmed that adopting the nuclear data library based on JENDL-4.0 reduced the deviations of the calculated isotopic compositions from the measurements for 238 Pu, 144 Nd, 145 Nd, 146 Nd, 148 Nd, 134 Cs, 154 Eu, 152 Sm, 154 Gd, and 157 Gd. On the other hand, the effect of the revision in the nuclear. data library on the neutronics analysis was not significant for major U and Pu isotopes. (authors)

  18. UO2 fuel pellets fabrication via Spark Plasma Sintering using non-standard molybdenum die

    Science.gov (United States)

    Papynov, E. K.; Shichalin, O. O.; Mironenko, A. Yu; Tananaev, I. G.; Avramenko, V. A.; Sergienko, V. I.

    2018-02-01

    The article investigates spark plasma sintering (SPS) of commercial uranium dioxide (UO2) powder of ceramic origin into highly dense fuel pellets using non-standard die instead of usual graphite die. An alternative and formerly unknown method has been suggested to fabricate UO2 fuel pellets by SPS for excluding of typical problems related to undesirable carbon diffusion. Influence of SPS parameters on chemical composition and quality of UO2 pellets has been studied. Also main advantages and drawbacks have been revealed for SPS consolidation of UO2 in non-standard molybdenum die. The method is very promising due to high quality of the final product (density 97.5-98.4% from theoretical, absence of carbon traces, mean grain size below 3 μm) and mild sintering conditions (temperature 1100 ºC, pressure 141.5 MPa, sintering time 25 min). The results are interesting for development and probable application of SPS in large-scale production of nuclear ceramic fuel.

  19. Grain boundary corrosion and alteration phase formation during the oxidative dissolution of UO2 pellets

    International Nuclear Information System (INIS)

    Wronkiewicz, D.J.; Buck, E.C.; Bates, J.K.

    1996-01-01

    Alteration behavior of UO 2 pellets following reaction under unsaturated drip-test conditions at 90 C for up to 10 years was examined by solid phase and leachate analyses. Sample reactions were characterized by preferential dissolution of grain boundaries between the original press-sintered UO 2 granules comprising the samples, development of a polygonal network of open channels along the intergrain boundaries, and spallation of surface granules that had undergone severe grain boundary corrosion. The development of a dense mat of alteration phases after 2 years of reaction trapped loose granules, resulting in reduced rates of particulate U release. The paragenetic sequence of alteration phases that formed on the present samples was similar to that observed in surficial weathering zones of natural uraninite (UO 2 ) deposits, with alkali and alkaline earth uranyl silicates representing the long-term solubility-limiting phases for U in both systems

  20. 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

  1. 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

  2. Thermal conductivity evaluation of high burnup mixed-oxide (MOX) fuel pellet

    International Nuclear Information System (INIS)

    Amaya, Masaki; Nakamura, Jinichi; Nagase, Fumihisa; Fuketa, Toyoshi

    2011-01-01

    The thermal conductivity formula of fuel pellet which contains the effects of burnup and plutonium (Pu) addition was proposed based on the Klemens' theory and reported thermal conductivities of unirradiated (U, Pu) O 2 and irradiated UO 2 pellets. The thermal conductivity of high burnup MOX pellet was formulated by applying a summation rule between phonon scattering parameters which show the effects of plutonium addition and burnup. Temperature of high burnup MOX fuel was evaluated based on the thermal conductivity integral which was calculated from the above-mentioned thermal conductivity formula. Calculated fuel temperatures were plotted against the linear heat rates of the fuel rods, and were compared with the fuel temperatures measured in a test reactor. Since both values agreed well, it was confirmed that the proposed thermal conductivity formula of MOX pellets is adequate.

  3. The UO2 pellets plant experimental background of the established process

    International Nuclear Information System (INIS)

    Aparicio Arroyo, E.; Alonso Folgueras, J. A.

    1969-01-01

    An account of the UO 2 research and development carried out at the JEN is first given. This includes a 10 tons/year pellet Plant construction. Experimental background of the process is established, pointing out both milling advantages and risks, granulation devices, automatic press selection, binder removing and sintering furnaces. Origin, surface area, grain size and 0/U rate are considered as raw material reception parameters, although this process shows a wide scope. (Author) 13 refs

  4. Dissolution of intact UO2 pellet in batch and rotary dissolver conditions

    International Nuclear Information System (INIS)

    Jayendra Kumar Gelatar; Bijendra Kumar; Sampath, M.; Shekhar Kumar; Kamachi Mudali, U.; Natarajan, R.

    2015-01-01

    Comparative dissolution of intact un-irradiated UO 2 pellet of PHWR fuel dimensions was performed in batch and dynamic rotary dissolver conditions in aqueous nitric acid solutions at elevated temperatures. The extent of dissolution was estimated by determining the uranium concentration of the resulting aqueous solution. It was observed that rate of dissolution was much faster in dynamic conditions as compared to static batch conditions. (author)

  5. Modeling fission gas release in high burnup ThO2-UO2 fuel

    International Nuclear Information System (INIS)

    Long, Y.; Yuan, Y.; Pilat, E.E.; Rim, C.S.; Kazimi, M.S.

    2001-01-01

    A preliminary fission gas release model to predict the performance of thoria fuel using the FRAPCON-3 computer code package has been formulated. The following modeling changes have been made in the code: - Radial power/burnup distribution; - Thermal conductivity and thermal expansion; - Rim porosity and fuel density; - Diffusion coefficient of fission gas in ThO 2 -UO 2 fuel and low temperature fission gas release model. Due to its lower epithermal resonance absorption, thoria fuel experiences a much flatter distribution of radial fissile products and radial power distribution during operation as compared to uranian fuel. The rim effect and its consequences in thoria fuel, therefore, are expected to occur only at relatively high burnup levels. The enhanced conductivity is evident for ThO 2 , but for a mixture the thermal conductivity enhancement is small. The lower thermal fuel expansion tends to negate these small advantages. With the modifications above, the new version of FRAPCON-3 matched the measured fission gas release data reasonably well using the ANS 5.4 fission gas release model. (authors)

  6. 3D Finite Elements Modelling for Design and Performance Analysis of UO2 Pellets

    International Nuclear Information System (INIS)

    Demarco, G.L.; Marino, A.C.; Demarco, G.L.; Marino, A.C.

    2011-01-01

    The geometry of a fuel pellet is a compromise among the intention to maximize UO 2 content and minimize the temperature gradient taking into account the thermomechanical behaviour, the economy, and the safety of the fuel management during and after irradiation. Dishing, shoulders, chamfers, and/or a central hole on a cylinder with an improved l/d relation (length of the pellet/diameter) are introduced in order to optimize the shape of the pellet. The Me Com tools coupled with the BaCo code constitutes a complete system for the 3D analysis of the stress strain state of the pellet under irradiation. CANDU and PHWR MOX fuel will be used to illustrate the excellent qualitative agreement between experimental

  7. Transient fission gas release from UO2 fuel for high temperature and high burnup

    International Nuclear Information System (INIS)

    Szuta, M.

    2001-01-01

    In the present paper it is assumed that the fission gas release kinetics from an irradiated UO 2 fuel for high temperature is determined by the kinetics of grain growth. A well founded assumption that Vitanza curve describes the change of uranium dioxide re-crystallization temperature and the experimental results referring to the limiting grain size presented in the literature are used to modify the grain growth model. Algorithms of fission gas release due to re-crystallization of uranium dioxide grains are worked out. The defect trap model of fission gas behaviour described in the earlier papers is supplemented with the algorithms. Calculations of fission gas release in function of time, temperature, burn-up and initial grain sizes are obtained. Computation of transient fission gas release in the paper is limited to the case where steady state of irradiation to accumulate a desired burn-up is performed below the temperature of re-crystallization then the subsequent step temperature increase follows. There are considered two kinds of step temperature increase for different burn-up: the final temperature of the step increase is below and above the re-crystallization temperature. Calculations show that bursts of fission gas are predicted in both kinds. The release rate of gas liberated for the final temperature above the re-crystallization temperature is much higher than for final temperature below the re-crystallization temperature. The time required for the burst to subside is longer due to grain growth than due to diffusion of bubbles and knock-out release. The theoretical results explain qualitatively the experimental data but some of them need to be verified since this sort of experimental data are not found in the available literature. (author)

  8. Proposal for Ultrasonic Technique for evaluation elastic constants in UO2 pellets

    International Nuclear Information System (INIS)

    Lopes, Alessandra Susanne Viana Ragone; Baroni, Douglas Brandao; Bittencourt, Marcelo de Siqueira Queiroz; Souza, Mauro Carlos Lopes

    2015-01-01

    Pellets of uranium dioxide are used as fuel in nuclear power reactors, in which are exposed to high thermal gradients. This high energy will initiate fusion in the central part of the pellet. The expansion of the uranium dioxide pellets, resulting from fission products, can cause fissures or cracks, therefore, the study of their behavior is important. This work aims to develop and propose an ultrasonic technique to evaluate the elastic constants of UO 2 pellets. However, because of the difficulties in handling nuclear material, we proposed an initial study of alumina specimens. Alumina pellets are also ceramic material and their porosity and dimensions are in the similar range of dioxide uranium pellets. They also are used as thermal insulation in the fuel rods, operating under the same conditions. They were fabricated and used in two different sets of 10 alumina pellets with densities of 92% and 96%. The developed ultrasonic technique evaluates the traveling time of ultrasonic waves, longitudinal and transverse, and correlates the observed time and the elastic constants of the materials. Equations relating the speed of the ultrasonic wave to the elastic modulus, shear modulus and Poisson's ratio have led to these elastic constants, with graphics of correlation that showed excellent agreement with the literature available for Alumina. In view of the results and the ease of implementation of this technique, we believe that it may easily be used for dioxide uranium pellets, justifying further studies for that application. (author)

  9. Isotopic analyses and calculation by use of JENDL-3.2 for high burn-up UO2 and MOX spent fuels

    International Nuclear Information System (INIS)

    Sasahara, Akihiro; Matsumura, Tetsuo; Nicolaou, G.; Betti, M.; Walker, C.T.

    1997-01-01

    The post irradiation examinations (PIE) were carried out for high burn-up UO 2 spent fuel (3.8%U235, average burn-up:60GWd/t) and mixed oxide (MOX) spent fuel (5.07%Pu, average burn-up:45GWd/t). The PIE includes, a) isotopic analysis, b) electron probe microanalysis (EPMA) in pellet cross section and so on. The results of isotopic analyses and EPMA were compared with ORIGEN2/82 and VIM-BURN calculation results. In VIM-BURN calculation, the nuclear data of actinides were proceeded from new data file, JENDL-3.2. The sensitivities of power history and moderator density to nuclides composition were investigated by VIM-BURN calculation and consequently power history mainly effected on Am241 and Am242m and moderator density effected on fissile nuclides. From EPMA results of U and Pu distribution in pellet, VIM-BURN calculation showed reasonable distribution in pellet cross section. (author)

  10. Behaviour of (Th, U)O2 microspheres under compression tests and pelletization

    International Nuclear Information System (INIS)

    Ferreira, R.A.N.

    1982-12-01

    The interrelation between the behaviour of isolated microspheres in compression tests and the microstructure of sintered pellets obtained with these microspheres, was investigated. Various batches of (Th, 5 w/o U)O 2 microspheres were produced applying the so-called gel process. The production parameters were diversified both as to the composition and to the heat treatments. The resulting products underwent compression tests in an universal tension and compression machine as single microspheres and, as bulk material, were compacted and sintered. The results of the compression tests revealed the existence of two distinct classes of fragmentation behaviour. Each of these classes causes a distinct behaviour during the pelletization, too, resulting in fuel pellets with quite different microstructures. It was evidenced that there is a relationship between these differences in the microstructure and the behaviour of the single microspheres in the compression test. (Author) [pt

  11. Study of secular equilibrium reinstatement on UO2 pellets manufactured by AUC route

    International Nuclear Information System (INIS)

    Carnaval, João Paulo R.; Beltran, Dalton J.M.C.; Oliveira, Carlos A.

    2017-01-01

    The fuel assemblies manufactured by INB for Angra-1 power plant has axial blanket fuel rods which must be inspected due the columns formed by different enrichment pellets. The equipment used for inspection is built with a group of BGO scintillators detectors which measurement principle is based on the absorption of gamma rays emitted from Uranium decay. The commercial grade UF 6 used by INB is stored into cylinders type 30B. The uranium inside these cylinders is in secular equilibrium before the processing. It has been found that the AUC route causes the loss of that equilibrium because the UF 6 is volatilized from the cylinder and the uranium daughters remain in the container. As AUC is converted to powder and pellets, the secular equilibrium is restored through time. The purpose of this work is to present a study of the secular equilibrium reinstatement on UO 2 pellets manufactured by AUC route before being inspected on Rod Scanner. (author)

  12. Direct dissolution and supercritical fluid extraction of uranium from UO2 powder, granule, green pellet and sintered pellet

    International Nuclear Information System (INIS)

    Rao, Ankita; Kumar, Pradeep; Ramakumar, K.L.

    2009-01-01

    In the present work, direct dissolution and extraction of UO 2 from the solid rejects various stages of fuel fabrication viz. powder granules green pellet and, sintered pellet has been studied. Powder and granules could be easily dissolved in TBP-HNO 3 complex at 50 deg C., whereas in case of green and sintered pellets at elevated temperature at raised to 80 deg C in TBP-HNO 3 complex. With supercritical (SC) CO 2 alone the efficiency was ∼70%. But with SC CO 2 +2.5% TBP, the efficiency was ∼95% for powder and granules, and ∼60% for green and sintered pellets. Nearly complete extraction (∼99%) was achievable for SC CO 2 + 2.5 % TTA in all cases. The method has distinct advantage of elimination of acid usage and minimization of liquid waste generation. (author)

  13. A method of eliminating the surface defect in low-temperature oxidation powder added UO2 pellet

    International Nuclear Information System (INIS)

    Yoo, H. S.; Lee, S. J.; Kim, J. I.; Jeon, K. R.; Kim, J. W.

    2002-01-01

    A study on methods to eliminate surface defect shown in low-temperature oxidation powder added UO 2 pellet has been performed. Powders oxidized at 350 .deg. C for 4 hrs were prepared and mixed with UO 2 powder after crushing them. After being sintered, surfaces of the pellet were inspected both visually and optically. A large number of defects were observed on the surface of the specimens in which low-temperature oxidation powders were directly mixed or master mixed with UO 2 powder while both specimens produced from mixed powders including milled oxidation powders and powders that were milled totally after mixing had clean surfaces. However, optical examination showed considerably large defected pores in the milled oxidation powder added pellet and it was confirmed that the inner defects can be eliminated completely only when milling the entire mixture on UO 2 and low-temperature oxidation powder, but not by crushing only oxidation powder

  14. Modelling of high burnup structure in UO2 fuel with the RTOP code

    International Nuclear Information System (INIS)

    Likhanskii, V.; Zborovskii, V.; Evdokimov, I.; Kanyukova, V.; Sorokin, A.

    2008-01-01

    The present work deals with self-consistent physical approach aimed to derive the criterion of fuel restructuring avoiding correlations. The approach is based on study of large over pressurized bubbles formation on dislocations, at grain boundaries and in grain volume. At first, stage of formation of bubbles non-destroyable by fission fragments is examined using consistent modelling of point defects and fission gas behavior near dislocation and in grain volume. Then, evolution of formed large non-destroyable bubbles is considered using results of the previous step as initial values. Finally, condition of dislocation loops punching by sufficiently large over pressurized bubbles is regarded as the criterion of fuel restructuring onset. In the present work consideration of large over pressurized bubbles evolution is applied to modelling of the restructuring threshold depending on temperature, burnup and grain size. Effect of grain size predicted by the model is in qualitative agreement with experimental observations. Restructuring threshold criterion as an analytical function of local burnup and fuel temperature is derived and compared with HBRP project data. To predict rim-layer width formation depending on fuel burnup and irradiation conditions the model is implemented into the mechanistic fuel performance code RTOP. Calculated dependencies give upper estimate for the width of restructured region. Calculations show that one needs to consider temperature distribution within pellet which depends on irradiation history in order to model rim-structure formation

  15. Effects of MnO-Al2O3 on the grain growth and high-temperature deformation strain of UO2 fuel pellets

    International Nuclear Information System (INIS)

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

    2010-01-01

    The fabrication and high-temperature deformation strain of MnO-Al 2 O 3 -doped UO 2 pellets were studied. The effects of additive composition and amount on the microstructure evolution of a UO 2 pellet were investigated. The compressive creep behaviors of MnO-Al 2 O 3 -doped UO 2 pellets were examined. The results indicated that a MnO-Al 2 O 3 binary additive can effectively promote the grain growth of UO 2 pellets. In addition, the high-temperature deformation strain of the UO 2 pellet can be improved significantly with 1,000 ppm 95MnO-5Al 2 O 3 (mol%). The developed MnO-Al 2 O 3 -additive-containing UO 2 pellets can be a potential candidate for a high-burn-up fuel and a pellet-cladding interaction (PCI) remedy. (author)

  16. Completion of UO2 pellets production and fuel rods load for the RA-8 critical facility

    International Nuclear Information System (INIS)

    Marajofsky, Adolfo; Perez, Lidia E.; Thern, Gerardo G.; Altamirano, Jorge S.; Benitez, Ana M.; Cardenas, Hugo R.; Becerra, Fabian A.; Perez, Aldo E.; Fuente, Mariano de la

    1999-01-01

    The Advanced Fuels Division produced fuel pellets of 235 U with 1.8% and 3.6% enrichment and Zry-4 cladding loads for the RA-8 reactor at Pilcaniyeu Technological Unit. For economical and availability reasons, the powder acquired was initially UO 2 with 3.4% enrichment in 235 U, therefore the 235 U powder with 1.8% enrichment was produced by mechanical mixture. The production of fuel pellets for both enrichments was carried out by cold pressing and sintering processes in reducing atmosphere. The load of Zry-4 claddings was performed manually. The production stages can be divided into setup, qualification and production. This production allows not only to fulfill satisfactorily the new fuel rods supply for the RA-8 reactor but also to count with a new equipment and skilled personnel as well as to meet quality and assurance control methods for future pilot-scale production and even new fuel elements production. (author)

  17. The effect of dissolved hydrogen on the dissolution of 233U doped UO2(s) high burn-up spent fuel and MOX fuel

    International Nuclear Information System (INIS)

    Carbol, P.; Spahiu, K.

    2005-03-01

    In this report the results of the experimental work carried out in a large EU-research project (SFS, 2001-2004) on spent fuel stability in the presence of various amounts of near field hydrogen are presented. Studies of the dissolution of 233 U doped UO 2 (s) simulating 'old' spent fuel were carried out as static leaching tests, autoclave tests with various hydrogen concentrations and electrochemical tests. The results of the leaching behaviour of a high burn-up spent fuel pellet in 5 M NaCl solutions in the presence of 3.2 bar H 2 pressure and of MOX fuel in dilute synthetic groundwater under 53 bar H 2 pressure are also presented. In all the experimental studies carried out in this project, a considerable effect of hydrogen in the dissolution rates of radioactive materials was observed. The experimental results obtained in this project with a-doped UO 2 , high burn-up spent fuel and MOX fuel together with literature data give a reliable background to use fractional alteration/dissolution rates for spent fuel of the order of 10 -6 /yr - 10 -8 /yr with a recommended value of 4x10 -7 /yr for dissolved hydrogen concentrations above 10 -3 M and Fe(II) concentrations typical for European repository concepts. Finally, based on a review of the experimental data and available literature data, potential mechanisms of the hydrogen effect are also discussed. The work reported in this document was performed as part of the Project SFS of the European Commission 5th Framework Programme under contract no FIKW-CT-2001-20192 SFS. It represents the deliverable D10 of the experimental work package 'Key experiments using a-doped UO 2 and real spent fuel', coordinated by SKB with the participation of ITU, FZK-INE, ENRESA, CIEMAT, ARMINES-SUBATECH and SKB

  18. Fracture properties of ThO2-UO2 pellets by Hertzian indentation technique

    International Nuclear Information System (INIS)

    Kutty, T.R.G.; Rath, B.N.; Balakrishnan, K.S.

    2005-01-01

    Fracture toughness (K Ic ) and fracture surface energy (γ s ) of ThO 2 -UO 2 pellets with varying UO 2 contents were measured using Hertzian indentation technique. The knowledge of fracture toughness (K Ic ) and fracture surface energy values are important for fuel designers since these values are used in fuel modeling. Cracks in nuclear fuel act as a path for fission gas release and enhances fuel cladding mechanical interaction. Microstructural features like grain size and presence of second phase play a significant role in controlling the fracture behavior. Since the fracture properties of nuclear materials are of primary design consideration, it is important that these properties should be evaluated with good precision. There have been several attempts to use Hertzian indentation for evaluating the fracture toughness of brittle materials. The main principle of this method depends on the interaction of the elastic stress field with a pre-existing surface flaw of the sample. One significant advantage of Hertzian indentation over that of Vickers is that the substrate's deformation is entirely elastic until fracture occurs. This avoids the complications arising from the ill-defined residual stress that is normally associated with indentations brought about by pointed indenters like that of Vickers. The material properties that may be determined by this test include (a) fracture toughness and fracture surface energy of the near surface material, (b) the densities and sizes of surface cracks, and (c) residual stresses in the near surface material. This paper deals with experimental procedure for the evaluation of fracture properties of ThO 2 -UO 2 of varying U content and results thus obtained are also presented. The K Ic values thus obtained are explained in terms of their microstructures and the U content. (author)

  19. 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)

  20. An exercise to establish optimum procedures for the characterisation of porosity in UO2 fuel pellets

    International Nuclear Information System (INIS)

    Small, G.J.

    1980-05-01

    A standard metallographic preparation technique for UO 2 is proposed. The criteria for choosing the optimum route are that the specimen should be scratch-free and that the pores inherent to any sintered UO 2 pellet should be neither enlarged nor filled-in during preparation. Having met these criteria one has a specimen suitable for quantitative metallography which can be used to monitor porosity changes due to in-pile sintering. A procedure for analysing the porosity is suggested. This consists of imaging the specimen surface over a range of magnifications using both optical and Scanning Electron Microscopy in order to cover the range of pore sizes of interest (0.1 μm to 10 μm diameter). These images are then analysed to obtain figures for the distribution of pores as a function of diameter. Two methods of pore-size analysis are reviewed, the manual Zeiss Particle Size Analyser and a more sophisticated electronic instrument - the Quantimet. A comparison is made between these two instruments on the basis of accuracy, reproducibility and ease of operation. (author)

  1. Fission gas release and pellet microstructure change of high burnup BWR fuel

    International Nuclear Information System (INIS)

    Itagaki, N.; Ohira, K.; Tsuda, K.; Fischer, G.; Ota, T.

    1998-01-01

    UO 2 fuel, with and without Gadolinium, irradiated for three, five, and six irradiation cycles up to about 60 GWd/t pellet burnup in a commercial BWR were studied. The fission gas release and the rim effect were investigated by the puncture test and gas analysis method, OM (optical microscope), SEM (scanning electron microscope), and EPMA (electron probe microanalyzer). The fission gas release rate of the fuel rods irradiated up to six cycles was below a few percent; there was no tendency for the fission gas release to increase abruptly with burnup. On the other hand, microstructure changes were revealed by OM and SEM examination at the rim position with burnup increase. Fission gas was found depleted at both the rim position and the pellet center region using EPMA. There was no correlation between the fission gas release measured by the puncture test and the fission gas depletion at the rim position using EPMA. However, the depletion of fission gas in the center region had good correlation with the fission gas release rate determined by the puncture test. In addition, because the burnup is very large at the rim position of high burnup fuel and also due to the fission rate of the produced Pu, the Xe/Kr ratio at the rim position of high burnup fuel is close to the value of the fission yield of Pu. The Xe/Kr ratio determined by the gas analysis after the puncture test was equivalent to the fuel average but not to the pellet rim position. From the results, it was concluded that fission gas at the rim position was released from the UO 2 matrix in high burnup, however, most of this released fission gas was held in the porous structure and not released from the pellet to the free volume. (author)

  2. Finite element simulation of fission gas release and swelling in UO2 fuel pellets

    International Nuclear Information System (INIS)

    Denis, Alicia C.

    1999-01-01

    A fission gas release model is presented, which solves the atomic diffusion problem with xenon and krypton elements tramps produced by uranium fission during UO 2 nuclear fuel irradiation. The model considers intra and intergranular precipitation bubbles, its re dissolution owing to highly energetic fission products impact, interconnection of intergranular bubbles and gas sweeping by grain border in movement because of grain growth. In the model, the existence of a thermal gradient in the fuel pellet is considered, as well as temporal variations of fission rate owing to changes in the operation lineal power. The diffusion equation is solved by the finite element method and results of gas release and swelling calculation owing to gas fission are compared with experimental data. (author)

  3. Development of ultrasonic technique for measure of porosity of UO2 pellets

    International Nuclear Information System (INIS)

    Baroni, Douglas Brandao

    2008-01-01

    The characterization of nuclear fuel is of great importance to guarantee the efficiency and even the safety in the power stations. At present, the techniques used implicate elevated costs with equipment, materials and installations of radiological protection. Besides, because of being destructive techniques, they impose that the checking of the characteristics of this material is done by sampling. In this work a not destructive technique was developed for measures of porosity in ceramic materials with efficiency and precision. The objective of this work is to this technique will be able to be used in laboratory practice for measures in UO 2 pellets, so it would become viable the inspection of up to 100% of the nuclear fuel, guaranteeing bigger control of the characteristics of the used material, turning in increasing safety, efficiency and economy. The innovation of the technique is due to the fact of analysing the specter of frequency of the ultrasonic wrist, and not his time of course in the material, frequently used. In this work 40 ceramic pellets of alumina were used with values of porosity between 5,09% and 37,30%. A system of recognition of signs using artificial neural networks made possible to distinguish pellets with differences of porosity of 0,04%. It was observed that this technique can be used for several others aims, for example, in the determination of the void fraction in regimen of two-phase flow, what is very important to guarantee the efficiency and safety of nuclear reactors. (author)

  4. Effect of additives in sintering UO2-7wt%Gd2O3 fuel pellets

    International Nuclear Information System (INIS)

    Santos, L.R.; Riella, H.G.

    2009-01-01

    Gadolinium has been used as burnable poison for reactivity control in modern PWRs. The incorporation of Gd 2 O 3 powder directly into the UO 2 powder enables longer fuel cycles and optimized fuel utilization. Nevertheless, processing by this method leads to difficulties while obtaining sintered pellets with the minimum required density. The process for manufacturing UO 2 - Gd 2 O 3 generates scraps that should be reused. The main scraps are green and sintered pellets, which must be calcined to U 3 O 8 to return to the fabrication process. Also, the incorporation of Gd 2 O 3 in UO 2 requires the use of an additive to improve the sintering process, in order to achieve the physical properties specified for the mixed fuel, mainly density and microstructure. This paper describes the effect of the addition of fabrication scraps on the properties of the UO 2 -Gd 2 O 3 fuel. Aluminum hydroxide Al(OH) 3 was also incorporated to the fuel as a sintering aid. The results shown that the use of 2000 ppm of Al(OH) 3 as additive allow to fabricate good pellets with up to 10 wt% of recycled scraps. (author)

  5. Correlation between UO2 powder and pellet quality in PHWR fuel manufacturing

    International Nuclear Information System (INIS)

    Glodeanu, F.; Spinzi, M.; Balan, V.

    1988-01-01

    Natural uranium dioxide fuel for heavy water reactors has a series of very tightly controlled quality factors: Chemical purity, density and microstructures. Although the fabrication history may consistently affect the fuel quality, the quality factor mentioned above are function mainly of the quality of the powder used as raw material. As regards the fulfilment of the requirements for very high density of the pellets, it was found that in a definite technology the raw material plays the decisive part. Except for the powder sinterability, one found other important subtile parameters, such as the degree of agglomeration and structural homogeneity. The fuel microstructure, very important for in-serive performances of the fuel, is related to a great extent to some powder characteristics (homogeneity, sinterability). This is why much stress was laid on UO 2 power quality evaluation both by standard methods and non-conventional ones (agglomeration, microscopy, X-rays). Some of the characteristics defined by product specification, such as powder sinterability, should be better defined to guarantee the final product quality. (orig.)

  6. The radial distribution of plutonium in high burnup UO2 fuels

    International Nuclear Information System (INIS)

    Lassmann, K.; O'Carroll, C.; Laar, J. van de; Walker, C.T.

    1994-01-01

    A new model (TUBRNP) is described which predicts the radial power density distribution as a function of burnup (and hence the radial burnup profile as a function of time) together with the radial profile of uranium and plutonium isotopes. Comparisons between measurements and the predictions of the TUBRNP model are made on fuels with enrichments in the range 2.9 to 8.25% and with burnups between 21 000 and 64 000 MWd/t. It is shown to be in excellent agreement with experimental measurements and is a marked improvement on earlier versions. (orig.)

  7. Preparation of high density (Th, U)O2 pellets by sol-gel microsphere pelletization and 1300 C air sintering

    International Nuclear Information System (INIS)

    Yamagishi, Shigeru; Takahashi, Yoshihisa

    1994-01-01

    The fabrication of high density (Th, U)O 2 pellets by the sol-gel microsphere pelletization (SGMP) process was studied. To prepare source ThO 2 -UO 3 microspheres, isopropyl alcohol was substituted for the water in gel and thereafter removed by evacuating and subsequently by heating at 200 C in air. After humidifying the microspheres up to the moisture content ranging 10-21%, they were compacted into a pellet under 150-500 MPa and sintered in air at 1300 C. Even at the relatively low temperature, the maximum density reached 98% TD or higher for the U/(Th+U) ratios of 5-20 mol%. Such high density products survived as firm pellets with a similarly high density of 99% TD during the reduction into (Th, U)O 2 in Ar-4% H 2 at 1300 C. ((orig.))

  8. Manufacturing at industrial level of UO2 pellets for the fuel elements of the Atucha I Nuclear Power Plant

    International Nuclear Information System (INIS)

    Dyment, I.G.; Noguera Rojas, Francisco

    1982-01-01

    The interest to produce fuel elements within a policy of self sufficiency arose with the installation of Atucha I. The first steps towards this goal consisted in processing the uranium oxide, transforming it into fuel pellets of high density. The developments towards the fabrication of said pellets, performed by CNEA since 1968, first at a laboratory level and afterwards on an industrial scale, allowed CNEA to obtain its own technological capability to produce 400 kg of UO 2 per day. The fuel pellets manufacturing method developed by CNEA is a powder-metallurgical process, which, besides conventional equipment, involves the use of special equipment that required the performance of systematic testing programmes, as well as special training at operational level. The developed processes respond to a modern and advanced technology. A general scheme of the process, starting with a directly sinterable UO 2 powder, is described, including compacting of the powder into pellets, sintering, control of the temperature in the sintering and reduction zones and of the time of permanence in both zones, and cylindric rectifying of the pellets. During the whole process, specialized personnel controls the operations, after which the material is released by the Quality Control Department. The national contribution to the manufacturing technology of the pellets for fuel elements of power and research reactors was of 100%. (M.E.L.) [es

  9. Dose rate measurements in the beta-photon radiation field from UO2 pellets and glazed ceramics containing uranium

    International Nuclear Information System (INIS)

    Piesch, E.; Burgkhardt, B.

    1986-01-01

    In the nuclear fuel cycle, the handling of UO 2 pellets results in a significant exposure, mainly due to beta rays. Depth dose distributions have been investigated at source-to-detector distances of 5 to 80 cm using LiF detectors of different thicknesses. Detailed data for the dose equivalent quantities H(0.07), H(3) and H(10) are presented. These data are compared with those found for the use of glazed tiles and ceramics containing natural uranium. (author)

  10. 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)

  11. Influence of process parameters on the fabrication of UO2-PuO2 pellets using the granulation technique

    International Nuclear Information System (INIS)

    Vollath, D.; Wedemeyer, H.

    1982-01-01

    The preparation of UO 2 -PuO 2 fuel pellets from ground and granulated powders results in the formation of mixed oxide solid solutions which is decisive for the solubility of the fuel. Compared with granulated powders made from sintered and ground material, the compaction of green powders leads to much lower compaction densities at the same compaction pressure. Discontinuities of the porosity of sintered pellets made from long-time ground powders probably reflect the rupture of the granulate structure during compaction. Unusual high values of contact numbers obtained by gas effusion measurements indicate a network of cracks in the sintered material. (orig.)

  12. On possible mechanisms of rim-layer formation in the high-burnup UO2 fuel

    International Nuclear Information System (INIS)

    Zborovskii, V.; Likhanskii, V.

    2006-01-01

    Two models determining threshold conditions for onset of UO 2 fuel restructuring are developed. In the first model the conditions for fuel restructuring are related with development of the Kinoshita instability. The second model is based upon attainment of critical values by radius of over pressurised bubbles. Possibility of large bubbles formation on dislocation lines is considered with account of Xe atoms drift in the field of mechanical strain of dislocation and irradiation-induced Xe drift in vacancy concentration gradient. Computer simulations of behaviour of point defects and Xe atoms near dislocation core are carried out, results are compared with experimental data. The computer program is developed which consistently calculates point defects and Xe atoms distributions inside fuel grain with account of their behaviour near dislocation core

  13. Performance of Bruce natural UO2 fuel irradiated to extended burnups

    International Nuclear Information System (INIS)

    Zhou, Y.N.; Floyd, M.R.; Ryz, M.A.

    1995-11-01

    Bruce-type bundles XY, AAH and GF were successfully irradiated in the NRU reactor at Chalk River Laboratories to outer-element burnups of 570-900 MWh/kgU. These bundles were of the Bruce Nuclear Generating Station (NGS)-A 'first-charge' design that contained gas plenums in the outer elements. The maximum outer-element linear powers were 33-37 kW/m. Post-irradiation examination of these bundles confirmed that all the elements were intact. Bundles XY and AAH, irradiated to outer-element burnups of 570-700 MWh/kgU, experienced low fission-gas release (FGR) ( 500 MWh/kgU (equivalent to bundle-average 450 MWh/kgU) when maximum outer-element linear powers are > 50 kW/m. The analysis in this paper suggests that CANDU 37-element fuel can be successfully irradiated (low-FGR/defect-free) to burnups of at least 700 MWh/kgU, provided maximum power do not exceed 40 kW/m. (author). 5 refs., 1 tab., 8 figs

  14. Study of scrap recovery for mixed oxide pellet by means of UO2+5wt.%CeO2

    International Nuclear Information System (INIS)

    Joung, Chang Young; Kim, Si Hyung; Kim, Han Soo; Lee, Yong Woo

    2000-01-01

    The recovery method of scrap powder was established using UO 2 -5wt.%CeO 2 powder in the way of of technology development of scrap recovery and recycle in the mixed oxide pellet fabrication process. And pellet density and microstructure, which depend on quantity of scrap, powder treatment method, and sintering condition, was analyzed. As a result of oxidation of sintered pellet in the air the powdering at below 400 degrees C occurred smoothly, and at above 400 degrees C powdering process was not proceeding well as temperature increased and powder particle size grew bigger. M 3 O 8 scrap powder which was powdered through oxidation method was added to UO 2 -5wt.%CeO 2 mixed powder. The results after the powder was treated by means of mixing, crushing, attrition milling, pelletizing, and sintering showed that its density and grain size in the case of reductive sintering decreased as scrap addition increased, but the result in the case of crushing showed increase in grain size. In attrition milling case both density and grain size showed the tendency of increase, particularly grain size grew up to 12 μm. In the oxidative sintering with scrap added mixed powder sintering was accelerated under the oxygen environment and the effect of powder treatment showed the tendency of relatively decreasing. (Hong, J. S.)

  15. Simulation of the neutron-physical properties of the classical UO2 fuel and of MOX fuel during the burn-up by Transuranus

    International Nuclear Information System (INIS)

    Breza, J. jr.; Necas, V.; Daoeilek, P.

    2005-01-01

    The classical nuclear fuel UO 2 is well known for VVER reactors. Nevertheless, in the near future it will be possible to replace this fuel by novel, advanced kinds of fuel, for instance MOX, inert matrices fuel, etc., that will allow to increase the level of burn-up and minimize the amount of hazardous waste. The code Transuranus [2], designed at ITU Karlsruhe, is intended for thermal and mechanical analyses of fuel elements in nuclear reactors. We have utilized the code Transuranus to simulate the neutron-physical properties of the classical UO 2 fuel and of MOX fuel during the burn-up to a level of 40 MWd/kgHM. We compare obtained results of uranium and plutonium nuclides concentrations, their changes during burn-up, with results obtained by code HELIOS [3], which is well-validated code for this kind of applications. We performed calculations of fission gasses concentrations, namely xenon and krypton. (author)

  16. 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

  17. 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)

  18. Irradiation of UO2

    International Nuclear Information System (INIS)

    Stevanovic, M.

    1965-10-01

    Based on the review of the available literature concerned with UO 2 irradiation, this paper describes and explains the phenomena initiated by irradiation of the UO 2 fuel in a reactor dependent on the burnup level and temperature. A comprehensive review of UO 2 radiation damage studies is given as a broad research program. This part includes the abilities of our reactor as well as needed elements for such study. The third part includes the definitions of the specific power, burnup level and temperature in the center of the fuel element needed for planning and performing the irradiation. Methods for calculating these parameters are included [sr

  19. Modeling the UO2 ex-AUC pellet process and predicting the fuel rod temperature distribution under steady-state operating condition

    Science.gov (United States)

    Hung, Nguyen Trong; Thuan, Le Ba; Thanh, Tran Chi; Nhuan, Hoang; Khoai, Do Van; Tung, Nguyen Van; Lee, Jin-Young; Jyothi, Rajesh Kumar

    2018-06-01

    Modeling uranium dioxide pellet process from ammonium uranyl carbonate - derived uranium dioxide powder (UO2 ex-AUC powder) and predicting fuel rod temperature distribution were reported in the paper. Response surface methodology (RSM) and FRAPCON-4.0 code were used to model the process and to predict the fuel rod temperature under steady-state operating condition. Fuel rod design of AP-1000 designed by Westinghouse Electric Corporation, in these the pellet fabrication parameters are from the study, were input data for the code. The predictive data were suggested the relationship between the fabrication parameters of UO2 pellets and their temperature image in nuclear reactor.

  20. 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

  1. Sintering of uranium dioxide pellets (UO2) in an oxidizing atmosphere (C O2)

    International Nuclear Information System (INIS)

    Santos, G.R.T.

    1992-01-01

    This work consists in the study of the sintering process of U O 2 pellets in an oxidizing atmosphere. Sintering tests were performed in an CO 2 atmosphere and the influence of temperature and time on the pellets density and microstructure were verified. The results obtained were compared to those from the conventional sintering process and its efficiency was confirmed. (author)

  2. Effect of technological parameters and microstructure on mechanical strength of UO2 fuel pellets

    International Nuclear Information System (INIS)

    Radford, K.

    1980-01-01

    The effect of various peculiarities of tablet microstructure namely, sammury porosity (tablet density), grain size and pore distribution over sizes on technological parameters, is studied. It is shown that density decrease leads to a fast reduction of UO 2 tablet strength. The maximum effect on strength is produced by pore distribution over sizes, characterized by a median size, and not by the grain size, though a combined effect of those two factors is also observed. The important role of the technology of tablet production manifests itself in the fact that all operations bringing about the increase of pore or grain sizes leads to a reduction of strength. Such factors as powder origin, granule sizes, U 3 O 8 content and the amount of additions do not cause any considerable changes in the strength of tablets. Bend tests under conditions of biaxial loading should be considered as an ideal method of determining fuel tablets strength [ru

  3. Spectrophotometric method for the determination of thorium in UO2 pellets

    International Nuclear Information System (INIS)

    Acosta L, E.

    1995-04-01

    This report presents a procedure with the description of the spectrophotometric method for the determination of the thorium element in uranium products including powders and pellets of uranium dioxide. Quantities can be determined starting from 1 ppm. (Author)

  4. UO2 Fuel pellet impurities, pellet surface roughness and n(18O)/n(16O) ratios, applied to nuclear forensic science

    International Nuclear Information System (INIS)

    Pajo, L.

    2001-01-01

    In the last decade, law enforcement has faced the problem of illicit trafficking of nuclear materials. Nuclear forensic science is a new branch of science that enables the identification of seized nuclear material. The identification is not based on a fixed scheme, but further identification parameters are decided based on previous identification results. The analysis is carried out by using traditional analysis methods and applying modern measurement technology. The parameters are generally not unambiguous and not self-explanatory. In order to have a full picture about the origin of seized samples, several identification parameters should be used together and the measured data should be compared to corresponding data from known sources. A nuclear material database containing data from several fabrication plants is installed for the purpose. In this thesis the use of UO 2 fabrication plant specific parameters, fuel impurities, fuel pellet surface roughness and oxygen isotopic ratio in UO 2 were investigated for identification purposes in nuclear forensic science. The potential use of these parameters as 'fingerprints' is discussed for identification purposes of seized nuclear materials. Impurities of the fuel material vary slightly according to the fabrication method employed and a plant environment. Here the impurities of the seized UO 2 were used in order to have some clues about the origin of the fuel material by comparing a measured data to nuclear database information. More certainty in the identification was gained by surface roughness of the UO 2 fuel pellets, measured by mechanical surface profilometry. Categories in surface roughness between a different fuel element type and a producer were observed. For the time oxygen isotopic ratios were determined by Thermal Ionisation Mass Speckometry (TIMS). Thus a TIMS measurement method, using U 16 O + and U 18 0 + ions, was developed and optimised to achieve precise oxygen isotope ratio measurements for the

  5. 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

  6. 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

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  8. Selective alpha autoradiography for monitoring thorium distribution in UO2-ThO2 fuel pellets

    International Nuclear Information System (INIS)

    Shriwastwa, B.B.; Raghunath, B.; Ghosh, J.K.

    1992-01-01

    Although natural uranium and thorium decay with similar alpha energies (4.20 and 3.98 MeV), their daughter products have different alpha characteristics. This has been exploited for selective alpha autoradiography for thoria in urania-thoria mixed nuclear fuel pellets. Difficulties in getting sufficient track density in alpha sensitive films due to the very low specific activity of natural uranium and thorium material were overcome by using a special film with annealing and pre-etching treatment. (orig./HP) [de

  9. 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

  10. On the thermal conductivity of UO2 nuclear fuel at a high burn-up of around 100 MWd/kgHM

    International Nuclear Information System (INIS)

    Walker, C.T.; Staicu, D.; Sheindlin, M.; Papaioannou, D.; Goll, W.; Sontheimer, F.

    2006-01-01

    A study of the thermal conductivity of a commercial PWR fuel with an average pellet burn-up of 102 MWd/kgHM is described. The thermal conductivity data reported were derived from the thermal diffusivity measured by the laser flash method. The factors determining the fuel thermal conductivity at high burn-up were elucidated by investigating the recovery that occurred during thermal annealing. It was found that the thermal conductivity in the outer region of the fuel was much higher than it would have been if the high burn-up structure were not present. The increase in thermal conductivity is a consequence of the removal of fission products and radiation defects from the fuel lattice during recrystallisation of the fuel grains (an integral part of the formation process of the high burn-up structure). The gas porosity in the high burn-up structure lowers the increase in thermal conductivity caused by recrystallisation

  11. Long-term safety of radioactive waste disposal: Chemical reaction of fabricated and high burnup spent UO2 fuel with saline brines. Final report

    International Nuclear Information System (INIS)

    Grambow, B.; Casas, I.; Pablo, J. de; Gimenez, J.; Torrero, M.E.

    1996-03-01

    This is the final report of a large EU-research project on spent fuel stability in saline repository environments. Static dissolution experiments with high burnup spent fuel samples and unirradiated UO 2 were performed for about two years in anaerobic NaCl solutions and deionized water with and without container material (iron) being present. Experiments performed at 25 and 150 C gave similar results. Dissolution rates were similar to those measured in the Swedish, or Canadian program for granite media. Rates are strongly influenced by the specific sample surface area, probably related to the mass balance of consumption and production of radiolytic oxidants. In the competition between the oxidizing effect of radiolysis and the reducing effect of iron, the metal corrosion process dominates. Processes controlling radionuclide release are matrix dissolution, solubility, coprecipitation sorption phenomena and colloid formation. In the absence of iron release rates of Sr90, Tc99, Np237, Sb125 and at low reaction progress Ru106 were controlled by matrix dissolution whereas concentrations of tetra-, hexa-, and trivalent actinides (U, Pu, Am, Cm) were controlled by solubility or coprecipitation. The presence of iron did effectively reduce the rates of fuel dissolution and the concentration of many, though not all radionuclides. Solubilities of U were similar for uniradiated UO 2 and for spent fuel both in the case of oxidizing and reducing conditions. In contrast, due to the effect of radiolysis, reaction rates of spent fuel were higher than UO 2 dissolution rates. (orig.) [de

  12. Review on quality control techniques of UO2 pellets under pilot-plant conditions, at Instituto de Energia Atomica, Sao Paulo, Brazil

    International Nuclear Information System (INIS)

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

    1977-01-01

    The Instituto de Energia Atomica's Metallurgy Division Pilot Plant has been established to develop fabrication and control techniques of ceramic fuel elements, to train personnel and to acquire experience in quality control of fuel pellets. Its close association with the Institute's Chemical Engineering Division, where pilot-plant development on uranium and thorium purification is carried out, affords a direct way to ascertain the influence of salt processing variables on the behaviour of oxides derived from such uranium salts (ammonium diuranate and ammonium uranyl carbonate). The pilot plant, with a capacity of about 5 tons of UO 2 pellets per year, has ample flexibility in equipment, installations and procedures for such work, comprising uranium salts calcining, UO 2 reduction, UO 2 pellet fabrication, sintering, inspection, centerless grinding and adequate controls, both on powders and on pellets produced. It comprises several self-contained sub-units, corresponding to each particular operation, arranged in such a way that work can be carried independently and asssuring in each good control of accountablity.Quality control techniques are exerted both on powder and on pellets lines. In the powders line, besides the current routine control tests, special ones have been developed and used, comprising grain size microscopy, electron scanning examination of particle shape and sedimentation tests. These controls allow fabrication of oxide powders (mainly natural uranium up to now) to meet the specifications for the particular programs that have been tackled. In the pellets line, with ample flexibility on fabrication steps, both low- (90 to 93 pct density) and medium-density pellets (93 to 95 pct) are produced. Besides the usal routine controls, special tests on quantitative pore and grain sizes distribution through quantitative optical microscopy, electron scanning microscopy and fractographic tests were developed to evaluate the homogeneity and the geometry of pore

  13. Advances in fuel pellet technology for improved performance at high burnup. Proceedings of a Technical Committee meeting

    International Nuclear Information System (INIS)

    1998-08-01

    The IAEA has recently completed two co-ordinated Research Programmes (CRPs) on The Development of Computer Models for Fuel Element Behaviour in Water Reactors, and on Fuel Modelling at Extended Burnup. Through these CRPs it became evident that there was a need to obtain data on fuel behaviour at high burnup. Data related o thermal behaviour, fission gas release and pellet to clad mechanical interaction were obtained and presented at the Technical Committee Meeting on Advances in Fuel Pellet Technology for Improved Performance at High Burnup which was recommended by the International Working Group on Fuel Performance and Technology (IWGFPT). The 34 papers from 10 countries are published in this proceedings and presented by a separate abstract. The papers were grouped in 6 sessions. First two sessions covered the fabrication of both UO 2 fuel and additives and MOX fuel. Sessions 3 and 4 covered the thermal behaviour of both types of fuel. The remaining two sessions dealt with fission gas release and the mechanical aspects of pellet to clad interaction

  14. Pu-rich MOX agglomerate-by-agglomerate model for fuel pellet burnup analysis

    International Nuclear Information System (INIS)

    Chang, G.S.

    2004-01-01

    In support of potential licensing of the mixed oxide (MOX) fuel made from weapons-grade (WG) plutonium and depleted uranium for use in United States reactors, an experiment containing WG-MOX fuel is being irradiated in the Advanced Test Reactor (ATR) at the Idaho National Engineering and Environmental Laboratory (INEEL). The WG-MOX comprises five percent PuO 2 and 95% depleted UO 2 . Based on the Post Irradiation Examination (PIE) observation, the volume fraction (VF) of MOX agglomerates in the fuel pellet is about 16.67%, and PuO 2 concentration of 30.0 = (5 / 16.67 x 100) wt% in the agglomerate. A pressurized water reactor (PWR) unit WG-MOX lattice with Agglomerate-by-Agglomerate Fuel (AbAF) modeling has been developed. The effect of the irregular agglomerate distribution can be addressed through the use of the Monte Carlo AbAF model. The AbAF-calculated cumulative ratio of Agglomerate burnup to U-MAtrix burnup (AG/MA) is 9.17 at the beginning of life, and decreases to 2.88 at 50 GWd/t. The MCNP-AbAF-calculated results can be used to adjust the parameters in the MOX fuel fission gas release modeling. (author)

  15. Fabrication, irradiation and post-irradiation examinations of MO2 and UO2 sphere-pac and UO2 pellet fuel pins irradiated in a PWR loop

    International Nuclear Information System (INIS)

    Linde, A. van der; Lucas Luijckx, H.J.B.; Verheugen, J.H.N.

    1982-01-01

    The document reports in detail the fuel pin fabrication data and describes the irradiation conditions and history. All the relevant results of the non-destructive and destructive post-irradiation examinations are reported. They include: visual inspection and chemical analysis of crud; length and diameter measurements; neutron radiography and gamma scanning; juncture tests and fission gas analysis (including residual gas in fuel samples); microscopy and alpha + beta/gamma autoradiography; microprobe investigations; burn-up and isotopic analysis; and hydrogen analysis in clad. The data and observations obtained are discussed in detail and conclusions are given. The irradiation and post-irradiation examinations of the R-109 pins have shown the safe, pre-calculable performance of LWR fuel pins containing mixed-oxide sphere-pac fuel with the fissile material mainly present in the large spheres

  16. Burn-up credit criticality safety benchmark phase VII - UO2 fuel: study of spent fuel compositions for long-term disposal

    International Nuclear Information System (INIS)

    2012-01-01

    After spent nuclear fuel (SNF) is discharged from a nuclear reactor, fuel composition and reactivity continue to vary as a function of time due to the decay of unstable nuclides. Accurate predictions of the concentrations of long-lived radionuclides in SNF, which represent a significant potential hazard to human beings and to the environment over a very long period, are particularly necessary for radiological dose assessments. This report assesses the ability of existing computer codes and associated nuclear data to predict isotopic compositions and their corresponding neutron multiplication factor (k eff ) values for pressurised-water-reactor (PWR) UO 2 fuel at 50 GWd/MTU burn-up in a generic spent fuel cask configuration. Fuel decay compositions and k eff values have been calculated for 30 post-irradiation time steps out to one million years

  17. Determination of oxygen to metal ratio for varying UO2 content in sintered (U,Th)O2 pellet by oxidation-reduction method using thermo-gravimetric analyser

    International Nuclear Information System (INIS)

    Mahanty, B.N.; Khan, F.A.; Karande, A.P.; Prakash, A.; Afzal, Md.; Panakkal, J.P.

    2009-01-01

    Experiments were carried out to determine oxygen to metal ratio in 4%, 6%, 10%, 20%, 50% and 80% UO 2 in sintered (U, Th) O 2 pellets by oxidation-reduction method using thermo gravimetric analyser. (author)

  18. Critical assessment of the pore size distribution in the rim region of high burnup UO_2 fuels

    International Nuclear Information System (INIS)

    Cappia, F.; Pizzocri, D.; Schubert, A.; Van Uffelen, P.; Paperini, G.; Pellottiero, D.; Macián-Juan, R.; Rondinella, V.V.

    2016-01-01

    A new methodology is introduced to analyse porosity data in the high burnup structure. Image analysis is coupled with the adaptive kernel density estimator to obtain a detailed characterisation of the pore size distribution, without a-priori assumption on the functional form of the distribution. Subsequently, stereological analysis is carried out. The method shows advantages compared to the classical approach based on the histogram in terms of detail in the description and accuracy within the experimental limits. Results are compared to the approximation of a log-normal distribution. In the investigated local burnup range (80–200 GWd/tHM), the agreement of the two approaches is satisfactory. From the obtained total pore density and mean pore diameter as a function of local burnup, pore coarsening is observed starting from ≈100 GWd/tHM, in agreement with a previous investigation. - Highlights: • A new methodology to analyse porosity is introduced. • The method shows advantages compared to the histogram. • Pore density and mean diameter data vs. burnup are presented. • Pore coarsening is observed starting from ≈100 GWd/tHM.

  19. Surface analysis using X-ray photoelectron spectroscopy and X-ray diffraction of UO2 fuel pellets oxidised in air at 2300C and 2700C

    International Nuclear Information System (INIS)

    Tempest, P.A.; Tyler, J.W.

    1987-08-01

    Factors which affect the UO 2 → U 3 O 8 transformation have been investigated by sequentially oxidising UO 2 fuel pellets in air at 230 0 C and 270 0 C and monitoring the growth of U 3 O 7 and U 3 O 8 using X-ray photoelectron spectroscopy, X-ray diffraction and scanning electron microscopy. Initially oxidation proceeded at a linear rate by the inward diffusion of oxygen to form a complete layer of sub-stoichiometric U 3 O 7 . This phase was tetragonal with a c/a ratio of 1.015, significantly less than the value of 1.03 measured on UO 2 powder when oxidised under identical conditions. This difference and the preferred orientation exhibited by surface grains were caused by growth stresses induced in the pellet surface. Both intergranular and transgranular cracking occurred and became nucleation sites for the growth of U 3 O 8 . The linear oxidation period associated with U 3 O 7 growth was much shorter at 270 0 C than 230 0 C and U 3 O 8 nucleated earlier. Spallation and the production of particulate were only observed during the formation of U 3 O 8 when a 30% increase in volume arose from the U 3 O 7 → U 3 O 8 phase change. (author)

  20. Quality control and testing UO2 powder and sintering pellets for nuclear fuel for LWR in out of pile condition

    International Nuclear Information System (INIS)

    Djuricic, Lj.; Katanic, J.; Stefanovic, M.

    1976-01-01

    The analysis of chemical and physical characteristics of fuels based on UO2 from the point of view of requested properties in the nuclear application, of the foreign technical methods of characterisation and domestic experience is given as one of the first steps toward standardization in the field in the state

  1. 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)

  2. An evaluation of UO2-CNT composites made by SPS as an accident tolerant nuclear fuel pellet and the feasibility of SPS as an economical fabrication process for the nuclear fuel cycle

    Science.gov (United States)

    Cartas, Andrew R.

    The innovative and advanced purpose of this study is to understand and establish proper sintering procedures for Spark Plasma Sintering process in order to fabricate high density, high thermal conductivity UO2 -CNT pellets. Mixing quality and chemical reactions have been investigated by field emission scanning electron microscopy (FESEM), wavelength dispersive spectroscopy (WDS), and X-ray diffraction (XRD). The effect of various types of CNTs on the mixing and sintering quality of UO2-CNT pellets with SPS processing have been examined. The Archimedes Immersion Method, laser flash method, and FE-SEM will be used to investigate the density, thermal conductivity, grain size, pinning effects, and CNT dispersion of fabricated UO2-CNT pellets. Pre-fabricated CNT's were added to UO 2 powder and dispersed via sonication and/or ball milling and then made into composite nuclear pellets. An investigation of the economic impact of SPS on the nuclear fuel cycle for producing pure and composite UO2 fuels was conducted.

  3. Development of thermocouple re-instrumentation technique for irradiated fuel rod. Techniques for making center hole into UO2 pellets and thermocouple re-instrumentation to fuel rod

    International Nuclear Information System (INIS)

    Shimizu, Michio; Saito, Junichi; Oshima, Kunio

    1995-07-01

    The information on FP gas pressure and centerline temperature of fuel pellets during power transient is important to study the pellet clad interaction (PCI) mechanism of high burnup LWR fuel rods. At the Department of JMTR, a re-instrumentation technique of FP gas pressure gage for an irradiated fuel rod was developed in 1990. Furthermore, a thermocouple re-instrumentation technique was successfully developed in 1994. Two steps were taken to carry out the development program of the thermocouple re-instrumentation technique. In the first step, a drilling technique was developed for making a center hole of the irradiated fuel pellets. Various drilling tests were carried out using dummy of fuel rods consisted of Ba 2 FeO 3 pellets and Zry-2 cladding. On this work it is important to keep the pellets just the state cracked at a power reactor. In these tests, the technique to fix the pellets by frozen CO 2 was used during the drilling work. Also, diamond drills were used to make the center hole. These tests were completed successfully. A center hole, 54mm depth and 2.5mm diameter, was realized by these methods. The second step of this program is the in-pile demonstration test on an irradiated fuel rod instrumented dually a thermocouple and FP gas pressure gage. The demonstration test was carried out at the JMTR in 1995. (author)

  4. Argentina-LLNL-LANL Comparative Sample Analysis on UO2 fuel pellet CRM-125A for Nuclear Forensics

    Energy Technology Data Exchange (ETDEWEB)

    Kips, R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2017-12-01

    The recent workshop on analytical plan development provided context and background for the next step in this engagement, i.e. a comparative sample analysis on CRM 125-A. This is a commercially available certified low-enriched uranium oxide fuel pellet material from New Brunswick National Laboratory (NBL) (see certificate in Annex 1).

  5. Burn-up Credit Criticality Safety Benchmark-Phase II-E. Impact of Isotopic Inventory Changes due to Control Rod Insertions on Reactivity and the End Effect in PWR UO2 Fuel Assemblies

    International Nuclear Information System (INIS)

    Neuber, Jens Christian; Tippl, Wolfgang; Hemptinne, Gwendoline de; Maes, Philippe; Ranta-aho, Anssu; Peneliau, Yannick; Jutier, Ludyvine; Tardy, Marcel; Reiche, Ingo; Kroeger, Helge; Nakata, Tetsuo; Armishaw, Malcom; Miller, Thomas M.

    2015-01-01

    The report describes the final results of the Phase II-E Burn-up Credit Criticality Benchmark conducted by the Expert Group on Burn-up Credit Criticality Safety. The objective of Phase II of the Burn-up Credit Criticality Safety programme is to study the impact of axial burn-up profiles of PWR UO 2 spent fuel assemblies on the reactivity of PWR UO 2 spent fuel assembly configurations. The objective of the Phase II-E benchmark was to study the impact of changes on the spent nuclear fuel isotopic composition due to control rod insertion during depletion on the reactivity and the end effect of spent fuel assemblies with realistic axial burn-up profiles for different control rod insertion depths ranging from 0 cm (no insertion) to full insertion (i.e. to the case that the fuel assemblies were exposed to control rod insertion over their full active length). For this purpose two axial burn-up profiles have been extracted from an AREVA-NP-GmbH-owned 17x17-(24+1) PWR UO 2 spent fuel assembly burn-up profile database. One profile has an average burn-up of 30 MWd/kg U, the other profile is related to an average burn-up of 50 MWd/kg U. Two profiles with different average burn-up values were selected because the shape of the burn-up profile is affected by the average burn-up and the end effect depends on the average burn-up of the fuel. The Phase II-E benchmark exercise complements the Phase II-C and Phase II-D benchmark exercises. In Phase II-D different irradiation histories were analysed using different control rod insertion histories during depletion as well as irradiation histories without control rod insertion. But in all the histories analysed a uniform distribution of the burn-up and hence a uniform distribution of the isotopic composition were assumed; and in all the histories including any usage of control rods full insertion of the control rods was assumed. In Phase II-C the impact of the asymmetry of axial burn-up profiles on the reactivity and the end effect of

  6. Neutronics characteristics of micro-heterogeneous ThO2-UO2 PWR cores

    International Nuclear Information System (INIS)

    Zhao, X.; Driscoll, M.J.; Kazimi, S.

    2001-01-01

    A new fuel concept, axially-micro-heterogeneous ThO 2 -UO 2 fuel, where ThO 2 fuel pellets and UO 2 fuel pellets are stacked in separate layers in the fuel rods, is being studied at MIT as an option to reduce plutonium production in LWR fuel. Very interesting neutronic behavior is observed: (1) A reactivity increase of 3% to 4% at EOL for a given 235 U inventory which results in a 20-30% increase in average core discharge burnup; (2) For certain configurations, a ''burnable poison'' effect is observed. Analysis shows that these effects are achieved due to a combination of changes in self-shielding, local fissile worth, and conversion ratio, among which self-shielding is the dominant effect at the end of a reactivity-limited burnup. Other variations of micro-heterogeneous UO 2 -ThO 2 fuel including duplex pellets, checkerboard pin distribution, and checkerboard-axial combinations have also been investigated, and their neutronic performance compared. It is concluded that the axial fuel micro-heterogeneity provides the largest gain in reactivity-limited burnup. (author)

  7. Determination of nitrogen in UO2 pellets by hot extraction-TCD method without using flux material and evaluation of associated measurement uncertainties

    International Nuclear Information System (INIS)

    Begum, Zahida; Balaji Rao, Y.; Subba Rao, Y.

    2015-01-01

    Presence of non metallic impurity elements like nitrogen above the specified limit of 75 ppm (max.) in fuel pellets affects the sintered density of fuel and also effect the Zircaloy fuel clad by forming brittle nitride phase. Hence, estimation of nitrogen plays an important role in qualifying the fuel material. Conventionally, Kjeldahl steam distillation followed by UV-Visible Spectro-photometric method has been widely employed for the estimation of nitrogen in UO 2 pellets. However, inherent chemical treatment processes causes the blank variations and provides the scope for uncertainty in measurements apart from being time consuming method. This makes Kjeldahl method as an un-attractive choice for any industrial lab where high analytical loads usually exist to meet the production targets and also where urgency for quick analytical feedback is an issue. In view of this, a simple, rapid and reliable method using A LECO Model TN- 600 Nitrogen analyzer based on Hot Extraction (HT) without adding any flux material followed by Thermal Conductivity Detection (TCD) has been developed for routine analysis. Several flux materials like Pt or Ni are used to bring down the temperature of extraction due to formation of stable uranium nitrides and temperatures of 2500°C is maintained for quantitative extraction. However usage of flux materials leads to several practical problems in analysis like deposition of uranium carbide dust onto the upper electrode requiring tedious cleaning and even sometimes causing the breakage of crucibles during the analysis. Thus an attempt has been made to increase the temperatures closer to 3000°C without using flux. Working temperature of 2900°C was arrived by plotting quantity of nitrogen extracted with temperature of extraction and flattening of plot took place at 2900°C

  8. Technical Issues in the development of high burnup and long cycle fuel pellets

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-10-15

    Over the last half century, a nuclear fuel cycle, a fuel discharged burnup and a uranium enrichment of the LWR (Light Water Reactor) fuel have continuously increased. It was the efforts to reduce the LWR fuel cycle cost, and to make reactor operation more efficiently. Improved fuel and reactor performance contribute further to the reduction and management efficiency of spent fuels. The primary incentive for operating nuclear reactor fuel to higher burnup and longer cycle is the economic benefits. The fuel cycle costs could be reduced by extending fuel discharged burnup and fuel cycle length. The higher discharged burnup can increase the energy production per unit fuel mass or fuel assembly. The longer fuel cycle can increase reactor operation flexibility and reduce the fuel changing operation and the spent fuel management burden. The margin to storage capacity limits would be also increased because high burnup and long cycle fuel reduces the mass of spent fuels. However, increment of fuel burnup and cycle length might result in the acceleration of material aging consisting fuel assembly. Then, the safety and integrity of nuclear fuel will be degraded. Therefore, to simultaneously enhance the safety and economics of the LWR fuel through the fuel burnup and cycle extension, it is indispensable to develop the innovative nuclear fuel material concepts and technologies which can overcome degradation of fuel safety. New fuel research project to extend fuel discharged burnup and cycle length has been launched in KAERI. Main subject is to develop innovative LWR fuel pellets which can provide required fuel performance and safety at extended fuel burnup and cycle length. In order to achieve the mission, we need to know that what the impediments are and how to break through current limit of fuel pellet properties. In this study, the technical issues related to fuel pellets at high burnup were surveyed and summarized. We have collected the technical issues in the literatures

  9. Technical Issues in the development of high burnup and long cycle fuel pellets

    International Nuclear Information System (INIS)

    Kim, Dong Joo; Yang, Jae Ho; Oh, Jang Soo; Kim, Keon Sik; Rhee, Young Woo; Kim, Jong Hun; Nam, Ik Hui

    2012-01-01

    Over the last half century, a nuclear fuel cycle, a fuel discharged burnup and a uranium enrichment of the LWR (Light Water Reactor) fuel have continuously increased. It was the efforts to reduce the LWR fuel cycle cost, and to make reactor operation more efficiently. Improved fuel and reactor performance contribute further to the reduction and management efficiency of spent fuels. The primary incentive for operating nuclear reactor fuel to higher burnup and longer cycle is the economic benefits. The fuel cycle costs could be reduced by extending fuel discharged burnup and fuel cycle length. The higher discharged burnup can increase the energy production per unit fuel mass or fuel assembly. The longer fuel cycle can increase reactor operation flexibility and reduce the fuel changing operation and the spent fuel management burden. The margin to storage capacity limits would be also increased because high burnup and long cycle fuel reduces the mass of spent fuels. However, increment of fuel burnup and cycle length might result in the acceleration of material aging consisting fuel assembly. Then, the safety and integrity of nuclear fuel will be degraded. Therefore, to simultaneously enhance the safety and economics of the LWR fuel through the fuel burnup and cycle extension, it is indispensable to develop the innovative nuclear fuel material concepts and technologies which can overcome degradation of fuel safety. New fuel research project to extend fuel discharged burnup and cycle length has been launched in KAERI. Main subject is to develop innovative LWR fuel pellets which can provide required fuel performance and safety at extended fuel burnup and cycle length. In order to achieve the mission, we need to know that what the impediments are and how to break through current limit of fuel pellet properties. In this study, the technical issues related to fuel pellets at high burnup were surveyed and summarized. We have collected the technical issues in the literatures

  10. An improved UO2 thermal conductivity model in the ELESTRES computer code

    International Nuclear Information System (INIS)

    Chassie, G.G.; Tochaie, M.; Xu, Z.

    2010-01-01

    This paper describes the improved UO 2 thermal conductivity model for use in the ELESTRES (ELEment Simulation and sTRESses) computer code. The ELESTRES computer code models the thermal, mechanical and microstructural behaviour of a CANDU® fuel element under normal operating conditions. The main purpose of the code is to calculate fuel temperatures, fission gas release, internal gas pressure, fuel pellet deformation, and fuel sheath strains for fuel element design and assessment. It is also used to provide initial conditions for evaluating fuel behaviour during high temperature transients. The thermal conductivity of UO 2 fuel is one of the key parameters that affect ELESTRES calculations. The existing ELESTRES thermal conductivity model has been assessed and improved based on a large amount of thermal conductivity data from measurements of irradiated and un-irradiated UO 2 fuel with different densities. The UO 2 thermal conductivity data cover 90% to 99% theoretical density of UO 2 , temperature up to 3027 K, and burnup up to 1224 MW·h/kg U. The improved thermal conductivity model, which is recommended for a full implementation in the ELESTRES computer code, has reduced the ELESTRES code prediction biases of temperature, fission gas release, and fuel sheath strains when compared with the available experimental data. This improved thermal conductivity model has also been checked with a test version of ELESTRES over the full ranges of fuel temperature, fuel burnup, and fuel density expected in CANDU fuel. (author)

  11. Determination of carbon content of UO2, (U, Gd)O2 and (U, Pu)O2 powders and sintered pellets - Combustion in a high-frequency induction furnace -Infrared absorption spectrometry

    International Nuclear Information System (INIS)

    2008-01-01

    This International Standard describes a method for determining the carbon content in UO 2 , (U,Gd)O 2 and (U,Pu)O 2 powder and sintered pellets by combustion in an induction furnace and infrared absorption spectroscopy measurement. It is applicable for determining 10 μg/g to 500 μg/g of carbon in UO 2 , (U,Gd)O 2 and (U,Pu)O 2 powder and pellets. The sample is heated to a temperature above 1500 deg. C in an induction furnace, under pure oxygen atmosphere, to convert any carbon compounds to carbon dioxide gas. The resulting carbon dioxide gas is filtered and dried before measurement using infrared spectroscopy to measure the carbon dioxide signal at 2350 cm -1 . The result is converted into the carbon content of the material analysed

  12. Fission gas release from ThO2 and ThO2--UO2 fuels (LWBR development program)

    International Nuclear Information System (INIS)

    Goldberg, I.; Spahr, G.L.; White, L.S.; Waldman, L.A.; Giovengo, J.F.; Pfennigwerth, P.L.; Sherman, J.

    1978-08-01

    Fission gas release data are presented from 51 fuel rods irradiated as part of the LWBR irradiations test program. The fuel rods were Zircaloy-4 clad and contained ThO 2 or ThO 2 -UO 2 fuel pellets, with UO 2 compositions ranging from 2.0 to 24.7 weight percent and fuel densities ranging from 77.8 to 98.7 percent of theoretical. Rod diameters ranged from 0.25 to 0.71 inches and fuel active lengths ranged from 3 to 84 inches. Peak linear power outputs ranged from 2 to 22 kw/ft for peak fuel burnups up to 56,000 MWD/MTM. Measured fission gas release was quite low, ranging from 0.1 to 5.2 percent. Fission gas release was higher at higher temperature and burnup and was lower at higher initial fuel density. No sensitivity to UO 2 composition was evidenced

  13. Grain growth in UO2

    International Nuclear Information System (INIS)

    Hastings, I.J.; Scoberg, J.A.; Walden, W.

    1979-06-01

    Grain growth studies have been carried out on UO 2 to provide data for the fuel modelling program and to evaluate fuel fabricated in commissioning the Mixed Oxide Fuel Fabrication Laboratory at Chalk River Nuclear Laboratories. Fuel examined includes natural UO 2 commercially fabricated from ADU powder for CANDU reactors; natural UO 2 commercially fabricated from AU powder; natural UO 2 from ADU and AU powder, fabricated in the MOFFL; and commercially fabricated UO 2 enriched 1.7, 4.5, and 9.6 wt. percent U-235 in U. Samples were step-annealed in vacuo at 1870-2070 K for up to 32.5 h. All data fit a (grain size)sup(2.5) versus annealing time relationship. Apparent activation energy for grain growth, Q, depends on fuel type and varies from 150+-10 kJ/mol for early AU powder to 360+-10 kJ/mol for pellets from ADU fabricated in the MOFFL. Grain sizes calculated using the laboratory equation in a fuel performance code tend to be greater than those measured in irradiated natural fuel, suggesting irradiation-induced inhibition of grain growth. However, any inhibition is equivalent to that expected for a systematic 5 percent underpredicition in reactor power. (author)

  14. 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.)

  15. 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)

  16. High burnup MOX fuel assembly

    International Nuclear Information System (INIS)

    Blanpain, P.; Brunel, L.

    1999-01-01

    From the outset, the MOX product was required to have the same performance as UO 2 in terms of burnup and operational flexibility. In fact during the first years the UO 2 managements could not be applied to MOX. The changeover to an AFA 2G type fuel allowed an improvement in NPP operational flexibility. The move to the AFA 3G design fuel will enable an increase in the burnup of the MOX assemblies to the level of the UO 2 ones ('MOX Parity' project). But the FRAMATOME fuel development objective does not stop at the obtaining of parity between the current MOX and UO 2 products: this parity must remain guaranteed and the MOX managements must evolve in the same way as the UO 2 managements. The goal of the MOX product development programmes underway with COGEMA and the CEA is the demonstration over the next 10 years of a fuel capable of reaching burnups of 70 GWD/T. The research programmes focus on the fission gas release aspect, with three issues explored: optimization of pellet microstructures and validation in experimental reactor ; build-up of experience feedback from fission gas release at elevated burnups in commercial reactors, both for current and experimental products; adaptation and qualification of the design models and tools, over the ranges and for the products concerned. The product arising from these development programmes should be offered on the market around 2010. While meeting safety requirements, it will cater for the needs of the utilities in terms of product reliability, personnel dosimetry and kWh output costs (increase in burnup, NPP maneuverability and availability, minimization of process waste). (authors)

  17. Modification of UO2 grain re-crystallization temperature in function of burn-up as a base for Vitanza experimental curve reconstruction

    International Nuclear Information System (INIS)

    Szuta, M.; Dąbrowski, L.

    2013-01-01

    Crossing the experimental critical fuel temperature dependent on burn-up, an onset of fission gas burst release is observed. This observed phenomena can be explained by assumption that the fission gas immobilization in the uranium dioxide irradiated to a fluency of greater than 10 19 fissions/cm 3 is mainly due to radiation induced chemical activity. Application of the “ab initio” method show that the bond energy of Xenon and Krypton is equal to –1.23 eV, and –3.42 eV respectively. Assuming further that the gas chemically bound can be released mainly in the process of re-crystallization and modifying the differential equation of Ainscough of grain growth by including the burn-up dependence and the experimental data of limiting grain size in function of the fuel temperature for the un-irradiated and irradiated fuel we can re-construct the experimental curve of Vitanza. (authors)

  18. 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.

  19. 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

  20. 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

  1. A simulation of the temperature overshoot observed at high burnup in annular fuel pellets

    Energy Technology Data Exchange (ETDEWEB)

    Baron, D [Electricite de France, Moret-sur-Loing (France); Couty, J C [Electricite de France (EDF), 69 - Villeurbanne (France)

    1997-08-01

    Instrumented experiments have been carried out in recent years to calibrate and improve temperature calculations at high burnup in PWR nuclear fuel rods. The introduction of a thermocouple in the fuel stack allows the experiment to record the centre-line temperature all along the irradiation or re-irradiation. The results obtained on fresh fuel have not revealed any abnormal behavior as have observations done on high burnup rods. In this case, a sudden overshoot has been recorded on the thermocouple temperature above an average power threshold. Several hypotheses have been suggested. Only two seem to be acceptable: one in relation to an effect of grain decohesion, another based on a modification of fuel chemistry. The apparent reversibility of the phenomena when power decreases led us to prefer the first explanation. Indeed, the introduction of a thermocouple means that annular fuel pellets must be used. These are either initially manufactured with a central hole or drilled after base irradiation, using the ``RISOE`` technique. One must bear in mind that the use of such annular pellets drastically changes the crack pattern as irradiation proceeds. This is due to a different stress field which, combined with a weakening of the grain binding energy, leads to a partial grain decohesion on the inner face of the annular pellet. Modification of the grain binding energy is related to the presence of an increasing local population of gas bubbles and metallic precipitates at grain boundaries, as swelling creates intergranular local stresses which also could probably enhance the grain decohesion process. This grain decohesion concerns a 250 to 350 {mu}m depth and shows a narrow cracks network through which released fission gas can flow, temporarily pushing the resident helium gas out. The low conductivity of these gaseous fission products and the numerous gas layers created this way could partly explain the unexpected temperatures measured in high burnup fuels. (Abstract

  2. Mechanism of UO2 selfdisintegration by oxidation

    International Nuclear Information System (INIS)

    Ohai, D.; Furtuna, I.; Dumitrescu, I.

    2008-01-01

    Full text: The paper present the results of the study of UO 2 sintered pellets oxidation, part of FIPRED (Fission Product Release from Debris Bed) Project. The FIPRED Project is dedicated to the study the fission products release from irradiated pellets existing in debris bed. The product release is produced by oxidative self disintegration of sintered pellets at air ingress and it depends on temperature. The experimental program covered experiments of 300-1000 deg. C in air diluted with nitrogen at different oxygen concentrations. The experiments were performed using the SETARAM thermo gravimetric equipment and the FIPRED EQ equipment designed and manufactured especially for this type of experiment. The powders (fragments), resulted from UO 2 pellets self disintegration, were characterized by sieving and SEM. The self disintegration mechanism was demonstrated using the experimental results obtained and thermodynamical data of uranium oxides. (authors)

  3. UO2 fuel behaviour at rod burn-ups up to 105 MWd/kgHM. A review of 10 years of high burn-up examinations commissioned by AREVA NP

    International Nuclear Information System (INIS)

    Goll, W.; Hoffmann, P.B.; Hellwig, C.; Sauser, W.; Spino, J.; Walker, C.T.

    2007-01-01

    Irradiation experience gained on fuel rods with burn-ups greater than 60 MWd/kgHM irradiated in the Nuclear Power Plant Goesgen, Switzerland, is described. Emphasis is placed on the fuel behaviour, which has been analysed by hot cell examinations at the Institute for Transuranium Elements and the Paul-Scherrer-Institute. Above 60 MWd/kgHM, the so-called high burn-up structure (HBS) forms and the fission gas release increases with burn-up and rod power. Examinations performed in the outer region of the fuel revealed that most if not all of the fission gas created was retained in the HBS, even at 25% porosity. Furthermore, the HBS has a relatively low swelling rate, greatly increased plasticity, and its thermal conductivity is higher than expected from the porosity. The post-irradiation examinations showed that the HBS has no detrimental effects on the performance of stationary irradiated PWR fuel irradiated to the high burn-ups that can be achieved with 5 wt% U-235 enrichment. On the contrary, the HBS results in fuel performance that is generally better than it would have been if the HBS had not formed. (orig.)

  4. Deposition of Cr, Nb, V, and Ti coatings on UO2-25w/oPuO2 fuel pellets by sputtering

    International Nuclear Information System (INIS)

    Gibby, R.L.; McClanahan, E.D.

    1976-01-01

    A sputtering deposition process was developed for application of metallic coatings on either the ends or circumferences of LMFBR mixed-oxide fuel pellets. Coatings of Cr, Nb, V and Ti were applied to over 860 pellets. Ceramography, emission spectrography, and spark source spectroscopy were used to characterize the coatings. Coating thicknesses were controlled to within +-0.0005 cm (0.0002 inch) for a coating thickness of 0.00127 cm (0.0005 inch) on the circumference and 0.00254 cm (0.001 inch) on the ends of pellets. Chemical impurities in the coatings were generally less than 0.5 wt percent. The coatings were adherent in all cases, although some interfacial separations were noted with Ti coatings. The results indicated that further optimization of coatings' parameters would result in improvement of the coatings

  5. 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

  6. A simulation of the temperature overshoot observed at high burnup in annular fuel pellets

    International Nuclear Information System (INIS)

    Baron, D.; Couty, J.C.

    1997-01-01

    Instrumented experiments have been carried out in recent years to calibrate and improve temperature calculations at high burnup in PWR nuclear fuel rods. The introduction of a thermocouple in the fuel stack allows the experiment to record the centre-line temperature all along the irradiation or re-irradiation. The results obtained on fresh fuel have not revealed any abnormal behavior as have observations done on high burnup rods. In this case, a sudden overshoot has been recorded on the thermocouple temperature above an average power threshold. Several hypotheses have been suggested. Only two seem to be acceptable: one in relation to an effect of grain decohesion, another based on a modification of fuel chemistry. The apparent reversibility of the phenomena when power decreases led us to prefer the first explanation. Indeed, the introduction of a thermocouple means that annular fuel pellets must be used. These are either initially manufactured with a central hole or drilled after base irradiation, using the ''RISOE'' technique. One must bear in mind that the use of such annular pellets drastically changes the crack pattern as irradiation proceeds. This is due to a different stress field which, combined with a weakening of the grain binding energy, leads to a partial grain decohesion on the inner face of the annular pellet. Modification of the grain binding energy is related to the presence of an increasing local population of gas bubbles and metallic precipitates at grain boundaries, as swelling creates intergranular local stresses which also could probably enhance the grain decohesion process. This grain decohesion concerns a 250 to 350 μm depth and shows a narrow cracks network through which released fission gas can flow, temporarily pushing the resident helium gas out. The low conductivity of these gaseous fission products and the numerous gas layers created this way could partly explain the unexpected temperatures measured in high burnup fuels. The purpose of

  7. French analytic experiment on the high specific burnup of PWR fuels in normal conditions

    International Nuclear Information System (INIS)

    Bruet, M.; Atabek, R.; Houdaille, B.; Baron, D.

    1982-04-01

    Hydrostatic density determinations made on UO 2 pellets of different kinds irradiated in conditions representative of PWR conditions enable the internal swelling rate of the UO 2 to be ascertained. A mean value of 0.8% per 10 4 MWdt -1 (u) up to a specific burnup of 45000 MWdt -1 (u) may be deduced from this experimental basis. These results agree well with those obtained in the TANGO experiments in which UO 2 balls were irradiated in quasi isothermal conditions and without stress. Further, the open porosity of oxide closes progressively and the change in the total porosity is thus very limited (under 1% at 45000 MWdt -1 (u)). With respect to the swelling of the pellets the rise in the specific burnup would not appear therefore to be a problem. The behaviour of recrystallized zircaloy 4 claddings remains satisfactory with respect to creep and growth during irradiation [fr

  8. 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

  9. 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)

  10. Sinterability of mixtures of UO2 of different morphological features

    International Nuclear Information System (INIS)

    Villegas de Maroto, Marina; Celora de Lavagnino, Julia; Marajofsky, Adolfo; Leyva, A.G.

    1981-01-01

    The reprocessing of scrap in the production of UO2 pellets, is important from an economical view-point of the fuel cycle. The recovery method by means of a humid process, tested for UO2 scrap, includes the dissolution of the pellets in a nitric media at boiling point, the precipitation of ammonium diuranates (ADU) and its conversion into UO2 at 600 deg C. The microestructural results and the sintering density of the pellets produced in these tests are compared. It is shown that, although the addition of said UO2 powders impaires the performance of the original mixture produced by the factory, the results thus obtained are, nevertheless, within specifications. This facts show that the mixture would then be able for production. (M.E.L.) [es

  11. Micropore sintering in UO2 irradiated below 11000C in the Winfrith SGHWR

    International Nuclear Information System (INIS)

    Banks, D.A.; Clough, D.J.

    1979-01-01

    The densification behaviour of UO 2 irradiated in the Winfrith steam generating heavy water reactor has been examined using optical and scanning electron microscope techniques and by measuring the densities of whole pellets and small cores removed from pellet sections. Micrography has shown that in the temperature range 400 to 1000 0 C densification is due to the in-factor sintering of micropores (approximately< 2 m in diameter) with a temperature dependency defined by an activation energy of 21.7 kJ/mol. Pore area measurements, allied to previously published work, have led to the derivation of an equation which can be used to predict micropore volume changes occurring during irradiation and takes account of both the temperature and burn-up dependency of pore sintering. (author)

  12. Fuel chemistry and pellet-clad interaction related to high burnup fuel. Proceedings of the technical committee

    International Nuclear Information System (INIS)

    2000-10-01

    The purpose of the meeting was to review new developments in clad failures. Major findings regarding the causes of clad failures are presented in this publication, with the main topics being fuel chemistry and fission product behaviour, swelling and pellet-cladding mechanical interaction, cladding failure mechanism at high burnup, thermal properties and fuel behaviour in off-normal conditions. This publication contains 17 individual presentations delivered at the meeting; each of them was indexed separately

  13. 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)

  14. 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.

  15. Survey of the power ramp performance testing of KWU'S PWR UO 2, fuel

    Science.gov (United States)

    Ga¨rtner, M.; Fischer, G.

    1987-06-01

    To determine the power ramp performance of KWU's PWR UO 2 fuel, 134 fuel rodlets with burnups of up to 46 GWd/ t (U) and several fuel assemblies with 19 to 30 GWd/t (U) burnup were ramped in power in the research reactors HFR Petten/The Netherlands and R2 Studsvik/Sweden and in the power plants KWO and KWB-A/Germany, respectively. The power ramp tests demonstrate decreasing resistance of the PWR fuel rods to PCI (pellet-to-clad interaction) up to fuel burnups of 35 GWd/t (U) and a reversal effect at higher burnups. The fuel rods can be operated free of defects at fast power transients to linear heat generation rates of up to 400 W/cm, at least.Power levels of up to 490 W/cm can be reached without defects by reducing the ramp rate. After reshuffling according to an out-in scheme, 1-cycle fuel assemblies may return to rod powers of up to 480 W/cm with a power increase rate of up to 10 W/(cm min) without fuel rod damage. Set points basing on these test results and incorporated into the power distribution control and power density limitation system of KWU's advanced power plants guarantee safe plant operation under normal and load follow operating conditions.

  16. 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

  17. A prediction of the UO2 fission gas release data of Bellamy and Rich using a model recently developed by combustion engineering

    International Nuclear Information System (INIS)

    Freeburn, H.R.; Pati, S.R.

    1983-01-01

    The trend in the Light Water Reactor industry to higher discharge burnups of UO 2 fuel rods has initiated the modification of existing fuel rod models to better account for high burnup effects. A model recently developed by Combustion Engineering, Inc. (C-E) for fission gas release from UO 2 fuel recognizes the separate effects of temperature-dependent and temperature-independent release mechanisms. This model accounts for a moderate burnup enhancement that is based on a concept of a saturation inventory existing for the intra- and inter-grannular storage of fission gas within the fuel pellet. The saturation inventory, as modelled, is strongly dependent on the local temperature and the changing grain size of the fuel with burnup. Although the fitting constants of the model were determined solely from more current gas release data from fuel more typical of the C-E product line, the model, nonetheless, provides an excellent prediction of the Bellamy and Rich data over the entire burnup range represented by the data (+-1.6% gas release at a 1σ level). The ability to obtain a good comparison with this data base provides additional support for the use of the particular separation of the effects of thermal diffusion and burnup enhancement on fission gas release that is embodied in the model. Furthermore, the degree of burnup enhancement in the model is believed to be moderate enough to suggest that this high burnup effect should not impede the extension of discharge burnup limits associated with current design fuel rods for Pressurized Water Reactors

  18. 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

  19. A Knowledge- Based Computer System for UO2 Characterization According to ASTM Requirements

    International Nuclear Information System (INIS)

    Afifi, Y.K.; El-Hakim, E.

    2000-01-01

    The uranium dioxde (UO 2 ) powder properties and the pellets fabrication processes determine the characteristics of the sintered UO 2 pellets. The powder properties include chemical and physical characteristics. The physical and chemical properties of UO 2 powder are normally checked to ensure consistency and reproducibility of the sintered UO 2 pellets. Powder characteristics are known to influence the subsequent manufacturing performance or the fuel properties. The aim of this paper is to provide the nuclear industry with a program dealing with the processes and the related requirements to determine the specifications of UO 2 powder according to the American Standards for Testing and Materials (ASTM). This program covers the physical and chemical characteristics of UO 2 powder. A group of logic flow charts dealing with the data and information available in the ASTM for each step in the characterization of UO 2 powder process and the technical assistance are constructed. These logic flow charts are collected to form a module of the software to qualify the UO 2 powder. The program contains 8 modules, each one deals with one object. This program saves time, is also considered as a collective schema for all the required UO 2 powder characterization and the related processes, and could be used as a training tool for less skilled personnel involved in UO 2 powder characterization laboratories

  20. 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

  1. 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

  2. Technological aspects of UO2 sintering at low temperature

    International Nuclear Information System (INIS)

    Thern, Gerardo G.; Dominguez, Carlos A.; Benitez, Ana M.; Marajofsky, Adolfo

    1999-01-01

    Within the Fuel Cycle Program of CNEA, the knowledge that plant personnel has on sintering at low temperature was evaluated, because this process could decrease costs for UO 2 and (U,Gd)O 2 pellets production, simplify the furnace maintenance and facilitate the automation of the production process, specially convenient for uranium recovery. By applying this technology, some companies have achieved production at pilot-scale and irradiated a significant number of pellets. (author)

  3. Effect of PCMI restraint on bubble size distribution in the rim structure of UO2 fuel

    International Nuclear Information System (INIS)

    Oh, Je-Yong; Koo, Yang-Hyun; Cheon, Jin-Sik; Lee, Byung-Ho; Sohn, Dong-Seong

    2005-01-01

    Generally, the bubble size in the rim structure of UO 2 is not dependent on the fuel burnup and the bubble pressure is higher than that in the equilibrium condition. However it was also observed that if the fuel pellet is not restrained, the size of the bubbles in the rim structure could be larger than that in the restraint condition. Although the wide variety of rim bubble sizes and porosities possibly result from an external restrain effect, the quantitative method to analyze the effect of PCMI restraint on bubble distribution in the rim is not available at the moment. In this paper, a method is developed which can be used to analyze the effect of PCMI restraint on the bubble distribution in the rim structure of UO 2 fuel based on the data in the literatures. The total number of Xe atoms in the rim bubbles per unit rim volume could be derived by a summation of the number of Xe atoms of each rim bubble in a unit rim volume. The number of Xe atoms of each rim bubble could be calculated by the Van der Waals equation of state and the pressure expressed by p=σ+C/r, where C is an unknown constant to be determined as a function of the temperature and the burnup. On the other hand, the total number of Xe atoms in the rim bubbles per unit rim volume can also be calculated by Xe depression data. If the fuel pellet is not restrained, the uniform hydrostatic stress, σ is zero. Hence if the data of the fuel disk without a restraint is used, a constant C can be obtained at 823K and a local burnup of 90 GWd/t. Although the local burnup of PCMI restraint case is slightly different from that without PCMI restraint, the value derived above is used for the analysis of PCMI restraint case. The calculated bubble distribution with PCMI restraint was similar to the measured one. Because the effect of PCMI restraint on bubble size increased with the bubble size, the development of a large bubble was suppressed. Hence, the PCMI restraint caused a typical bubble size in the rim and

  4. 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.)

  5. Design and analytic evaluation of a rim effect reduction type LWR fuel for extending burnup

    International Nuclear Information System (INIS)

    Matsumura, Tetsuo; Kameyama, Takanori; Kinoshita, Motoyasu

    1991-01-01

    We have designed a new concept fuel design 'Rim effect reduction type fuel' which has thin natural UO 2 layer on surface of a UO2 pellet. Our neutronic analyses with ANRB code show this fuel design can reduce rim effect (burnup at plelet rim) by about 30 GWd/t comparing a normal fuel. It is known that a high burnup fuel has different microstructure from as-fabricated one at fuel rim (which is called as rim region) due to rim effect. Therefore this fuel design can expect smaller rim region than a normal fuel. Our fuel performance analyses with EIMUS code show this fuel design can reduce fuel center temperature at high burnup if thermal conductivity of fuel pellet decreases with burnup in inverse proportion. However, this fuel design increases fuel center temperature at low and middle burnup than a normal fuel due to increase of thermal power density at pellet center. Additionally Irradiation experiment of this fuel design can be considered to offer important data which make clear the relation between rim effect and fuel performance. (author)

  6. Oxidation of UO2 at 400 to 1000 degrees C in air and its relevance to fission product release

    International Nuclear Information System (INIS)

    McCracken, D.R.

    1985-07-01

    Currently there is great interest in the behaviour of UO 2 under oxidizing conditions because irradiated uranium dioxide fuel can conceivably be exposed to a hot oxidizing atmosphere as a result of accidents. The temperature range covered in this paper is 400 to 1000 degrees C. At these high temperatures, UO 2 in air can oxidize rapidly to U 3 O 8 via U 3 O 7 and/or U 4 O 9 . The accompanying volume increase and corresponding stresses lead to fragmentation of the fuel pellets. The purpose of this work was to investigate the dependence of UO 2 oxidation on temperature, rate of air supply and residence time at temperature; to determine the rate controlling steps and rate of oxygen penetration; and to characterize the oxidation products and size of fragments. In addition, detailed metallography was related to X-ray diffraction studies of the oxidized UO 2 to facilitate future study of irradiated fuel, which is easier to do by metallography in hot-cells than by X-ray diffraction. Samples were heated in argon, then once at temperature they were exposed to air at a controlled flow-rate. Studies of the oxidation of unirradiated UO 2 pellets in air show two distinct types of oxidation with a change in mechanism at 600-700 degrees C. At temperatures ≤ 600 degrees C fragmentation accompanies the formation of U 3 O 8 while at T ≥ 800 degrees C, rapid grain growth occurs. In the first temperature region, volatile fission product releases are small, while in the second region, 100% release can be correlated with U 3 O 8 formation. In the first region, only the grain boundary inventory is released while in the other, 100% of the Xe, Kr, Ru, Sb, Cs and I are released. It appears that, within the error of present measurements, burnup does not affect rates of fission product release and oxidation in air at 400 to 1000 degrees C, so that oxidation rate data gathered using unirradiated pellets can be applied to irradiated fuel. 33 refs

  7. 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

  8. Acoustic emission from thermal-gradient cracks in UO2

    International Nuclear Information System (INIS)

    Kennedy, C.R.; Kupperman, D.S.; Wrona, B.J.

    1975-01-01

    A feasibility study has been conducted to evaluate the potential use of acoustic emission to monitor thermal-shock damage in direct electrical heating of UO 2 pellets. In the apparatus used for the present tests, two acoustic-emission sensors were placed on extensions of the upper and lower electrical feedthroughs. Commercially available equipment was used to accumulate acoustic-emission data. The accumulation of events displayed on a cathode-ray-tube screen indicates the total number of acoustic-emission events at a particular location within the pellet stack. These tests have indicated that acoustic emission can be used to monitor thermal-shock damage in UO 2 pellets subjected to direct-electrical heating. 8 references

  9. 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

  10. Defect trap model of gas behaviour in UO2 fuel during irradiation

    International Nuclear Information System (INIS)

    Szuta, A.

    2003-01-01

    Fission gas behaviour is one of the central concern in the fuel design, performance and hypothetical accident analysis. The report 'Defect trap model of gas behaviour in UO 2 fuel during irradiation' is the worldwide literature review of problems studied, experimental results and solutions proposed in related topics. Some of them were described in details in the report chapters. They are: anomalies in the experimental results; fission gas retention in the UO 2 fuel; microstructure of the UO 2 fuel after irradiation; fission gas release models; defect trap model of fission gas behaviour; fission gas release from UO 2 single crystal during low temperature irradiation in terms of a defect trap model; analysis of dynamic release of fission gases from single crystal UO 2 during low temperature irradiation in terms of defect trap model; behaviour of fission gas products in single crystal UO 2 during intermediate temperature irradiation in terms of a defect trap model; modification of re-crystallization temperature of UO 2 in function of burnup and its impact on fission gas release; apparent diffusion coefficient; formation of nanostructures in UO 2 fuel at high burnup; applications of the defect trap model to the gas leaking fuel elements number assessment in the nuclear power station (VVER-PWR)

  11. 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.)

  12. UO2: production based on two alternative lines

    International Nuclear Information System (INIS)

    Coppa, R.C.; Martin, H.R.

    1987-01-01

    The production of the uranium dioxide (UO 2 ) is carried out at the Cordoba factory, of the Argentine National Atomic Energy Commission, by the uranil carbonate method (AUC). The commercial uranium concentrates (yellow cake) is dissolved with HNO 3 and purificated with tributil phosphate (TBP). The pure uranium compound coming from the reextraction, is concentrated to 0.4 Kg U/l, then the precipitation with CO 2 and NH 3 gives the AUC crystalls. After conversion of AUC to UO 2 powder, the pellets are obtained by direct compacting. In the second experimental method used by CNEA, the yellow cake is dissolved with H 2 SO 4 , and then it is purified with a terciary amine and precipitated with (NH 4 ) 2 CO 3 . In this form the ammonium uranil tri-carbonate (AUT) crystals are obtained. The convertion to UO 2 is made under an atmosphere of dissociated NH 3 . (M.E.L.) [es

  13. Preparation of UO2 fragments for fuel-debris experiments

    International Nuclear Information System (INIS)

    Tinkle, M.C.; Kircher, J.A.; Zinn, R.M.; Eash, D.T.

    1982-01-01

    A unique process was developed for preparing multi-kilogram quantities of > 90% dense fragments of enriched and depleted UO 2 sized 20 mm to 0.038 mm for fuel debris experiments. Precipitates of UO 4 . xH 2 O were treated to obtain UO 2 powders that would yield large cohesive green pieces when isostatically pressed to 206 MPa. The pressed pieces were crushed into fragments that were about 30% oversized, and heated to 1800 0 C for 24 h in H 2 . Oversizing compensates for shrinkage during densification. Effort was dramatically reduced by working on a large scale and by presizing the green UO 2 instead of directly crushing densified pellets

  14. On the Role of the Electrical Field in Spark Plasma Sintering of UO2+x

    Science.gov (United States)

    Tyrpekl, Vaclav; Naji, Mohamed; Holzhäuser, Michael; Freis, Daniel; Prieur, Damien; Martin, Philippe; Cremer, Bert; Murray-Farthing, Mairead; Cologna, Marco

    2017-01-01

    The electric field has a large effect on the stoichiometry and grain growth of UO2+x during Spark Plasma Sintering. UO2+x is gradually reduced to UO2.00 as a function of sintering temperature and time. A gradient in the oxidation state within the pellets is observed in intermediate conditions. The shape of the gradient depends unequivocally on the direction of the electrical field. The positive surface of the pellet shows a higher oxidation state compared to the negative one. An area with larger grain size is found close to the positive electrode, but not in contact with it. We interpret these findings with the redistribution of defects under an electric field, which affect the stoichiometry of UO2+x and thus the cation diffusivity. The results bear implications for understanding the electric field assisted sintering of UO2 and non-stoichiometric oxides in general. PMID:28422164

  15. Highlights on R and D work related to the achievement of high burnup with MOX fuel in commercial reactors

    International Nuclear Information System (INIS)

    Lippens, M.; Maldague, Th.; Basselier, J.; Boulanger, D.; Mertens, L.

    2000-01-01

    Part of the R and D work made at BELGONUCLEAIRE in the field of high burnup achievement with MOX fuel in commercial LWRs is made through lnternational Programmes. Special attention is given to the evolution with burnup of fuel neutronic characteristics and of in-reactor rod thermal-mechanical behaviour. Pu burning in MOX is characterized essentially by a drop of Pu 239 content. The other Pu isotopes have an almost unchanged concentration, due to internal breeding. The reactivity drop of MOX versus burnup is consequently much less pronounced than in UO 2 fuel. Concentration of minor actinides Am and Cm becomes significant with burnup increase. These nuclides start to play a role on total reactivity and in the helium production. The thermal-mechanical behaviour of MOX fuel rod is very similar to that of UO 2 . Some specificities are noticed. The better PCI resistance recognized to MOX fuel has recently been confirmed. Three PWR MOX segments pm-irradiated up to 58 GWd/tM were ramped at 100 W/cm.min respectively to 430-450-500 W/cm followed by a hold time of 24 hours. No segment failed. MOX and UO 2 fuels have different reactivities and operate thus at different powers. Moreover, radial distribution of power in MOX pellet is less depressed at high burnup than in UO 2 , leading to higher fuel central temperature for a same rating. The thermal conductivity of MOX fuel decreases with Pu content, typically 4% for 10% Pu. The combination of these three elements (power level, power profile, and conductivity) lead to larger FGR at high burnup compared to UO 2 . Helium production remains low compared to fission gas production (ratio < 0.2). As faster diffusing element, the helium fractional release is much higher than that of fission gas, leading to rod pressure increase comparable to the one resulting from fission gas. (author)

  16. Preparation of high density (8 to 9) uranium oxide UO2

    International Nuclear Information System (INIS)

    Eichner, C.; Ertaud, A.; Ortel, Y.; Stohr, J.; Vautrey, L.

    1948-10-01

    This report describes the process elaborated for the preparation of high density UO 2 . The thermal decomposition of uranium peroxide leads to UO 3 which is reduced by an hydrogen flow to obtain UO 2 . A UO 2 powder of good quality is obtained for temperatures below 650 deg. C. The powder is pulverized to obtain an homogeneous grain size and compressed inside a die to make pellets. Pellets are sintered up to 1600 deg. C in a reducing atmosphere and following a temperature rise law of 150 deg. C/hour. The equipment used (furnaces, gases purifier, control equipment, power supplies, thermoregulation systems) is described at the end. (J.S.)

  17. Electrochemical Reduction of solid UO2 in Molten Fluoride Salts

    International Nuclear Information System (INIS)

    Gibilaro, Mathieu; Cassayre, Laurent; Massot, Laurent; Chamelot, Pierre; Malmbeck, Rikard; Dugne, Olivier; Allegri, Patrick

    2010-01-01

    The direct electrochemical reduction of UO 2 solid pellets was carried out in LiF-CaF 2 (+ 2wt % Li 2 O) at 850 deg. C. An inert gold anode was used instead of the usual reactive sacrificial carbon anode. In this case, reduction of oxide ions yields O 2 gas evolution on the anode. Electrochemical characterisations of UO 2 pellets have been performed by linear sweep voltammetry at 10 mV/s and reduction waves associated to its direct reduction have been observed at a potential 150 mV more positive in comparison with the solvent reduction. Then, galvano-static electrolyses runs have been realised and products were characterised by SEM-EDX, EPMA/WDS and XRD. In one of the runs, uranium oxide was partially reduced and three phases were observed: non reduced UO 2 in the centre, pure metallic uranium on the external layer and an intermediate phase representing the initial stage of reduction taking place at the grain boundaries. In another run, the UO 2 sample was fully reduced. Due to oxygen removal, the U matrix had a typical coral-like structure which is characteristic of the pattern observed after the electroreduction of solid oxides. (authors)

  18. Use of UO 2 films for electrochemical studies

    Science.gov (United States)

    Miserque, F.; Gouder, T.; Wegen, D. H.; Bottomley, P. D. W.

    2001-10-01

    UO 2 films have been prepared by dc reactive sputtering of a uranium metal target in an Ar/O 2 atmosphere. We have used the films deposited on gold substrates as working electrodes for electrochemical investigations as simulating the surfaces of fuel pellets. Film composition was determined by photoelectron spectroscopy (XPS and UPS) and X-ray diffraction (XRD). The oxide stoichiometry as a function of deposition conditions was determined and the appropriate conditions for UO 2.0 formation established. AC impedance and cyclic voltammetry measurements were performed. A double RC electrical equivalent circuit was used to fit the data from impedance measurements, similar to those used in unirradiated UO 2 or spent fuel pellets. However due to the porosity or adhesion defects on the thin films that permitted a direct contact between the solution and the gold substrate, we were obliged to add a contribution simulating the water-gold system. Cyclic voltammetry measurements show the influence of pH on the dissolution mechanism. Alkaline solutions permit the formation of an oxidised layer (UO 2.33) which is not present in the acidic solutions. In both pH=2 and pH=6 solutions, a U VI species layer is formed.

  19. Results of the irradiation of mixed UO2 - PuO2 oxide fuel elements

    International Nuclear Information System (INIS)

    Mikailoff, H.; Mustelier, J.P.; Bloch, J.; Ezran, L.; Hayet, L.

    1966-01-01

    In order to study the behaviour of fuel elements used for the first charge of the reactor Rapsodie, a first batch of eleven needles was irradiated in the reactor EL3 and then examined. These needles (having a shape very similar lo that of the actual needles to be used) were made up of a stack of sintered mixed-oxide pellets: UO 2 containing about 10 per cent of PuO 2 . The density was 85 to 97 per cent of the theoretical, value. The diametral gap between the oxide and the stainless steel can was between 0,06 and 0,27 mm. The specific powers varied from 1230 to 2700 W/cm 3 and the can temperature was between 450 and 630 C. The maximum burn-up attained was 22000 MW days/tonne. Examination of the needles (metrology, radiography and γ-spectrography) revealed certain macroscopic changes, and the evolution of the fuel was shown by micrographic studies. These observations were used, together with flux measurements results, to calculate the temperature distribution inside the fuel. The volume of the fission gas produced was measured in some of the samples; the results are interpreted taking into account the temperature distribution in the oxide and the burn-up attained. Finally a study was made both of the behaviour of a fuel element whose central part was molten during irradiation, and of the effect of sodium which had penetrated into some of the samples following can rupture. (author) [fr

  20. High Frequency Acoustic Microscopy for the Determination of Porosity and Young's Modulus in High Burnup Uranium Dioxide Nuclear Fuel

    Science.gov (United States)

    Marchetti, Mara; Laux, Didier; Cappia, Fabiola; Laurie, M.; Van Uffelen, P.; Rondinella, V. V.; Wiss, T.; Despaux, G.

    2016-06-01

    During irradiation UO2 nuclear fuel experiences the development of a non-uniform distribution of porosity which contributes to establish varying mechanical properties along the radius of the pellet. Radial variations of both porosity and elastic properties in high burnup UO2 pellet can be investigated via high frequency acoustic microscopy. For this purpose ultrasound waves are generated by a piezoelectric transducer and focused on the sample, after having travelled through a coupling liquid. The elastic properties of the material are related to the velocity of the generated Rayleigh surface wave (VR). A UO2 pellet with a burnup of 67 GWd/tU was characterized using the acoustic microscope installed in the hot cells of the JRC-ITU at a 90 MHz frequency, with methanol as coupling liquid. VR was measured at different radial positions. A good agreement was found, when comparing the porosity values obtained via acoustic microscopy with those determined using SEM image analysis, especially in the areas close to the centre. In addition, Young's modulus was calculated and its radial profile was correlated to the corresponding burnup profile and to the hardness radial profile data obtained by Vickers micro-indentation.

  1. 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

  2. Characterisation and compaction behaviour of UO2 powder prepared from ADU and AUC

    International Nuclear Information System (INIS)

    Rachmawati, M.

    2000-01-01

    UO 2 powder prepared from ADU and AUC route process are characterised primarily in terms of compaction and sintering behaviour. Scientific understanding of the phenomena will give useful information leading to processing and product improvement. The investigation has been done by characterising the particle size/shape distribution using SEM and compacting the powder at 4 and 5.4 tons/cm 2 . The behaviour of the powder under compaction is observed by characterizing the pellet length, green density, microstructure, and the compression strength using micrometer SEM, and Universal Testing Machine. The results of the experiment show that the UO 2 powder ex-AUC has particles of spherical type and separate individually which provide the flowable characteristic, important for the die filling aspect during compaction step. The UO 2 powder ex-ADU is more or less agglomerated and contains very fine particles causing the difficulty in pressing. Therefore the green density resulted from UO 2 ex-AUC (6.415 g/cm 3 ) is higher than UO 2 powder of UO 2 ex-ADU (6.117 g/cm 3 . UO 2 at lower pressure (4 tons/cm 3 ) the compression strength ex-AUC green pellet (47.144 kgf) is lower than UO 2 ex-ADU (63,364 kgf), and at higher temperature the compression strength of ex-AUC (92.86 kgf) is higher than UO 2 ex-ADU (82.664 kgf). It is suggested that UO 2 ex-ADU has to be precompacted and granulated in order to increase its flowability so that the pellet length can easily be controlled during pressing (improve reproducibility). (author)

  3. 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

  4. Fabrication of metallic channel-containing UO2 fuels

    International Nuclear Information System (INIS)

    Yang, Jae Ho; Song, Kun Woo; Kim, Keon Sik; Jung, Youn Ho

    2004-01-01

    The uranium dioxide is widely used as a fuel material in the nuclear industry, owing to many advantages. But it has a disadvantage of having the lowest thermal conductivity of all kinds of nuclear fuels; metal, carbide, nitride. It is well known that the thermal conductivity of UO 2 fuel is enhanced by making, so called, the CERMET (ceramic-metal) composite which consists of both continuous body of highly thermal-conducting metal and UO 2 islands. The CERMET fuel fabrication technique needs metal phase of at least 30%, mostly more than 50%, of the volume of the pellet in order to keep the metal phase interconnected. This high volume fraction of metal requires such a high enrichment of U that the parasitic effect of metal should be compensated. Therefore, it is attractive to develop an innovative composite fuel that can form continuous metal phase with a small amount of metal. In this investigation, a feasibility study was made on how to make such an innovative fuel. Candidate metals (W, Mo, Cr) were selected, and fabrication process was conceptually designed from thermodynamic calculations. We have experimentally found that a metal phase envelops perfectly UO 2 grains, forming continuous channel throughout the pellet, and improving the thermal conductivity of pellet

  5. 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

  6. Achieving higher productivity of UO2 fuel at NUOFP through improved in-plant quality surveillance

    International Nuclear Information System (INIS)

    Meena, R.; Pramanik, D.; Sairam, S.; Rajkumar, J.V.; Rao, R.V.R.L.V.; Sinha, T.K.; Santra, N.; Rao, G.V.S.H.; Jayaraj, R.N.

    2009-01-01

    At Nuclear Fuel Complex (NFC), in the production of UO 2 fuel for PHWRs, a standard set of process parameters are monitored regularly for every lot of powder and pellet. Quality of intermediate products in the production process like UNP, ADU(dry), U 3 O 8 , UO 2+x , UO 2 granules, green pellets, sintered pellets are also regularly analysed/monitored apart from the final finished pellet and ensured to be within specified range. This range is decided by final product specifications and sometimes also based on the feed requirement in the next process in the downstream of the flow sheet. Vast experience gained over the years, behavior of various equipment under given set of conditions, feed back from the customer plants etc; have been primary criteria hither to, for defining the process conditions and chemical/physical properties of intermediate products

  7. Relationship between changes in the crystal lattice strain and thermal conductivity of high burnup UO{sub 2} pellets

    Energy Technology Data Exchange (ETDEWEB)

    Amaya, Masaki, E-mail: amaya.masaki@jaea.go.j [Fuel Safety Research Group, Nuclear Safety Research Center, Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Nakamura, Jinichi; Fuketa, Toyoshi [Fuel Safety Research Group, Nuclear Safety Research Center, Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Kosaka, Yuji [Nuclear Development Corporation, 622-12, Funaishikawa, Tokai-mura, Naka-gun, Ibaraki 319-1111 (Japan)

    2010-01-01

    Two kinds of disk-shaped UO{sub 2} samples (4 mm in diameter and 1 mm in thickness) were irradiated in a test reactor up to about 60 and 130 GWd/t, respectively. The microstructures of the samples were investigated by means of optical microscopy, scanning electron microscopy/ electron probe micro-analysis (SEM/EPMA) and micro-X-ray diffractometry. The measured lattice parameters tended to be considerably smaller than the reported values, and the typical cauliflower structure which is often observed in high burnup fuel pellet is hardly seen in these samples. Thermal diffusivities of the samples were also measured by using a laser flash method, and their thermal conductivities were evaluated by multiplying the heat capacity of unirradiated UO{sub 2} and sample densities. While the thermal conductivities of sample 2 showed recovery after being annealed at 1500 K, those of sample 4 were not clearly observed even after being annealed at 1500 K. These trends suggest that the amount of accumulated irradiation-induced defects depends on the irradiation condition of each sample. From the comparison of the changes in the lattice parameter and strain energy density before and after the thermal diffusivity measurements, it is likely that the thermal conductivity recovery in the temperature region from 1200 to 1500 K is related to the migration of dislocation.

  8. Analysis of effects of pellet-cladding bonding on trapping of the released fission gases in high burnup KKL BWR fuels

    Energy Technology Data Exchange (ETDEWEB)

    Brankov, Vladimir [Laboratory for Reactor Physics and Systems Behaviour at the Paul Scherrer Institute, 5232 Villigen-PSI (Switzerland); Swiss Federal Institute of Technology Lausanne (EPFL), Route Cantonale, 1015 Lausanne (Switzerland); Khvostov, Grigori; Mikityuk, Konstantin [Laboratory for Reactor Physics and Systems Behaviour at the Paul Scherrer Institute, 5232 Villigen-PSI (Switzerland); Pautz, Andreas [Laboratory for Reactor Physics and Systems Behaviour at the Paul Scherrer Institute, 5232 Villigen-PSI (Switzerland); Swiss Federal Institute of Technology Lausanne (EPFL), Route Cantonale, 1015 Lausanne (Switzerland); Restani, Renato; Abolhassani, Sousan [Laboratory for Nuclear Materials at the Paul Scherrer Institute, 5232 Villigen-PSI (Switzerland); Ledergerber, Guido [Kernkraftwerk Leibstadt, 5325 Leibstadt (Switzerland); Wiesenack, Wolfgang [Institutt for Energiteknikk - OECD Halden Reactor Project, Os Allé 5, 1777 Halden (Norway)

    2016-08-15

    Highlights: • Explanation for the scatter in measured fission gas release in high-BU BWR fuel rods. • Partial fuel-clad bond layer formation in high-BU BWR fuel. • Hypothesis for fission gas trapping facilitated by the pellet-cladding bond layer. • Correlation between burnup asymmetry and the quantity of trapped fission gas. • Implications of the trapped FG in LOCA transient. - Abstract: The first part of the paper presents results of a numerical analysis of the fuel behavior during base irradiation in the Kernkraftwerk Leibstadt Boiling Water Reactor (KKL BWR) using EPRI’s FALCON code coupled to GRSW-A – an advanced model for fuel swelling and fission gas release. Post-irradiation examinations conducted at the Paul Scherrer Institute’s (PSI) hot laboratory gave evidence of a distinct circumferential non-uniformity of local burnup at pellet surfaces. For several fuel samples, intact pellet-cladding bonding areas on the high burnup sides of the pellets at high burnup above ∼70 MWd/kgU were observed. It is hypothesized that a part of the fission gases, which are expected to be released by those areas, can be trapped and do not reach the rod plenum. In this paper, a simple approach to modeling of fission gas trapping is employed which reveals a potential correlation between the position of the rod within the fuel assembly (and therefore the degree of circumferential burnup non-uniformity) and the degree of fission gas trapping. A model is suggested to correlate the amount of locally trapped gas with the integral of the local contact pressure and the degree of circumferential burnup non-uniformity. The model is calibrated with available measurements of FGR from rod puncturing at the level of the plenums. In future work, the hypothesis about the axial distribution of trapped fission gas will be extrapolated to the Loss-Of-Coolant Accident (LOCA) analysis as an attempt to explain the fission gas release observed in some samples fabricated from

  9. Sintering diagrams of UO2

    International Nuclear Information System (INIS)

    Mohan, A.; Soni, N.C.; Moorthy, V.K.

    1979-01-01

    Ashby's method (see Acta Met., vol. 22, p. 275, 1974) of constructing sintering diagrams has been modified to obtain contribution diagrams directly from the computer. The interplay of sintering variables and mechanisms are studied and the factors that affect the participation of mechanisms in UO 2 are determined. By studying the physical properties, it emerges that the order of inaccuracies is small in most cases and do not affect the diagrams. On the other hand, even a 10% error in activation energies, which is quite plausible, would make a significant difference to the diagram. The main criticism of Ashby's approach is that the numerous properties and equations used, communicate their inaccuracies to the diagrams and make them unreliable. The present study has considerably reduced the number of factors that need to be refined to make the sintering diagrams more meaningful. (Auth.)

  10. On the sintering kinetics in UO2

    International Nuclear Information System (INIS)

    Marajofsky, A.

    1998-01-01

    The fabrication process of UO 2 pellets from powders involve pressing and a sintering anneal at high temperature (1650 deg. C to 1750 deg. C) during two or more hours in a hydrogen atmosphere. An alternative method is the oxidative sintering, made at lower temperature (1000 deg. C to 1300 deg. C) in a CO 2 or CO/CO 2 atmosphere. The sintering phenomena consist in the densification of the material by a thermal treatment below the fusion point. For a compact made by pressing a powder, sintering is the process of annulation of the porosity present in the compact or pellet. Several theories describe the sintering phenomena dividing it in three stages, initial, intermediate and final: in all of them the densification is a continuous growing function of time. Nevertheless it has been experimentally reported that a reduction of the density occurs in the third step of the sintering. The phenomena has been called solarization. Solarization has been attributed to the effect of the evolved gases from additives or to the CO 2 atmosphere in oxidative sintering. Thus, it is convenient to distinguish between solarization in oxidative or reducing conditions. Reducing solarization is a consequence of the tendency towards equilibrium of intergranular pores. In oxidative sintering it occurs in the reducing anneal after the sintering and is due to the change in the lattice parameter. This work shows examples of both types of solarization and qualitative interpretation of this phenomena. Both situations show the need of strict control of the sintering and powder production conditions. (author)

  11. 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

  12. Effects of pellet-to-cladding gap design parameters on the reliability of high burnup PWR fuel rods under steady state and transient conditions

    International Nuclear Information System (INIS)

    Tas, Fatma Burcu; Ergun, Sule

    2013-01-01

    Highlights: • Fuel performance of a typical Pressurized Water Reactor rod is analyzed. • Steady state fuel rod behavior is examined to see the effects of pellet to cladding gap thickness and gap gas pressure. • Transient fuel rod behavior is examined to see the effects of pellet to cladding gap thickness and gap gas pressure. • The optimum pellet to cladding gap thickness and gap gas pressure values of the simulated fuel are determined. • The effects of pellet to cladding gap design parameters on nuclear fuel reliability are examined. - Abstract: As an important improvement in the light water nuclear reactor operations, the nuclear fuel burnup rate is increased in recent decades and this increase causes heavier duty for the nuclear fuel. Since the high burnup fuel is exposed to very high thermal and mechanical stresses and since it operates in an environment with high radiation for about 18 month cycles, it carries the risk of losing its integrity. In this study; it is aimed to determine the effects of pellet–cladding gap thickness and gap pressure on reliability of high burnup nuclear fuel in Pressurized Water Reactors (PWRs) under steady state operation conditions and suggest optimum values for the examined parameters only and validate these suggestions for a transient condition. In the presented study, fuel performance was analyzed by examining the effects of pellet–cladding gap thickness and gap pressure on the integrity of high burnup fuels. This work is carried out for a typical Westinghouse type PWR fuel. The steady state conditions were modeled and simulated with FRAPCON-3.4a steady state fuel performance code and the FRAPTRAN-1.4 fuel transient code was used to calculate transient fuel behavior. The analysis included the changes in the important nuclear fuel design limitations such as the centerline temperature, cladding stress, strain and oxidation with the change in pellet–cladding gap thickness and initial pellet–cladding gap gas

  13. 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

  14. In-Situ Observation of Sintering Shrinkage of UO2 Compacts Derived from Different Powder Routes

    International Nuclear Information System (INIS)

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

    2015-01-01

    In-situ observations on the shrinkage of green pellets with precisely controlled dimensions were carefully conducted by using TOM during H2 atmosphere sintering. The shrinkage retardation in IDR-UO 2 might be attributed to the larger primary particle size of IDRUO 2 than those of ADU- and AUC- UO 2 powders. It would be important to understand the different sintering characteristics of UO 2 powders according to the powder routes, when it comes to designing a new sintering process or choosing a sintering additive for new fuel pellet like PCI (Pellet Cladding Interaction) remedy pellet. In this paper, we have investigated the initial and intermediate sintering shrinkage of UO 2 from different powder routes by in-situ observation of green samples during H2 atmosphere sintering. Effect of powder characteristics of three different UO 2 powders on the initial and intermediate sintering were closely reviewed including crystal structure, powder size, specific surface area, primary crystal size, and O/U ratio

  15. 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

  16. 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

  17. 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

  18. 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)

  19. 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.

  20. 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)

  1. Westinghouse Advanced Doped Pellet - Characteristics and irradiation behavior

    International Nuclear Information System (INIS)

    Backman, K.; Hallstadius, L.; Roennberg, G.

    2009-01-01

    Full text: There are a number of trends in the nuclear power industry, which put additional requirements on the operational flexibility and reliability of nuclear fuel, for example power uprates and longer cycles in order to increase production, higher burnup levels in order to reduce the backend cost of the fuel cycle, and lower goals for activity release from power plant operation. These additional requirements can be addressed by increasing the fuel density, improving the FG retention, improving the PCI resistance and improving the post-failure performance. In order to achieve that, Westinghouse has developed ADOPT (Advanced Doped Pellet Technology) UO 2 fuel containing additions of chromium and aluminium oxides. The additives facilitate pellet densification during sintering, enlarge the pellet grain size, and increase the creep rate. The final manufactured doped pellets reach about 0.5 % higher density within a shorter sintering time and a five times larger grain size compared with standard UO 2 fuel pellets. Fuel rods with ADOPT pellets have been irradiated in several light water reactors (LWRs) since 1999, including two full SVEA Optima2 reloads in 2005. ADOPT pellets has been investigated in pool-side and hot cell Post Irradiation Examinations (PIEs), as well as in a ramp test and a fuel washout test in the Studsvik R2 test reactor. The investigations have identified three areas of improved operational behaviour: Reduced Fission Gas Release (FGR), improved Pellet Cladding Interaction (PCI) performance thanks to increased pellet plasticity and higher resistance against post-failure degradation. The better FGR behaviour of ADOPT has been verified with a pool side FGR gamma measurement performed at 55 MWd/kgU, as well as transient tests in the Studsvik R2 reactor. Creep measurements performed on fresh pellets show that ADOPT has a higher creep rate which is beneficial for the PCI performance. ADOPT has also been part of a high power Halden test (IFA-677). The

  2. High frequency acoustic microscopy for the determination of porosity and Young's modulus in high burnup uranium dioxide nuclear fuel

    International Nuclear Information System (INIS)

    Marchetti, M.; Laux, D.; Cappia, F.; Laurie, M.; Van Uffelen, P.; Rondinella, V.V.; Despaux, G.

    2015-01-01

    During irradiation UO 2 nuclear fuel experiences the development of a non-uniform distribution of porosity which contributes to establish varying mechanical properties along the radius of the pellet. Radial variations of the porosity and of elastic properties in high burnup UO 2 pellet can be investigated via high frequency acoustic microscopy. Ultrasound waves are generated by a piezoelectric transducer and focused on the sample, after having travelled through a coupling liquid. The elastic properties of the material are related to the velocity of the generated Rayleigh surface wave (VR). A 67 MWd/kgU UO 2 pellet was characterized using the acoustic microscope installed in the hot cells of the Institute of Transuranium Elements: 90 MHz frequency was applied, methanol was used as coupling liquid and VR was measured at different radial positions. By comparing the porosity values obtained via acoustic microscopy with those determined using ceramographic image analysis a good agreement was found, especially in the areas close to the centre. In addition Young's modulus was calculated and its radial profile was correlated to the corresponding burnup profile. (authors)

  3. Creep of UO2 at 25000C

    International Nuclear Information System (INIS)

    Slagle, O.D.

    1977-01-01

    Until an improved high temperature relationship is available, the previously derived low temperature relationship is a reasonable means for predicting the creep rates of UO 2 at 2500 0 C. The activation energy determined for creep at 2500 0 C is at least two times larger than that measured previously at the lower temperature. Stress induced grain growth under uniaxial compression at high temperatures in UO 2 results in preferential growth of grains having a cube axis closely aligned with the stress axis

  4. Innovative microstructures in ThO2-UO2 system

    International Nuclear Information System (INIS)

    Kutty, T.R.G.; Sengupta, A.K.; Majumdar, S.; Sah, D.N.; Kamath, H.S.

    2005-01-01

    The basic properties that really matter to the nuclear scientists are those that have greatest influence on microstructure: crystal structure, defects concentration and phase stability. The role of microstructure and crystal defects in determining the engineering properties are always acknowledged. Microstructure of nuclear fuels controls the in-pile fuel behavior like fission gas release, plasticity, in-pile creep and swelling. Conventional nuclear ceramic fabrication process consists of a number of stages, including calcination, milling, incorporating additives, pressing, drying and densification. Since each of these steps affects the microstructure of fuel pellets they must all be understood and a more holistic approach is required when processing nuclear ceramics compared to metals and polymers. It is possible to obtain a wide range of microstructures for ThO 2 -UO 2 system if a proper fabrication route is chosen. It is possible to tailor microstructure as per our requirement so that an improved behaviour during irradiation is expected. The improvement in plasticity and fission gas release can be attained by modifying the microstructure during fabrication. This paper deals with fabrication of ThO 2 -UO 2 pellets of varying U content and its characterization with the help of optical microscopy, XRD, SEM and EPMA. The microstructures are characterized in terms grain size, pore size and its distribution and homogeneity of uranium. (author)

  5. Acoustic emission during the compaction of brittle UO2 particles

    International Nuclear Information System (INIS)

    Hegron, Lise

    2014-01-01

    One of the options considered for recycling minor actinides is to incorporate about 10% to UO 2 matrix. The presence of open pores interconnected within this fuel should allow the evacuation of helium and fission gases to prevent swelling of the pellet and ultimately its interaction with the fuel clad surrounding it. Implementation of minor actinides requires working in shielded cell, reducing their retention and outlawing additions of organic products. The use of fragmentable particles of several hundred micrometers seems a good solution to control the microstructure of the green compacts and thus control the open porosity after sintering. The goal of this study is to monitor the compaction of brittle UO 2 particles by acoustic emission and to link the particle characteristics to the open porosity obtained after the compact sintering. The signals acquired during tensile strength tests on individual granules and compacts show that the acoustic emission allows the detection of the mechanism of fragmentation and enables identification of a characteristic waveform of this fragmentation. The influences of compaction stress, of the initial particle size distribution and of the internal cohesion of the granules, on the mechanical strength of the compact and on the microstructure and open porosity of the sintered pellets, are analyzed. By its ability to identify the range of fragmentation of the granules during compaction, acoustic emission appears as a promising technique for monitoring the compaction of brittle particles in the manufacture of a controlled porosity fuel. (author) [fr

  6. Development of UO2/PuO2 dispersed in uranium matrix CERMET fuel system for fast reactors

    International Nuclear Information System (INIS)

    Sinha, V.P.; Hegde, P.V.; Prasad, G.J.; Pal, S.; Mishra, G.P.

    2012-01-01

    CERMET fuel with either PuO 2 or enriched UO 2 dispersed in uranium metal matrix has a strong potential of becoming a fuel for the liquid metal cooled fast breeder reactors (LMR’s). In fact it may act as a bridge between the advantages and disadvantages associated with the two extremes of fuel systems (i.e. ceramic fuel and metallic fuel) for fast reactors. At Bhabha Atomic Research Centre (BARC), R and D efforts are on to develop this CERMET fuel by powder metallurgy route. This paper describes the development of flow sheet for preparation of UO 2 dispersed in uranium metal matrix pellets for three different compositions i.e. U–20 wt%UO 2 , U–25 wt%UO 2 and U–30 wt%UO 2 . It was found that the sintered pellets were having excellent integrity and their linear mass was higher than that of carbide fuel pellets used in Fast Breeder Test Reactor programme (FBTR) in India. The pellets were characterized by X-ray diffraction (XRD) technique for phase analysis and lattice parameter determination. The optical microstructures were developed and reported for all the three different U–UO 2 compositions.

  7. Development of UO2/PuO2 dispersed in uranium matrix CERMET fuel system for fast reactors

    Science.gov (United States)

    Sinha, V. P.; Hegde, P. V.; Prasad, G. J.; Pal, S.; Mishra, G. P.

    2012-08-01

    CERMET fuel with either PuO2 or enriched UO2 dispersed in uranium metal matrix has a strong potential of becoming a fuel for the liquid metal cooled fast breeder reactors (LMR's). In fact it may act as a bridge between the advantages and disadvantages associated with the two extremes of fuel systems (i.e. ceramic fuel and metallic fuel) for fast reactors. At Bhabha Atomic Research Centre (BARC), R & D efforts are on to develop this CERMET fuel by powder metallurgy route. This paper describes the development of flow sheet for preparation of UO2 dispersed in uranium metal matrix pellets for three different compositions i.e. U-20 wt%UO2, U-25 wt%UO2 and U-30 wt%UO2. It was found that the sintered pellets were having excellent integrity and their linear mass was higher than that of carbide fuel pellets used in Fast Breeder Test Reactor programme (FBTR) in India. The pellets were characterized by X-ray diffraction (XRD) technique for phase analysis and lattice parameter determination. The optical microstructures were developed and reported for all the three different U-UO2 compositions.

  8. Statistical model for grain boundary and grain volume oxidation kinetics in UO2 spent fuel

    International Nuclear Information System (INIS)

    Stout, R.B.; Shaw, H.F.; Einziger, R.E.

    1989-09-01

    This paper addresses statistical characteristics for the simplest case of grain boundary/grain volume oxidation kinetics of UO 2 to U 3 O 7 for a fragment of a spent fuel pellet. It also presents a limited discussion of future extensions to this simple case to represent the more complex cases of oxidation kinetics in spent fuels. 17 refs., 1 fig

  9. The effect Of Fine UO2 Powder Composition On Density Of UO2 pellet Compaction

    International Nuclear Information System (INIS)

    Widjaksana; Mutiara, E.

    1998-01-01

    This research has been done to solve it. The work was done through making variation on the fine powder composition from 0% up to 50% and variation on compacting pressure from 1 MP up to 5 MP. Each variation combination was replicated eight times. The result showed that the data were within confidence interval of the density and the ANOVA showed that the parameters affected the density. The affect of them was showed in the various correlation. The maximum density was reached at the range of fine powder composition between 20% and 30%. When Heckel theory was applied to the data, the conclusion showed that the fine powder composition did not affect the mechanical strength of material, but when Smith theory was applied, the conclusion showed that the fine powder composition affect the compressibility of powder. The maximum compressibility of powder was reached at the range of fine powder composition between 20% and 30%

  10. Estimate of the instant release fraction for UO2 and MOX fuel at t=0

    International Nuclear Information System (INIS)

    Johnson, L.; Poinssot, C; Ferry, C.; Lovera, P.

    2004-07-01

    values, which results in significant overprediction of average IRF values. Best estimate IRF values are determined for moderate burnup UO 2 fuel for nuclides for which data exist, because the understanding and data is sufficient. Only pessimistic IRF values are estimated for radionuclides for which little data is available and in the case of MOX fuel and higher burnup UO 2 fuel. Special attention is given to several phenomena occurring in the outer region of fuel pellets (rim region) resulting in restructuring of fuel grains. These include: a) high fission density as a result of high yields of 239 Pu arising from capture of epithermal neutrons; b) increased porosity; c) reduction in grain size; d) increased thermal release of fission gas from the grains. From the perspective of assessing the release of fission products from spent fuel under disposal conditions, the restructuring process is important

  11. Development of advanced LWR fuel pellet technology

    International Nuclear Information System (INIS)

    Song, Kun Woo; Kang, K.W.; Kim, K. S.; Yang, J. H.; Kim, Y. M.; Kim, J. H.; Bang, J. B.; Kim, D. H.; Bae, S. O.; Jung, Y. H.; Lee, Y. S.; Kim, B. G.; Kim, S. H.

    2000-03-01

    A UO 2 pellet was designed to have a grain size of larger than 12 μm, and a new duplex design that UO 2 -Gd 2 O 3 is in the core and UO 2 -Er 2 O 3 in the periphery was proposed. A master mixing method was developed to make a uniform mixture of UO 2 and additives. The open porosity of UO 2 pellet was reduced by only mixing AUC-UO 2 powder with ADU-UO 2 or milled powder. Duplex compaction tools (die and punch) were designed and fabricated, and duplex compacting procedures were developed to fabricate the duplex BA pellet. In UO 2 sintering, the relations between sintering variables (additive, sintering gas, sintering temperature) and pellet properties (density, grain size, pore size) were experimentally found. The UO 2 -U 3 O 8 powder which is inherently not sinterable to high density could be sintered well with the aid of additives. U 3 O 8 single crystals were added to UO 2 powder, and homogeneous powder mixture was pressed and sintered in a reducing atmosphere. This technology leads to a large-grained pellet of 12-20 μm. In UO 2 -Gd 2 O 3 sintering, the relations between sintering variables (additives, sintering gas) and pellet properties (density, grain size) were experimentally found. The developed technology of fabricating a large-grained UO 2 pellet has been optimized in a lab scale. Pellet properties were investigated in the fields of (1) creep properties, (2) thermal properties, (3) O/M ratios and (4) unit cell lattice. (author)

  12. An Analysis of the Thermal and Structure Behaviour of the UO2-PuO2-Fuel in the Irradiation Experiment of the UO2-PuO2-Fuel in the Irradiation Experiment FR2 Capsule Test Series 5a

    International Nuclear Information System (INIS)

    Lopez Jimenez, J.; Helmut, E.

    1981-01-01

    In the Karlsruhe research reactor FR2 nine fuel pins were irradiated within three irradiation capsules in the course of the test series 5a. The pins contained UO 2 -PuO 2 fuel pellets. They reached bump values of about 6, 17 and 47 Mwd/Kg Me with linear rod powers of 400 to 600 W/cm and clad surface temperature between 500 and 700 degree centigree. A detailed analysis of the fuel structuration data (columnar-grain and equiaxed- -grain growth regions) have allowed to determine, with the help of physic-mathematical models, the radii of these regions and the heat transfer through the contact zone between fuel and clad depending on the bump. The results of the analysis showed that the fuel surface temperature rose with increasing burnup. (Author) 16 refs

  13. 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)

  14. Fabrication and post-irradiation examination of a zircaloy-2 clad UO2-1.5 wt% PuO2 fuel pin irradiated in PWL, CIRUS

    International Nuclear Information System (INIS)

    Sah, D.N.; Sahoo, K.C.; Chatterjee, S.; Majumdar, S.; Kamath, H.S.; Ramachandran, R.; Bahl, J.K.; Purushottam, D.S.C.; Ramakumar, M.S.; Sivaramakrishnan, K.S.; Roy, P.R.

    1977-01-01

    A zircaloy-2 clad UO 2 -1.5 wt% PuO 2 fuel pin was fabricated at the Radiometallurgy Section of the Bhabha Atomic Research Centre, Bombay, for irradiation in the pressurised water loop in CIRUS. Requisite development work related to powder conditioning, blending, pressing and sintering parameters was carried out to meet the exacting fuel pellet specifications of CANDU fuel. The fuel pin ruptured while being irradiated in the pressurised water loop in CIRUS, after experiencing a low burn-up of 507 MWD/MTM and was subsequently examined at the Radiometallurgy Hot Cells Facility. The results showed that internal clad hydriding led to primary failure of the fuel pin. Subsequent ingress of the coolant water caused excessive swelling of the thermal insulating magnesia pellets located at the ends of the fuel column. The swelling of magnesia pellets caused severe rupturing of the fuel pin at the two ends. The delayed rupturing of the fuel pin at the upper end, caused the fuel column to be displaced downwards by 5.85mm. (author)

  15. Boiling point measurements on liquid UO2

    International Nuclear Information System (INIS)

    Bober, M.; Singer, J.; Trapp, M.

    1986-01-01

    In analogy to the classic boiling point method, a quasi-stationary millisecond laser-heating technique was applied to measure the saturated-vapour pressure curve of liquid UO 2 in the temperature range of 3500 to 4500 K. The result is represented by log p(MPa) 5.049 -23042/T(K) according to an average heat of vaporization of 441 kJ/mol and a normal boiling point of 3808 K. Besides, spectral emissivities of liquid UO 2 were measured at the pyrometer wavelengths of 752 and 1064 nm. (author)

  16. 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

  17. 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)

  18. The corrosion of spent UO2 fuel in synthetic groundwater

    International Nuclear Information System (INIS)

    Forsyth, R.S.; Werme, L.D.; Bruno, J.

    1985-10-01

    Leaching of high burnup BWR fuel for up to 3 years showed that both U and Pu attain saturation rapidly at pH 8.1, giving values of 1-2 mg/l and 1 μg/l respectively. The leaching rate for Sr-90 decreased from about 10 -5 /d to 10 -7 /d but was always higher than the rates for U, Pu, Cm, Ce, Eu and Ru. Congruent dissolution was only attained at pH values of about 4. When reducing conditions were imposed on the pH 8.1 groundwater by means of H 2 /Ar in the presence of a Pd catalyst, significanly lower leach rates were attained. The hypothesis that alpha radiolytic decomposition of water is a driving force for UO 2 corrosion even under reducing conditions has been examined in leaching tests on low burnup (low alpha dose-rate) fuel. No significant effect of alpha radiolysis under the experimental conditions was detected. Thermodynamically the calculated uranium solubilities in the pH range 4-8.2 generally agreed, well with the measured ones, although assumptions made for certain parameters in the calculations limit the validity of the results. (Author)

  19. Developments in MOX fuel pellet fabrication technology: Indian experience

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  20. Study of uranium dioxide pellets by micro-acoustic techniques

    International Nuclear Information System (INIS)

    Roque, V.

    1999-01-01

    In order to reduce the volume of spent fuel to reprocess and to improve the productivity and the safety of the nuclear reactor, 'Electricite De France' aim to increase the average fuel discharge burn-up. To elaborate the safety reports, EDF develops codes to simulate the thermo-mechanical behaviour of the nuclear fuel element. These numeric simulations need to evaluate accurately and locally the evolution of the material and of its properties. One of the major concern today is the local characterisation of the intrinsic volume fraction porosity and the mechanical properties of the irradiated fuel. The fuel pellet fragmentation, the steep radial gradient in its physical properties evolution and the chemical evolution of the irradiated material make difficult nay the use of the conventional techniques. This leads EDF to pay interest for the use of two complementary techniques: micro-indentation on the one hand and acoustic methods on the other hand (acoustic microscopy and micro-echography), with an additional constrain to perform on active materials. The objective of this work has been to adapt the acoustic methods for an application on uranium dioxide pellets, used as nuclear fuel in Water Pressurised Reactor. Acquisitions protocols have been set to measure accurately the Rayleigh velocity and the longitudinal velocity of the UO 2 . Using these protocols, we have calibrated these acoustic methods by analysing non irradiated nuclear pellet which properties were well known. This process enable to quantify the effects of different physico-chemical parameters of the UO 2 on the ultrasonic velocities measured. Particularly, the large influence of the porosity has been demonstrated and empirical laws to express the evolution of the acoustic velocities as a function of the volume fraction porosity were established. Moreover, we have established a methodology to characterise the intrinsic elastic constants and the volume fraction porosity on irradiated UO 2 fuel pellets

  1. Simulation of the behaviour of nuclear fuel under high burnup conditions

    International Nuclear Information System (INIS)

    Soba, Alejandro; Lemes, Martin; González, Martin Emilio; Denis, Alicia; Romero, Luis

    2014-01-01

    Highlights: • Increasing the time of nuclear fuel into reactor generates high burnup structure. • We analyze model to simulate high burnup scenarios for UO 2 nuclear fuel. • We include these models in the DIONISIO 2.0 code. • Tests of our models are in very good agreement with experimental data. • We extend the range of predictability of our code up to 60 MWd/KgU average. - Abstract: In this paper we summarize all the models included in the latest version of the DIONISIO code related to the high burnup scenario. Due to the extension of nuclear fuels permanence under irradiation, physical and chemical modifications are developed in the fuel material, especially in the external corona of the pellet. The codes devoted to simulation of the rod behaviour under irradiation need to introduce modifications and new models in order to describe those phenomena and be capable to predict the behaviour in all the range of a general pressurized water reactor. A complex group of subroutines has been included in the code in order to predict the radial distribution of power density, burnup, concentration of diverse nuclides and porosity within the pellet. The behaviour of gadolinium as burnable poison also is modelled into the code. The results of some of the simulations performed with DIONISIO are presented to show the good agreement with the data selected for the FUMEX I/II/III exercises, compiled in the NEA data bank

  2. Modeling of WWER-440 Fuel Pin Behavior at Extended Burn-up

    International Nuclear Information System (INIS)

    El-Koliel, M.S.; Abou-Zaid, A.A.; El-Kafas, A.A.

    2004-01-01

    Currently, there is an ongoing effort to increase fuel discharge burn-up of all LWRs fuel including WWER's as much as possible in order to decrease power production cost. Therefore, burn-up is expected to be increased to 60 to 70 Mwd/kg U. The change in the fuel radial power distribution as a function of fuel burn up can affect the radial fuel temperature distribution as well as the fuel microstructure in the fuel pellet rim. In this paper, the radial burn-up and fissile products distributions of WWER-440 UO 2 fuel pin were evaluated using MCNP 4B and ORIGEN2 codes. The impact of the thermal conductivity on predicted fission gas release calculations is needed. For the analysis, a typical WWER-440 fuel pin and surrounding water moderator are considered in a hexagonal pin cell well. The thermal release and the athermal release from the pellet rim were modeled separately. The fraction of the rim structure and the excessive porosity in the rim structure in isothermal irradiation as a function of the fuel burn-up was predicted. a computer program; RIMSC-01, is developed to perform the required FGR calculations. Finally, the relevant phenomena and the corresponding models together with their validation are presented

  3. Mechanical properties and structure of Zircaloy attached by UO2+x and fission products

    International Nuclear Information System (INIS)

    Holub, F.

    1987-08-01

    The aim of this project was to determine the combined long-term effect of simulated fission products and hyperstoichiometric uranium dioxide on the mechanical properties and structure of Zircaloy. Three groups of fission product elements or compounds were defined: The rare earth oxides CeO 2 , La 2 O 3 , Nd 2 O 3 , Y 2 O 3 ; The metals No, Ru, Ag; The low melting elements Te, Sb and Cd. Each of these groups of fission products was mixed with UO 2+x in proportion related for burnups of 5, 10 and 30%. The simulated fuel mixtures were filled into tubular Zircaloy casings, plugged and welded. These specimens were annealed at 350, 500 and 700 deg. C up to 17,500 hours. The test results indicate different kinds of action of the simulated fuel constituents. Mixtures of rare earth oxides and UO 2+x embrittle Zircaloy drastically at higher temperatures. There exists a mutual intensifying effect of rare earth oxides and UO 2+x . UO 2+x and (Mo + Ru + Ag) and their mixtures act very similar on Zircaloy. The low melting fission products (Te + Sb + Cd) influence the ductility of Zircaloy in an advantageous manner, compared to pure UO 2+x fuel. The layer of zirconium tellurides seems to protect the Zircaloy metal against the embrittling attack of oxygen from UO 2+x . The most important events of tensile tests at 400 deg. C are the high values of the elongation of specimens which are brittled at room temperature. It should guarantee the integrity of fuel elements, which have been attacked chemically by fission products at temperatures of 400 deg. C and higher

  4. Equi-axed and columnar grain growth in UO2

    International Nuclear Information System (INIS)

    White, R.J.

    1997-01-01

    The grain size of UO 2 is an important parameter in the actual performance and the modelling of the performance of reactor fuel elements. Many processes depend critically on the grain size, for example, the degree of initial densification, the evolution rate of stable fission gases, the release rates of radiologically hazardous fission products, the fission gas bubble swelling rates and the fuel creep. Many of these processes are thermally activated and further impact on the fuel thermal behavior thus creating complex feedback processes. In order to model the fuel performance accurately it is necessary to model the evolution of the fuel grain radius. When UO 2 is irradiated, the fission gases xenon and krypton are created from the fissioning uranium nucleus. At high temperatures these gases diffuse rapidly to the grain boundaries where they nucleate immobile lenticular shaped fission gas bubbles. In this paper the Hillert grain growth model is adapted to account for the inhibiting ''Zener'' effects of grain boundary fission gas porosity on grain boundary mobility and hence grain growth. It is shown that normal grain growth ceases at relatively low levels of irradiation. At high burnups, high temperatures and in regions of high temperature gradients, columnar grain growth is often observed, in some cases extending over more than fifty percent of the fuel radius. The model is further extended to account for the de-pinning of grains in the radial direction by the thermal gradient induced force on a fission gas grain boundary bubble. The observed columnar/equi-axed boundary is in fair agreement with the predictions of an evaporation/condensation model. The grain growth model described in this paper requires information concerning the scale of grain boundary porosity, the local fuel temperature and the local temperature gradient. The model is currently used in the Nuclear Electric version of the ENIGMA fuel modelling code. (author). 14 refs, 3 figs, 1 tab

  5. Irradiation of UO2 specimens with molten cores in a pressurized water loop. Test X-2-x

    International Nuclear Information System (INIS)

    Bain, A.S.

    1961-08-01

    Two Zircaloy-2 clad specimens containing stoichiometric UO 2 pellets were irradiated in a pressurized water loop for 379 hours at heat ratings sufficient to cause central melting of the UO 2 . There was no appearance of localized overheating or accelerated corrosion of the sheath, but the diametral increases were considerably larger than those observed in loop specimens irradiated at lower heat ratings. The length increases, however, were approximately the same as those measured for specimens at lower ratings. There was a clearly visible demarcation between UO 2 that had been molten and that which had not. The value of ∫ 500 o C Tm kdθ = 74 ± W/cm was essentially the same as that obtained from the short-duration tests in the Hydraulic Rabbit, indicating there is no marked decrease in thermal conductivity of the UO 2 fuel in irradiations up to 379 hours. (author)

  6. TCA UO2/MOX core analyses

    International Nuclear Information System (INIS)

    Tahara, Yoshihisa; Noda, Hideyuki

    2000-01-01

    In order to examine the adequacy of nuclear data, the TCA UO 2 and MOX core experiments were analyzed with MVP using the libraries based on ENDF/B-VI Mod.3 and JENDL-3.2. The ENDF/B-VI data underpredict k eff values. The replacement of 238 U data with the JENDL-3.2 data and the adjustment of 235 ν-value raise the k eff values by 0.3% for UO 2 cores, but still underpredict k eff values. On the other hand, the nuclear data of JENDL-3.2 for H, O, Al, 238 U and 235 U of ENDF/B-VI whose 235 ν-value in thermal energy region is adjusted to the average value of JENDL-3.2 give a good prediction of k eff . (author)

  7. Leaching of irradiated CANDU UO2 fuel

    International Nuclear Information System (INIS)

    Vandergraaf, T.T.; Johnson, L.H.; Lau, D.W.P.

    1980-01-01

    Irradiated fuel, leached at room temperature with distilled water and with slightly chlorinated river water, releases approx. 4% of its cesium inventory over a comparatively sort period of a few days but releases its actinides and rare earths more slowly. The matrix itself dissolves at a rate conservatively calculated to be less than approx. 2 x 10 -6 g UO 2 /cm 2 day and, with time, the leach rates of the various nuclides approach this value

  8. UO2 production process with methanol washing

    International Nuclear Information System (INIS)

    Sondermann, T.

    1978-01-01

    The invention refers to a process for the recovery of methanol used for washing the ammonium uranyl carbonate obtained during UO 2 production. The methanol contains about 50% H 2 O, about 10% (NH 4 ) 2 CO 3 , and is radioactive. According to the invention the methanol is purified at reduced pressure in a distillation unit and then led back to the washing unit. (UWI) 891 HP/UWI 892 MBE [de

  9. Crystal-field effect in UO2

    International Nuclear Information System (INIS)

    Gajek, Z.; Lahalle, M.P.; Krupa, J.C.; Mulak, J.

    1988-01-01

    Simple ab initio model perturbation calculations of the crystal-field parameters for the U 4+ ion in UO 2 crystals are reported. The crystal-field parameters obtained, B 0 4 = -7130 cm -1 and B 0 6 = 2890 cm -1 , turn out to be much lower in value, particularly the first one, than those usually assumed for this compound. They are found, however, to agree with new spectroscopic data and recent inelastic neutron scattering measurements. (orig.)

  10. Thermal stress in UO2 during sintering as a possible cause of cracking

    International Nuclear Information System (INIS)

    Aragones, M.A.; Tobias, E.; Tulli, I.; Naquid, C.

    1980-01-01

    Thermal stresses arising during sintering of UO 2 pellets are evaluated numerically by the solution of coupled equations for heat transfer through the sample. Results are compared with those of a semiempirical approach reported in the literature. Better insight into the heat transfer process is obtained from the solution of the coupled equations rather than from the empirical approach. The two approaches give different results for the thermal stresses arising during sintering. The use of heating and cooling rates of approximately 0.5 0 Cs -1 is found to prevent the possibility of cracking in UO 2 pellets of radii varying from 0.6 cm to 1 cm during sintering in hydrogen or argon-hydrogen atmospheres. (author)

  11. 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)

  12. Preliminary study of cost benefits associated with duplex fuel pellets of the LOWI type

    International Nuclear Information System (INIS)

    Ainscough, J.B.; Coucill, D.N.; Howl, D.A.; Jensen, A.; Misfeldt, I.

    1983-01-01

    Duplex UO 2 pellets, which consist of an outer enriched annulus and a depleted or natural core, can provide a solution to the problem of stress corrosion cracking failures, which have led to constraints being placed on ramp rates in power reactors. An analysis of the reactor physics and the performance of duplex pellets is presented in the context of a 17 X 17 pressurized water reactor fuel rod design. The study has been based on the particular type of duplex pellet in which the core and the annulus are physically separate; this is called ''LOWI'' after the Danish design. At low burnup, this fuel shows a significant improvement in power ramp performance compared with standard fuel. At higher burnup, the benefits are less certain but as the severity of the ramp will usually be less in high burnup fuel simply because of the reduced rating, the reduction in benefit may not be significant. If the gap between the core and annulus persists to high burnup, there will be no loss of benefit. Economic calculations and a cost-benefit analysis are presented to show the number of extra full-power hours of reactor operation that must be obtained in order to outweigh the additional fabrication costs associated with this fuel

  13. 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)

  14. 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

  15. Production of molten UO2 pools by internal heating: apparatus and preliminary experimental heat transfer results

    International Nuclear Information System (INIS)

    Chasanov, M.G.; Gunther, W.H.; Baker, L. Jr.

    1977-01-01

    The capability for removal of heat from a pool of molten fuel under postaccident conditions is an important consideration in liquid-metal fast breeder reactor safety analysis. No experimental data for pool heat transfer from molten UO 2 under conditions simulating internal heat generation by fission product decay have been reported previously in the literature. An apparatus to provide such data was developed and used to investigate heat transfer from pools containing up to 7.5 kg of UO 2 ; the internal heat generation rates and pool depths attained cover most of the ranges of interest for postaccident heat removal analysis. It was also observed in these studies that the presence of simulated fission products corresponding to approximately 150,000 kW-day/kg burnup had no significant effect on the observed heat transfer

  16. High burn-up structure in nuclear fuel: impact on fuel behavior - 4005

    International Nuclear Information System (INIS)

    Noirot, J.; Pontillon, Y.; Zacharie-Aubrun, I.; Hanifi, K.; Bienvenu, P.; Lamontagne, J.; Desgranges, L.

    2016-01-01

    When UO 2 and (U,Pu)O 2 fuels locally reach high burn-up, a major change in the microstructure takes place. The initial grains are replaced by thousands of much smaller grains, fission gases form micrometric bubbles and metallic fission products form precipitates. This occurs typically at the rim of the pellets and in heterogeneous MOX fuel Pu rich agglomerates. The high burn-up at the rim of the pellets is due to a high capture of epithermal neutrons by 238 U leading locally to a higher concentration of fissile Pu than in the rest of the pellet. In the heterogeneous MOX fuels, this rim effect is also active, but most of the high burn-up structure (HBS) formation is linked to the high local concentration of fissile Pu in the Pu agglomerates. This Pu distribution leads to sharp borders between HBS and non-HBS areas. It has been shown that the size of the new grains, of the bubbles and of the precipitates increase with the irradiation local temperatures. Other parameters have been shown to have an influence on the HBS initiation threshold, such as the irradiation density rate, the fuel composition with an effect of the Pu presence, but also of the Gd concentration in poisoned fuels, some of the studied additives, like Cr, and, maybe some of the impurities. It has been shown by indirect and direct approaches that HBS formation is not the main contributor to the increase of fission gas release at high burn-up and that the HBS areas are not the main source of the released gases. The impact of HBS on the fuel behavior during ramp on high burn-up fuels is still unclear. This short paper is followed by the slides of the presentation

  17. SEM hot stage sintering of UO2

    International Nuclear Information System (INIS)

    Miller, D.J.

    1976-06-01

    The sintering of hyperstoichiometric uranium dioxide powder compacts, in the hot stage of a scanning electron microscope, was continuously monitored using 16 mm time lapse movies. From alumina microspheres placed on the surface of the compacts, shrinkage measurements were obtained. Converting shrinkage measurements into densification profiles indicates that a maximum densification rate is reached at a critical density, independent of the constant heating rates. At temperatures above 1350 0 C, the movement of the reference microspheres made shrinkage measurements impossible. It is believed the evolution of UO 3 gas from hyperstoichiometric UO 2 is the cause of this limitation

  18. Modelling of UO2 oxidation in steam

    International Nuclear Information System (INIS)

    Brito, A.C.; Iglesias, F.C.; Liu, Y.

    1996-01-01

    A computer model has been developed for calculating oxidation of UO 2 at high temperatures in steam oxidising conditions. Several methods to calculate the partial pressure of oxygen in the fuel and in the environment surrounding the fuel are available. The various methodologies have been compared and the best models have been compiled into a computer model which will be implemented into fuel thermal/mechanical behaviour codes such as FACTAR 2.0 (LOECI) and ELESIM/ELOCA. Calculations from the computer model have been compared to experimental results. The calculated oxidation reaction kinetics are in good agreement with the experimental data. (author)

  19. Physics of the fuel cycle. Evaluation of methods, uncertainties and estimation of the material balance for fuels UO2 and UO2-PuO2

    International Nuclear Information System (INIS)

    Monier, C.

    1997-09-01

    The research works of this thesis are aimed to evaluate the methods and the associated uncertainties for the material balances estimation of the burn-up UO 2 and MOX fuels which intervene in the fuel cycle physics. The studies carried out are used to qualify the cycle 'package' DARWIN for the PWRs material balances estimation. The elaboration and optimisation of the calculation routes are carried out following a very specific methodology, aimed at estimating the bias introduced by the modelizations simplification by a comparison with almost exact reference modelizations. Depending on the precision goals and the informations, the permissible approximation will be determined. Two calculation routes have been developed and the qualified by applying them to the used fuels isotopic analysis interpretation: one 'industry-oriented' calculation route which can calculate full UO 2 assemblies material balances with a 2 % precision on the main actinides, respecting the industrial specifications. This route must run with a reasonable calculation time and stay user-friendly; one reference calculation route for the precise interpretation of fuel samples made of pieces of burn-up MOX rods. Aiming to provide material balances with the best possible precision, this route does not have the same specifications concerning its use and its calculation time performance. (author)

  20. Out-of-pile UO2/Zircaloy-4 experiments under severe fuel damage conditions

    International Nuclear Information System (INIS)

    Hofmann, P.

    1983-01-01

    Chemical interactions between UO 2 fuel and Zircaloy-4 cladding up to the melting point of zircaloy (Zry) are described. Out-of-pile UO 2 /zircaloy reaction experiments have been performed to investigate the chemical interaction behavior under possible severe fuel damage conditions (very high temperatures and external overpressure). The tests have been conducted in inert gas (1 to 80 bar) with 10-cm-long zircaloy cladding specimens filled with UO 2 pellets. The annealing temperature varied between 1000 and 1700 deg. C and the annealing period between 1 and 150 min. The extent of the chemical reaction depends decisively on whether or not good contact between UO 2 and zircaloy has been established. If solid contact exists, zircaloy reduces the UO 2 to form oxygen-stabilized α-Zr(O) and uranium metal. The uranium reacts with zircaloy to form a (U,Zr) alloy rich in uranium. The (U,Zr) alloy, which is liquid above approx. 1150 deg. C, lies between two α-Zr(O) layers. The UO 2 /zircaloy reaction obeys a parabolic rate law. The degree of chemical interaction is determined by the extent of oxygen diffusion into the cladding, and hence by the time and temperature. The affinity of zirconium for oxygen, which results in an oxygen gradient across the cladding, is the driving force for the reaction. The growth of the reaction layers can be represented in an Arrhenius diagram. The UO 2 /Zry-4 reaction occurs as rapidly as the steam/Zry-4 reaction above about 1100 deg. C. The extent of the interaction is independent of external pressure above about 10 bar at 1400 deg. C and 5 bar at 1700 deg. C. The maximum measured oxygen content of the cladding is approx. 6wt.%. Up to approx. 9 volume % of the UO 2 can be chemically dissolved by the zircaloy. In an actual fuel rod, complete release of the fission products in this region of the fuel must therefore be assumed. (author)

  1. 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)

  2. Modelling of thermal mechanical behaviour of high burn-Up VVER fuel at power transients with special emphasis on the impact of fission gas induced swelling of fuel pellets

    International Nuclear Information System (INIS)

    Novikov, V.; Medvedev, A.; Khvostov, G.; Bogatyr, S.; Kuzetsov, V.; Korystin, L.

    2005-01-01

    This paper is devoted to the modelling of unsteady state mechanical and thermo-physical behaviour of high burn-up VVER fuel at a power ramp. The contribution of the processes related to the kinetics of fission gas to the consequences of pellet-clad mechanical interaction is analysed by the example of integral VVER-440 rod 9 from the R7 experimental series, with a pellet burn-up in the active part at around 60 MWd/kgU. This fuel rod incurred ramp testing with a ramp value ΔW 1 ∼ 250 W/cm in the MIR research reactor. The experimentally revealed residual deformation of the clad by 30-40 microns in the 'hottest' portion of the rod, reaching a maximum linear power of up to 430 W/cm, is numerically justified on the basis of accounting for the unsteady state swelling and additional degradation of fuel thermal conductivity due to temperature-induced formation and development of gaseous porosity within the grains and on the grain boundaries. The good prediction capability of the START-3 code, coupled with the advanced model of fission gas related processes, with regard to the important mechanical (residual deformation of clad, pellet-clad gap size, central hole filling), thermal physical (fission gas release) and micro-structural (profiles of intra-granular concentration of the retained fission gas and fuel porosity across a pellet) consequences of the R7 test is shown. (authors)

  3. 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

  4. BNFL assessment of methods of attaining high burnup MOX fuel

    International Nuclear Information System (INIS)

    Brown, C.; Hesketh, K.W.; Palmer, I.D.

    1998-01-01

    It is clear that in order to maintain competitiveness with UO 2 fuel, the burnups achievable in MOX fuel must be enhanced beyond the levels attainable today. There are two aspects which require attention when studying methods of increased burnups - cladding integrity and fuel performance. Current irradiation experience indicates that one of the main performance issues for MOX fuel is fission gas retention. MOX, with its lower thermal conductivity, runs at higher temperatures than UO 2 fuel; this can result in enhanced fission gas release. This paper explores methods of effectively reducing gas release and thereby improving MOX burnup potential. (author)

  5. 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)

  6. 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)

  7. Data report on leach tests of Pu-doped UO2 in PBB1 brine: Salt Repository Project

    International Nuclear Information System (INIS)

    Gray, W.J.

    1987-10-01

    This report provides results from a series of leach tests conducted using nonirradiated uranium dioxide (UO 2 ) doped with plutonium (Pu) to simulate the alpha activity of spent fuel specimens used in recent spent fuel leach tests. The purpose was to determine whether alpha radiation from the spent fuel could be responsible for uranium release values in spent fuel leach tests in salt brine that were at least 100 times greater than from similar tests with nonirradiated UO 2 pellets. The data in this data report are preliminary; they have been neither analyzed nor evaluated. 2 refs., 2 figs., 8 tabs

  8. HELIOS calculations for UO2 lattice benchmarks

    International Nuclear Information System (INIS)

    Mosteller, R.D.

    1998-01-01

    Calculations for the ANS UO 2 lattice benchmark have been performed with the HELIOS lattice-physics code and six of its cross-section libraries. The results obtained from the different libraries permit conclusions to be drawn regarding the adequacy of the energy group structures and of the ENDF/B-VI evaluation for 238 U. Scandpower A/S, the developer of HELIOS, provided Los Alamos National Laboratory with six different cross section libraries. Three of the libraries were derived directly from Release 3 of ENDF/B-VI (ENDF/B-VI.3) and differ only in the number of groups (34, 89 or 190). The other three libraries are identical to the first three except for a modification to the cross sections for 238 U in the resonance range

  9. FBR pellet fabrication - density and dimensional control

    International Nuclear Information System (INIS)

    Rasmussen, D.E.; Schaus, P.S.

    1982-01-01

    The fuel pellet fabricating experience described in this paper involved pellet processing tests using mixed oxide (PuO 2 -UO 2 ) powders to produce fast breeder reactor (FBR) fuel pellets. Objectives of the pellet processing tests were to establish processing parameters for sintered-to-size fuel pellets to be used in an irradiation test in the Fast Flux Test Facility and to establish baseline fabrication control information. 26 figures, 7 tables

  10. Micrography of UO2 power of PPNY and France using JSM T-20

    International Nuclear Information System (INIS)

    Kasilani, N.S.; Hidayati; Hartati, P.

    1996-01-01

    As a quality control of processing to produce a fuel element using UO 2 powder, its necessary to be known the physical characteristic of the shape, size and surface condition of particle. This physical character influence the flow ability of powder particles and density of pellet. To create photomicrograph used a electron microscope, influenced by the condition of tools, specimen, and the skilled of the processing of pictures. The current of JSM T-20 is about 5 until 20 Kv, used 11 and 8 camera diaphragm with black and white films, specimen must be dried with electrical conductor property. The processing resulted an optimal photomicrograph. Micrograph of UO 2 powder of PPNY and France was investigated, yield a same grain form, surface structure are different, and range of size particle is 0.5 - 1.0 um. (author)

  11. 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

  12. Dissolution mechanism of UO2 at various parametric conditions

    International Nuclear Information System (INIS)

    Ollila, K.

    1988-04-01

    The aim of this experimental study is to investigate the solubility and dissolution mechanism of uranium dioxide under simulated disposal conditions of spent fuel. Unirradiated UO 2 is used as a surrogate for spent fuel. Two types of synthetic groundwaters were used in these experiments, on simulating the natural conditions deep in granitic bedrock (synthetic groundwater I) and the other simulating the effects of bentonite on groundwater (synthetic groundwater II). The effect of carbonate concentration was investigated by following dissolution in sodium bicarbonate solution as a function of bicarbonate concentration. Deionized wate was used as a reference water. All the experiments were carried out under both air-saturated, oxidizing and anoxic, reducing conditions. A separate test series under anoxic conditions was initiated in order to study the oxidation state of uranium. The experimental uranium solubilities are compared with the solubilities obtained from theoetical calculations by applying the geochemical code PHREEQ. The theoretical solubility values of uranium under oxidizing conditions calculated by PHREEQE are higher when compared to the corresponding experimental solubility values. The reason for the lower solubility values may be the mechanism of dissolution leading for example either to a situation where low dissolution rate is a limiting factor or to formation of some solid phase of uranium with lower solubility. Formation of a surface layer was observed on the pellet after dissolution in synthetic groundwater II. The theoretical solubility values under educing conditions calculated for uranium by PHREEQE appear to be in good agreement with the experimental solubility values

  13. Nuclear fuel pellet loading machine

    International Nuclear Information System (INIS)

    Kee, R.W.; Denero, J.V.

    1975-01-01

    An apparatus for loading nuclear fuel pellets on trays for transfer in a system is described. A conveyor supplies pellets from a source to a loading station. When the pellets reach a predetermined position at the loading station, a manual or automatically operated arm pushes the pellets into slots on a tray and this process is repeated until pellet sensing switches detect that the tray is full. Thereupon, the tray is lowered onto a belt or other type conveyor and transferred to other apparatus in the system, such as a furnace for sintering, and in some cases, reduction of UO 2 . 2 to UO 2 . The pellets are retained on the tray and subsequently loaded directly into fuel rods to be used in the reactor core. (auth)

  14. Modelling intragranular fission gas release in irradiation of sintered LWR UO2 fuel

    International Nuclear Information System (INIS)

    Loesoenen, Pekka

    2002-01-01

    A model for the release of stable fission gases by diffuion from sintered LWR UO 2 fuel grains is presented. The model takes into account intragranular gas bubble behaviour as a function of grain radius. The bubbles are assumed to be immobile and the gas migrates to grain boundaries by diffusion of single gas atoms. The intragranular bubble population in the model at low burn-ups or temperatures consists of numerous small bubbles. The presence of the bubbles attenuates the effective gas atom diffusion coefficient. Rapid coarsening of the bubble population in increased burn-up at elevated temperatures weakens significantly the attenuation of the effective diffusion coefficient. The solution method introduced in earlier papers, locally accurate method, is enhanced to allow accurate calculation of the intragranular gas behaviour in time varying conditions without excessive computing time. Qualitatively the detailed model can predict the gas retention in the grain better than a more simple model

  15. 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)

  16. Effects of hyperstoichiometry and fission products on the electrochemical reactivity of UO2 nuclear fuel

    International Nuclear Information System (INIS)

    Betteridge, J.S.; Scott, N.A.M.; Shoesmith, D.W.; Bahen, L.E.; Hocking, W.H.; Lucuta, P.G.

    1997-03-01

    The effects of hyperstoichiometry and fission products on the electrochemical reactivity Of UO 2 nuclear fuel have been systematically investigated using cyclic voltammetry and the O 2 reduction reaction. Significant constraints are placed on the active-site model for O 2 reduction by the modest impact of bulk hyperstoichiometry. Formation of the U 4 O 9 derivative phase was associated with a marked increase in transient surface oxidation/reduction processes, which probably involve localized attack and might be fostered by tensile stresses induced during oxidation. Electrocatalytic reduction Of O 2 on simulated nuclear fuel (SIMFUEL) has been determined to increase progressively with nominal burnup and pronounced enhancement of H 2 O reduction has been observed as well. Substitution of uranium by lower-valence (simulated) fission products, which was formerly considered the probable cause for this behaviour, has now been shown to merely provide good electrical conductivity. Instead, the enhanced reduction kinetics for O 2 and H 2 O on SIMFUEL can be fully accounted for by noble metals, which segregate to the UO 2 grain boundaries as micron-sized particles, despite their low effective surface area. Apparent convergence of the electrochemical properties Of UO 2 and SIMFUEL through natural corrosion likely reflects evolution toward a common active surface. (author)

  17. 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

  18. 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)

  19. Fission Gas Release in LWR Fuel Rods Exhibiting Very High Burn-Up

    DEFF Research Database (Denmark)

    Carlsen, H.

    1980-01-01

    Two UO2Zr BWR type test fuel rods were irradiated to a burn-up of about 38000 MWd/tUO2. After non-destructive characterization, the fission gas released to the internal free volume was extracted and analysed. The irradiation was simulated by means of the Danish fuel performance code WAFER-2, which...

  20. Structure change of fuel pellets

    International Nuclear Information System (INIS)

    Imanaka, Tetsuji

    1980-01-01

    The investigation of the broken pieces of fuel rods in Mihama No. 1 reactor was carried out in the Japan Atomic Energy Research Institute, and unexpectedly led to the post-irradiation tests. The investigation group of the Kyoto University Research Institute considers that the pursuit of the causes of accident by the government was insufficient, and the countermeasures are problematical, as the result of having examined various reports. In this study, the white foreign phase and swelling of cladding tubes were investigated, because these are especially important in view of the soundness of the fuel. Besides, the possibility of the oxidation of UO 2 pellets by cooling water was examined. It was found by metallographic test that the featuring phase different from UO 2 structure existed in the central part of pellets remaining in two broken fuel rod pieces. The report of JAERI judged that it is the product of solid phase reaction above a certain threshold temperature. The change of pellet structure observed in the white foreign phase and the swell of a cladding tube was caused by the oxidation of UO 2 pellets by primary coolant. The result of observation of the white foreign phase showed that it had been liquid phase at the time of the formation. From the thermodynamic examination based on oxygen potential, UO 2 is oxidized above 1100 deg C in the atmosphere of primary coolant. The liquid phase of the oxidized phase of UO 2 is formed above 1600 deg C. (Kako, I.)

  1. The MOX fuel behaviour test IFA-597.4/.5. Temperature and pressure data to a burn-up of 15 MWd/kg MOX

    International Nuclear Information System (INIS)

    Takano, K.

    1999-04-01

    The behaviour of MOX fuel should be investigated in detail for more effective use in the future, especially concerning its thermal performance and fission gas release. IFA-597.4 and IFA-597.5, containing two MOX fuel rods each with a fuel centre thermocouple and a pressure transducer, have been irradiated in the Halden Reactor to study the temperature threshold of fission gas release for MOX fuel and to explore potential differences in the thermal and fission gas release behaviour between solid and hollow pellets. The two rods of MOX fuel with an initial Pu-fissile content of 6.07 percent have solid pellets and hollow pellets respectively, and with an active length of about 220 mm. The diameter of the pellets is 8.05 mm with 180μm of diametral gap to the cladding. For the purpose of the test, power ramp operation, in which estimated peak temperature of the MOX pellets increases and decreases above and below the threshold for fission gas release in UO 2 fuel, is planned every 10 MWd/kgMOX of burn-up. The first ramp operation has been successfully performed at 10 MWd/kgMOX. When the estimated peak temperature of the fuel gets close to but below the threshold of UO 2 , fission gas release was observed at around 28 kW/m of power. Densification of the MOX pellets could be estimated to about 1.2 percent for the solid pellets and about 2,3 percent for the hollow pellets from normalised internal rod pressure. After 13.5 MWd/kgMOX the average assembly power has been operated low enough to observe swelling rate of MOX fuel pellets and behaviour after significant fission gas release. The burn-up had reached 15.5 MWd/kgMOX as of the end of 1998. The target burn-up of this MOX test is 60 MWd/kgMOX (author) (ml)

  2. Uranium dioxide pellets

    International Nuclear Information System (INIS)

    Zawidzki, T.W.

    1982-01-01

    A process for the preparation of a sintered, high density, large crystal grain size uranium dioxide pellet is described which involves: (i) reacting a uranyl nitrate of formula UO 2 (NO 3 ) 2 .6H 2 O with a sulphur source, at a temperature of from about 300 deg. C to provide a sulphur-containing uranium trioxide; (ii) reacting the thus-obtained modified uranium trioxide with ammonium nitrate to form an insoluble sulphur-containing ammonium uranate; (iii) neutralizing the thus-formed slurry with ammonium hydroxide to precipitate out as an insoluble ammonium uranate the remaining dissolved uranium; (iv) recovering the thus-formed precipitates in a dry state; (v) reducing the dry precipitate to UO 2 , and forming it into 'green' pellets; and (vi) sintering the pellets in a hydrogen atmosphere at an elevated temperature

  3. Simulations and Experimental Measurements of UO2 Thermal Conductivity

    International Nuclear Information System (INIS)

    Stanek, Christopher Richard; Gofryk, Krzysztof; Tonks, Michael; Andersson, Anders David Ragnar; Liu, Xiang-Yang; Lashley, Jason Charles; Uberuaga, Blas P.; Mcclellan, Kenneth James

    2015-01-01

    Spin-phonon interactions lead to low @@ of UO 2 (and behave like a defect), and this has implications for nuclear fuel performace. The inability to capture spin-phonon scattering leads to inherent errors. The interplay between magnetism and structural asymmetry in UO 2 displays rich physics. Grain boundary structure plays a role which must be taken into account.

  4. Influence of radiolysis on UO2 fuel matrix dissolution under disposal conditions. Literature Study

    International Nuclear Information System (INIS)

    Ollila, K.

    2011-05-01

    . Interaction tests between U(VI) and Fe 2+ have shown the reduction of U(VI) hydroxide complexes to take place by Fe 2+ in aqueous solution. A decrease in corrosion and release rates has been observed in the experiments with spent fuel and alpha-doped UO2 in the presence of hydrogen. Hydrogen seems to suppress the oxidation and dissolution of UO 2 . The H 2 concentration of around 10 -5 M seemed to be a threshold for oxidation of UO 2 to occur in the tests with 10% 233 U-doped UO 2 in 10 -2 M NaCl (+2 x 10 -3 M NaHCO 3 ) solution. Dissolved hydrogen concentrations from the anaerobic corrosion of iron are estimated to be in the 10 -2 to 10 -1 M range. High burnup fuel did not seem to display any enhanced dissolution in the presence of 3.3 x 10 -2 M H 2 . The level of U concentrations in solution has been very low in the presence of H 2 , ∼ 10 -11 M, lower than the solubility of amorphous UO 2 . The dissolved U concentrations at the same level have been measured in the presence of metallic Fe. A hypothesis of α radiation induced hydrogen annealing, or a crystallization process in the amorphous phase has been presented in literature to explain low U concentrations. A hypothetical mechanism to explain the effects of hydrogen under α radiolysis conditions has been presented. The observations suggest that the consumption of oxidants is a surface-catalyzed process. (orig.)

  5. 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.)

  6. Experimental simulation of irradiation effects on thermomechanical behaviour of UO2 fuel: Impact of solid and gaseous fission products

    International Nuclear Information System (INIS)

    Balland, J.

    2007-12-01

    Predictive simulation of thermomechanical behaviour of nuclear fuel has to take into account irradiation effects. Fission Products (FP) can modify the thermomechanical behaviour of UO 2 . During this thesis, differentiation was made between fission products which create a solid solution with UO 2 and gaseous products, generating pressurized bubbles. SIMFUELS containing gadolinium oxide and pressurized argon bubbles were manufactured, respectively by conventional process and by Gas Pressure Sintering. Brittle and ductile behaviour of UO 2 was investigated, under experimental conditions representative of Pellet-Cladding Interaction (PCI), respectively with 3 points bending tests and compressive creep tests. Investigation of brittle behaviour of UO 2 showed that fracture is mainly controlled by natural defects, like porosities, acting like starting points for cracks propagation. Addition of simulates fission products increase the brittle-to-ductile transition temperature of UO 2 , up to 400-500 C regarding FP in solid solution, and up to 200 C for gaseous products. Fission products although reduce fracture stresses, by a factor between 1.5 and 4, respectively for gas bubbles and solid solutions. Decrease of fracture stress is linked to an increase of microstructural defects due the solid solution and to pressurized bubbles located at grain boundaries. Pellets were tested under compressive solicitation at high temperatures. Experimental results of creep tests are well represented by Norton laws. Creep controlling mechanisms are evidenced by microstructural analysis performed on pellets at different strains. On the basis of calculations made for fuels having the same microstructures than the SIMFUELs, a creep factor is determined. It revealed a strong hardening effect of the solid solution, due to the fact that the added elements anchor the dislocations, whereas pressurized bubbles showed a coupling between hardening and softening effects. (author)

  7. 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.)

  8. 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

  9. 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)

  10. 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)

  11. 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

  12. 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

  13. Dissolution of unirradiated UO2 fuel in synthetic groundwater. Final report (1996-1998)

    International Nuclear Information System (INIS)

    Ollila, K.

    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 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 2 solubility under the disposal conditions of spent fuel, varied from strongly oxidising (air-saturated), anaerobic (N 2 , O 2 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 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 2 powder showed that the increase in the salinity ( -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 3 O 8 -UO 3 ) was identified with XRD when studying possible secondary phases after the contact time of one year with all groundwater compositions. Longer contact times are needed to identify secondary phases predicted by modelling (EQ3/6). In the anoxic dissolution experiments with UO 2 pellets, the

  14. Correlation between fuel structure and mechanical properties of UO2

    International Nuclear Information System (INIS)

    Blank, H.; Mandler, R.; Matzke, H.; Routbort, J.; Werner, P.

    1982-10-01

    The relation between the structure of a UO 2 fuel and its mechanical properties are discussed and illustrated for particular types of UO 2 by measurements of fracture surface energy, hardness, fracture stress and of compressive deformation at 1870 and 1970 0 K. This gives the background for treating the question whether it is possible to find a simple experimental method for correlating the mechanical properties of UO 2 before irradiation with those after various irradiation histories. Hardness measurements might be such a method if combined with a detailed structural analysis and sufficient knowledge about the irradiation history

  15. Ceramic UO2 powder production at Cameco Corporation

    International Nuclear Information System (INIS)

    Mulligan, J.J.

    2005-01-01

    This paper describes the various aspects of ceramic grade UO 2 powder production at Cameco Corporation's Port Hope conversion facility. It discusses the significant safety systems, production processes and plant monitoring and control systems. It also provides an insight into how various support groups such as Quality Assurance, Analytical Services, and Technology Development contribute to the consistent production of high quality UO 2 powder. 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 that has consistently met customer requirements. (author)

  16. 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)

  17. Experimental investigations of the meltdown phase of UO2-Zircaloy fuel rods under conditions of failure of emergency cooling

    International Nuclear Information System (INIS)

    Hagen, S.; Mack, A.; Malauschek, H.; Wallenfels, K.

    1975-01-01

    In the monoxidizing helium atmosphere at 1,850 0 C Zircaloy and UO 2 interact violently. The result is a combined meltdown of pellets and can. This phenomenon appears independent of the velocity of temperature rise. In air the oxid skin splits open at 1,890 0 C and the earlier molten material of the interior begins to flow out. When heating up to more than 2,200 0 C the oxid skin remains solid nevertheless. (orig.) [de

  18. 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

  19. TEM characterization of UO2-Gd2O3 nuclear fuels synthesized by coprecipitation method

    International Nuclear Information System (INIS)

    Soldati, A.; Gana Watkins, I.; Menghini, J.; Prado, M.

    2013-01-01

    We present a micro and nano structural characterization of 4% weight doped Gd 2 O 3 -UO 2 pellet using Transmission Electron Microscopy (TEM). Agglomerate morphology and crystallite sizes were determined using light/dark field and high resolution (HR-TEM) images. Convergent beam Energy Dispersive Spectroscopy (EDS) and Electron Diffraction (ED) were used to evaluate sample composition and homogeneity, even at the nanometer scale. We obtained an average crystallite size of 90±20 nm. Moreover, from TEM-EDS analyses we determined the presence of Gadolinium in all the analyzed crystallites but with 25% variation among their concentrations. These results show the capability of TEM analysis to characterize a nuclear fuel pellet with burnable poisons nano structure and homogeneity.(author)

  20. Overall models and experimental database for UO2 and MOX fuel increasing performance

    International Nuclear Information System (INIS)

    Bernard, L.C.; Blanpain, P.

    2001-01-01

    COPERNIC is an advanced fuel rod performance code developed by Framatome. It is based on the TRANSURANUS code that contains a clear and flexible architecture, and offers many modeling possibilities. The main objectives of COPERNIC are to accurately predict steady-state and transient fuel operations at high burnups and to incorporate advanced materials such as the Framatome M5-alloy cladding. An extensive development program was undertaken to benchmark the code to very high burnups and to new M5-alloy cladding data. New models were developed for the M5-alloy cladding and the COPERNIC thermal models were upgraded and improved to extend the predictions to burnups over 100 GWd/tM. Since key phenomena, like fission gas release, are strongly temperature dependent, many other models were upgraded also. The COPERNIC qualification range extends to 67, 55, 53 GWd/tM respectively for UO 2 , UO 2 -Gd 2 O 3 , and MOX fuels with Zircaloy-4 claddings. The range extends to 63 GWd/tM with UO 2 fuel and the advanced M5-alloy cladding. The paper focuses on thermal and fission gas release models, and on MOX fuel modeling. The COPERNIC thermal model consists of several submodels: gap conductance, gap closure, fuel thermal conductivity, radial power profile, and fuel rim. The fuel thermal conductivity and the gap closure models, in particular, have been significantly improved. The model was benchmarked with 3400 fuel centerline temperature data from many French and international programs. There are no measured to predicted statistical biases with respect to linear heat generation rate or burnup. The overall quality of the model is state-of-the-art as the model uncertainty is below 10 %. The fission gas release takes into account athermal and thermally activated mechanisms. The model was adapted to MOX and Gadolinia fuels. For the heterogeneous MOX MIMAS fuels, an effective burnup is used for the incubation threshold. For gadolinia fuels, a scaled temperature effect is used. The

  1. 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

  2. PECITIS-II, a computer program to predict the performance of collapsible clad UO2 fuel elements

    International Nuclear Information System (INIS)

    Anand, A.K.; Anantharaman, K.; Sarda, V.

    1978-01-01

    The Indian power programme envisages the use of PHWRs, which use collapsible clad UO 2 fuel elements. A computer code, PECITIS-II, developed for the analysis of this type of fuel is described in detail. The sheath strain and fission gas pressure are evaluated by this method. The pellet clad gap conductance is calculated by Ross and Solute model. The pellet thermal expansion is calculated by assuming a two zone model, i.e. a plastic core surrounded by an elastic cracked annulus. (author)

  3. 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)

  4. 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

  5. UO2-PuO2 fuel pin capsule-irradiations of the test series FR 2-5a

    International Nuclear Information System (INIS)

    Dienst, W.; Goetzmann, O.; Schulz, B.

    1975-06-01

    In the capsule-irradiation test series FR 2-5a, short UO 2 -PuO 2 fuel pins (80 mm fuel length) of 7 mm diameter were irradiated in a thermal neutron flux at mean rod powers of 400 - 450 W/cm and mean cladding surface temperatures of 500 - 550 0 C to burnups of 0.6, 1.8 and 5.0 at% (U + Pu). Void volume redistribution in the fuel pins was examined in micrographs of cross-sections by measuring crack widths, central void diameters, and fuel porosity. The width of the radial cracks at the outer fuel rim was taken as a basis for measuring the irradiation-induced densification of the UO 2 -PuO 2 fuel. The result was that the final fuel density after irradiation-induced densification amounted to 92 - 94% TD and had already been reached after 0.6 at% burnup. The porosity measurement on fuel cross-sections was to show a possible dependence of the radial porosity redistribution on the initial sintered density. Examining the fuel pin diameters after irradiation showed permanent cladding strains after 5 at% burnup, which must be due to mechanical interaction with the fuel. To judge if the chemical compatibility between the fuel and the cladding of Cr-Ni-stainless steel 1.4988, the depths of chemical attack on the cladding inside was measured by micrographs of fuel pin cross-sections. (orig./GSC) [de

  6. Automated generation of burnup chain for reactor analysis applications

    International Nuclear Information System (INIS)

    Tran, Viet-Phu; Tran, Hoai-Nam; Yamamoto, Akio; Endo, Tomohiro

    2017-01-01

    This paper presents the development of an automated generation of burnup chain for reactor analysis applications. Algorithms are proposed to reevaluate decay modes, branching ratios and effective fission product (FP) cumulative yields of a given list of important FPs taking into account intermediate reactions. A new burnup chain is generated using the updated data sources taken from the JENDL FP decay data file 2011 and Fission yields data file 2011. The new burnup chain is output according to the format for the SRAC code system. Verification has been performed to evaluate the accuracy of the new burnup chain. The results show that the new burnup chain reproduces well the results of a reference one with 193 fission products used in SRAC. Burnup calculations using the new burnup chain have also been performed based on UO_2 and MOX fuel pin cells and compared with a reference chain th2cm6fp193bp6T.

  7. Manufacture of a UO2-Based Nuclear Fuel with Improved Thermal Conductivity with the Addition of BeO

    Science.gov (United States)

    Garcia, Chad B.; Brito, Ryan A.; Ortega, Luis H.; Malone, James P.; McDeavitt, Sean M.

    2017-12-01

    The low thermal conductivity of oxide nuclear fuels is a performance-limiting parameter. Enhancing this property may provide a contribution toward establishing accident-tolerant fuel forms. In this study, the thermal conductivity of UO2 was increased through the fabrication of ceramic-ceramic composite forms with UO2 containing a continuous BeO matrix. Fuel with a higher thermal conductivity will have reduced thermal gradients and lower centerline temperatures in the fuel pin. Lower operational temperatures will reduce fission gas release and reduce fuel restructuring. Additions of BeO were made to UO2 fuel pellets in 2.5, 5, 7.5, and 10 vol pct concentrations with the goals of establishing reliable lab-scale processing procedures, minimizing porosity, and maximizing thermal conductivity. The microstructure was characterized with electron probe microanalysis, and the thermal properties were assessed by light flash analysis and differential scanning calorimetry. Reliable, high-density samples were prepared using compaction pressure between 200 and 225 MPa and sintering times between 4 and 6 hours. It was found that the thermal conductivity of UO2 improved approximately 10 pct for each 1 vol pct BeO added over the measured temperature range 298.15 K to 523.15 K (25 °C to 250 °C) with the maximum observed improvement being ˜ 100 pct, or doubled, at 10 vol pct BeO.

  8. Finite element analysis of local overheating within plutonium enriched UO2 fuel rods caused by PuO2 islands

    International Nuclear Information System (INIS)

    Sarmiento, G.S.

    1980-01-01

    Within natural UO 2 fuel elements enriched with plutonium, this last material should form PuO 2 solid solutions inside the UO 2 pellets, in a wide range of concentrations. If the solutions are obtained by mechanical mixing of the oxides, PuO 2 islands are formed in the UO 2 matrix. These islands may be the source of several problems in the fuel behaviour, the most important being the overheating of the matrix in the neighbourhood of the particles. It is caused by the large fission cross section of plutonium compared with that of uranium. A detailed study of the thermal effects produced by PuO 2 particles in the UO 2 matrix and the cladding is then important for the specification of their permissible size. A portion of the fuel rods with spherical particles in the most significant places was studied. In order to obtain the dimensionless overheating of the fuel and cladding produced by the presence of those particles, the spatial distribution of temperature was calculated, solving the stationary and linear bidimensional equation of heat conducting using a finite element code. Several geometrical variables and material properties have been taken as dimensionless parameters. A satisfactory convergence of the numerical results to an asymptotic limit with a well-known exact solution, has been obtained. (orig.)

  9. Fission gas behaviour in UO2 under steady state and transient conditions

    International Nuclear Information System (INIS)

    Zimmermann, H.

    1980-01-01

    Fission gas behaviour in UO 2 is determined by the limited capacity of the fuel to retain fission gas. This capacity depends primarily on temperature, but also on fission rate, pressure loading, and fuel microstructure. Under steady state irradiation conditions fission gas behaviour can be described qualitatively as follows: At the beginning of the irradiation most of the fission gas remains in the grains in irradiation-induced solution. With increasing gas content in the grains the gas transport to the grain boundaries increases, too. The fission gas release from the grain boundaries occurs primarily by interlinkage of inter-granular bubbles. The fission gas release without noticeable fuel swelling during the short-term heating in the LOCA tests and the powdering of the high burnup UO 2 in the annealing tests can only be accounted for by formation of inter-granular separations, which are caused by the fission gas accumulated in the grain boundaries. Besides this short-term effect there are diffusion-controlled long-term effects, such as growth and coalescence of bubbles and formation of inter-connected porosity, which result in time-dependent fission gas release and fuel swelling

  10. Gaseous swelling of B4C and UO2 fuel: similarities and differences

    International Nuclear Information System (INIS)

    Evdokimov, I.; Khoruzhii, O.; Kourtchatov, S.; Likhanskii, V.; Matweev, L.

    2001-01-01

    A major factor limiting the resource of control rods (CRs) for WWER-1000 reactors is their radiation damage. Radiation induced embrittlement of the CRs cladding, core swelling and gaseous internal pressure in CRs result in mechanical core-cladding interaction. This work is devoted to the physical analysis of processes that control the structural changes in neutron absorber elements with B 4 C under irradiation in water reactors. Particularly, the analysis of mechanisms of the helium porosity formation in B 4 C is undertaken. In view of the deficiency of experimental data on the subject, a fruitful approach to the problem is a comparative analysis of the swelling mechanisms in B 4 C absorber and UO 2 fuel. Using this similarity a phenomenological model of fission gas behavior in boron carbide is proposed. The model predictions for radial profile of 10 B burnup under influence of thermal and epithermal neutrons are compared with experimental results. The main results show that despite the external similarity of the process of fission gas accumulation in UO 2 and in B 4 C, phenomenology of gaseous swelling is much different for the fuel and the CR core. The reason for that difference is the distinction of physical conditions in irradiated fuel and CR core

  11. Study on the effect of UO2 composition on dissolution of sintered (Th-U)O2 MOX by microwave heating

    International Nuclear Information System (INIS)

    Singh, G.; Malav, R.K.; Fulzele, A.K.; Prakash, A.; Afzal, Md.; Panakkal, J.P.

    2010-01-01

    Full text: Complete dissolution of sample is a prerequisite for any chemical analysis in liquid form. Dissolution of ThO 2 based mixed oxide sample like (Th-U)O 2 , (Th-Pu)O 2 is a challenging job due to single oxidation state of thorium (IV). The present paper describes a study carried out on effect of UO 2 composition on dissolution of sintered (Th-U)O 2 mixed oxide pellets, in 0.05M HF prepared in 16 M HNO 3 . The experiments were performed in PTFE pressure vessels which could stand up to ∼ 250 deg C and safely operated up to 120 psi in an indigenous 700 watts microwave digestion system. ThO 2 , ThO 2 -3.75%UO 2 and ThO 2 -5%UO 2 pellets (∼ 6 g each) were dissolved in 60 mL of 16M HNO 3 /HF mixtures (0.05M HF in 16 M HNO 3 ) in PTFE (teflon) made pressure vessels (each experiment triplicate) at a pressure of ∼ 120 psi. Samples (two at an instant) were withdrawn after each hour and Th in the solution was determined by EDTA complexometric titration where end point was detected visually. Table 1 shows the results of percent dissolution of Th (mean of three experiments) for the sintered pellet after each interval of time until 100% dissolution. The plot for percent dissolution of Th (mean Th %) against time taken for sintered pellets is shown. Application of microwave heating has been applied for the dissolution of uncrushed sintered ThO 2 and (Th-U)O 2 pellets. It is quite evident from Th% dissolved versus time curves that the dissolution is faster as percentage of UO 2 in (Th-U)O 2 MOX solid solution increases. This is attributed to UO 2 as it can easily absorb microwave energy, leading to high temperature

  12. Behaviour of short-lived iodines in operating UO2 fuel elements

    International Nuclear Information System (INIS)

    Lipsett, J.J.; Hastings, I.J.; Hunt, C.E.L.

    1984-11-01

    Sweep gas experiments have been done to determine the behaviour of short-lived fission products within operating UO 2 fuel elements at linear powers of 45, 54, and 60 KW/m, and to burnups of 70, 80, and 50 MWh/kgU respectively. Although radioiodine transport was not observed directly during normal operation, equilibrium gap inventories for I-131 were deduced from the shutdown decay behaviour of the fission gases. These inventories were a strong function of fuel power and ranged from 10 GBq (0.27 Ci) to 100 GBq (2.7 Ci) over the range tested. We conclude that the iodine inventory was adsorbed onto the fuel and/or sheath surfaces with a volatile fraction of less than 10 -2 and a charcoal-filter-penetrating fraction of less than 2x10 -4

  13. Dissolution rates of unirradiated UO2, UO2 doped with 233U, and spent fuel under normal atmospheric conditions and under reducing conditions using an isotope dilution method

    International Nuclear Information System (INIS)

    Ollila, Kaija; Albinsson, Yngve; Oversby, Virginia; Cowper, Mark

    2003-10-01

    The experimental results given in this report allow us to draw the following conclusions. 1) Tests using unirradiated fuel pellet materials from two different manufacturers gave very different dissolution rates under air atmosphere testing. Tests for fragments of pellets from different pellets made by the same manufacturer gave good agreement. This indicates that details of the manufacturing process have a large effect on the behavior of unirradiated UO 2 in dissolution experiments. Care must be taken in interpreting differences in results obtained in different laboratories because the results may be affected by manufacturing effects. 2) Long-term tests under air atmosphere have begun to show the effects of precipitation. Further testing will be needed before the samples reach steady state. 3) Testing of unirradiated UO 2 in systems containing an iron strip to produce reducing conditions gave [U] less than detection limits ( 235 U added as spike was recovered, indicating that 90% of the spike had precipitated onto the solid sample or the iron strip. 9) Tests of UO 2 pellet materials containing 233 U to provide an alpha decay activity similar to that expected for spent fuel 3000 and 10,000 years after disposal showed that the pellet materials behaved as expected under air atmosphere conditions, showing that the manufacturing method was successful. 10) Early testing of the 233 U-doped materials under reducing conditions showed relatively rapid (30 minute) dissolution of small amounts of U at the start of the puff test procedure. Results of analyses of an acidified fraction of the same solutions after 1 or 2 weeks holding indicate that the solutions were inhomogeneous, indicating the presence of colloidal material or small grains of solid. 11) Samples from the 233 U-doped tests initially indicated dissolution of solid during the first week of testing, with some indication of more rapid dissolution of the material with the higher doping. 12) The second cycle of testing

  14. Fabrication of 0.5-inch diameter FBR mixed oxide fuel pellets

    International Nuclear Information System (INIS)

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

    1979-01-01

    Large diameter (0.535 inch) mixed oxide fuel pellets for Fast Breeder Reactor application were successfully fabricated by the cold-press-and-sinter technique. Enriched UO 2 , PuO 2 -UO 2 , and PuO 2 -ThO 2 compositions were fabricated into nominally 90% theoretical density pellets for the UO 2 and PuO 2 -UO 2 compositions, and 88% and 93% T.D. for the PuO 2 -ThO 2 compositions. Some processing adjustments were required to achieve satisfactory pellet quality and density. Furnace heating rate was reduced from 200 to 50 0 C/h for the organic binder burnout cycle for the large, 0.535-inch diameter pellets to eliminate pellet cracking during sintering. Additional preslugging steps and die wall lubrication during pressing were used to eliminate pressing cracks in the PuO 2 -ThO 2 pellets

  15. An Overview of Current and Past W-UO[2] CERMET Fuel Fabrication Technology

    International Nuclear Information System (INIS)

    Douglas E. Burkes; Daniel M. Wachs; James E. Werner; Steven D. Howe

    2007-01-01

    Studies dating back to the late 1940s performed by a number of different organizations and laboratories have established the major advantages of Nuclear Thermal Propulsion (NTP) systems, particularly for manned missions. A number of NTP projects have been initiated since this time; none have had any sustained fuel development work that appreciably contributed to fuel fabrication or performance data from this era. As interest in these missions returns and previous space nuclear power researchers begin to retire, fuel fabrication technologies must be revisited, so that established technologies can be transferred to young researchers seamlessly and updated, more advanced processes can be employed to develop successful NTP fuels. CERMET fuels, specifically W-UO2, are of particular interest to the next generation NTP plans since these fuels have shown significant advantages over other fuel types, such as relatively high burnup, no significant failures under severe transient conditions, capability of accommodating a large fission product inventory during irradiation and compatibility with flowing hot hydrogen. Examples of previous fabrication routes involved with CERMET fuels include hot isostatic pressing (HIPing) and press and sinter, whereas newer technologies, such as spark plasma sintering, combustion synthesis and microsphere fabrication might be well suited to produce high quality, effective fuel elements. These advanced technologies may address common issues with CERMET fuels, such as grain growth, ductile to brittle transition temperature and UO2 stoichiometry, more effectively than the commonly accepted 'traditional' fabrication routes. Bonding of fuel elements, especially if the fabrication process demands production of smaller element segments, must be investigated. Advanced brazing techniques and compounds are now available that could produce a higher quality bond segment with increased ease in joining. This paper will briefly address the history of CERMET

  16. 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)

  17. Nuclear fuel pellet production method and nuclear fuel pellet

    International Nuclear Information System (INIS)

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

    1993-01-01

    In a method of manufacturing nuclear fuel pellets by compression-molding UO 2 powders followed by sintering, a sintering agent having a composition of about 40 to 80 wt% of SiO 2 and the balance of Al 2 O 3 , a sintering agent at a ratio of 10 to 500 ppm based on the total amount of UO 2 and UO 2 powders are mixed, compression molded and then sintered at a sintering temperature of about 1500 of 1800degC. The UO 2 particles have an average grain size of about 20 to 60μm, most of the crystal grain boundary thereof is coated with a glassy or crystalline alumina silicate phase, and the porosity is about 1 to 4 vol%. With such a constitution, the sintering agent forms a single liquid phase eutectic mixture during sintering, to promote a surface reaction between nuclear fuel powders by a liquid phase sintering mechanism, increase their density and promote the crystal growth. Accordingly, it is possible to lower the softening temperature, improve the creep velocity of the pellets and improve the resistance against pellet-clad interaction. (T.M.)

  18. Application of boron and gadolinium burnable poison particles in UO2 and PUO2 fuels in HTRs

    International Nuclear Information System (INIS)

    Kloosterman, J.L.

    2003-01-01

    Burnup calculations have been performed on a standard HTR fuel pebble (fuel zone with radius of 2.5 cm surrounded with a 0.5 cm thick graphite layer) and burnable poison particles (BPPs) containing B 4 C made of pure 10 B or containing Gd 2 O 3 made of natural Gd. Two types of fuel were considered: UO 2 fuel made of 8% enriched uranium and PuO 2 fuel made of plutonium from LWR spent fuel. The radius of the BPP and the number of particles per fuel pebble were varied to find the flattest reactivity-to-time curve. For the UO 2 fuel, the reactivity swing is lowest (around 2%) for BPPs made of B 4 C with radius of 75 μm. In this case around 1070 BPPs per fuel pebble are needed. For the PuO 2 fuel to get a reactivity swing below 4%, the optimal radius of the BPP is the same, but the number of particles per fuel pebble should be around 1600. The optimal radius of the Gd 2 O 3 particles in the UO 2 fuel is about 10 times that of the B 4 C particles. The reactivity swing is around 3% when each fuel pebble contains only 9 BPPs with radius of 840 μm. The results of the Gd particles illustrate nicely the usage of black burnable poison particles introduced by Van Dam [Ann. Nuclear Energy 27 (2000) 733

  19. Burnup performance of rock-like oxide (ROX) fuel in small pebble bed reactor with accumulative fuel loading scheme

    International Nuclear Information System (INIS)

    Simanullang, Irwan Liapto; Obara, Toru

    2017-01-01

    Highlights: • Burnup performance using ROX fuel in PBR with accumulative fuel loading scheme was analyzed. • Initial excess reactivity was suppressed by reducing 235 U enrichment in the startup condition. • Negative temperature coefficient was achieved in all condition of PBR with accumulative fuel loading scheme using ROX fuel. • Core lifetime of PBR with accumulative fuel loading scheme using ROX fuel was shorter than with UO 2 fuel. • In PBR with accumulative fuel loading scheme using ROX fuel, achieved discharged burnup can be as high as that for UO 2 fuel. - Abstract: The Japan Atomic Energy Agency (JAEA) has proposed rock-like oxide (ROX) fuel as a new, once-through type fuel concept. Here, burnup performance using ROX fuel was simulated in a pebble bed reactor with an accumulative fuel loading scheme. The MVP-BURN code was used to simulate the burnup calculation. Fuel of 5 g-HM/pebble with 20% 235 U enrichment was selected as the optimum composition. Discharged burnup could reach up to 218 GWd/t, with a core lifetime of about 8.4 years. However, high excess reactivity occurred in the initial condition. Initial fuel enrichment was therefore reduced from 20% to 4.65% to counter the initial excess reactivity. The operation period was reduced by the decrease of initial fuel enrichment, but the maximum discharged burnup was 198 GWd/t. Burnup performance of ROX fuel in this reactor concept was compared with that of UO 2 fuel obtained previously. Discharged burnup for ROX fuel in the PBR with an accumulative fuel loading scheme was as high as UO 2 fuel. Maximum power density could be lowered by introducing ROX fuel compared to UO 2 fuel. However, PBR core lifetime was shorter with ROX fuel than with UO 2 fuel. A negative temperature coefficient was achieved for both UO 2 and ROX fuels throughout the operation period.

  20. Studies on the Sintering Behaviour of UO2-Gd2O3 Nuclear Fuel

    International Nuclear Information System (INIS)

    Durazzo, Michelangelo; Gracher Riella, Humberto

    2008-01-01

    The incorporation of gadolinium directly into nuclear power reactor fuel is important from the point of reactivity compensation and adjustment of power distribution enabling thus longer fuel cycles and optimized fuel utilization. The incorporation of Gd 2 O 3 powder directly into the UO 2 powder by dry mechanical blending is the most attractive process because of its simplicity. Nevertheless, processing by this method leads to difficulties while obtaining sintered pellets with the minimum required density. This is due to blockages during the sintering process. There is little information in published literature about the possible mechanism for this blockage and this is restricted to the hypothesis based on formation of a low diffusivity Gd rich (U,Gd)O 2 phase. Experimental evidences indicated the existence of phases in the (U,Gd)O 2 system with structure different from the fluorite type structure of UO 2 . The apparition of these new phases coincides with the lowering of the density after sintering and with the lowering of the interdiffusion coefficient. However, it has been shown experimentally that the sintering blockage phenomena cannot be explained on the basis of the formation of low diffusivity Gd rich (U,Gd)O 2 phases. The work was continued to investigate other possible blocking mechanism. (authors)

  1. The heating of UO_2 kernels in argon gas medium on the physical properties of sintered UO_2 kernels

    International Nuclear Information System (INIS)

    Damunir; Sri Rinanti Susilowati; Ariyani Kusuma Dewi

    2015-01-01

    The heating of UO_2 kernels in argon gas medium on the physical properties of sinter UO_2 kernels was conducted. The heated of the UO_2 kernels was conducted in a sinter reactor of a bed type. The sample used was the UO_2 kernels resulted from the reduction results at 800 °C temperature for 3 hours that had the density of 8.13 g/cm"3; porosity of 0.26; O/U ratio of 2.05; diameter of 1146 μm and sphericity of 1.05. The sample was put into a sinter reactor, then it was vacuumed by flowing the argon gas at 180 mmHg pressure to drain the air from the reactor. After that, the cooling water and argon gas were continuously flowed with the pressure of 5 mPa with 1.5 liter/minutes velocity. The reactor temperature was increased and variated at 1200-1500 °C temperature and for 1-4 hours. The sinters UO_2 kernels resulted from the study were analyzed in term of their physical properties including the density, porosity, diameter, sphericity, and specific surface area. The density was analyzed using pycnometer with CCl_4 solution. The porosity was determined using Haynes equation. The diameters and sphericity were showed using the Dino-lite microscope. The specific surface area was determined using surface area meter Nova-1000. The obtained products showed the the heating of UO_2 kernel in argon gas medium were influenced on the physical properties of sinters UO_2 kernel. The condition of best relatively at 1400 °C temperature and 2 hours time. The product resulted from the study was relatively at its best when heating was conducted at 1400 °C temperature and 2 hours time, produced sinters UO_2 kernel with density of 10.14 gr/ml; porosity of 7 %; diameters of 893 μm; sphericity of 1.07 and specific surface area of 4.68 m"2/g with solidify shrinkage of 22 %. (author)

  2. 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.)

  3. Creep behavior of UO2 above 20000C

    International Nuclear Information System (INIS)

    Slagle, O.D.

    1978-01-01

    A series of high temperature creep measurements were made for UO 2 in the temperature range from 2000 0 C to the melting temperature. The effects of temperature, stress and accrued strain on the creep rate have been measured. The results indicate that additional creep mechanisms are being activated at the higher temperatures

  4. 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

  5. Method for fluoride ion depletion of UO2 powders

    International Nuclear Information System (INIS)

    Beutner, R.; Ploeger, F.

    1978-01-01

    The method described consists in removing the hydrogen still present from the reduction during the preparation of UO 2 as completely as possible and in performing a pyrohydrolysis at temperatures above 650 0 C for at least 45 minutes. The removal of fluorine is necessary in order to avoid cladding tube damaging. (UA) [de

  6. 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)

  7. Effect of helium pressure on the response of unirradiated UO2 subjected to thermal transients

    International Nuclear Information System (INIS)

    Fenske, G.R.; Chapello, P.M.; Emerson, J.E.; Poeppel, R.B.

    1983-01-01

    The effect of helium pressure on the transient response of unirradiated depleted UO 2 subjected to simulated hypothetical loss-of-flow accidents in a gas-cooled fast reactor was examined by use of the direct electrical heating technique. Transient tests were performed at pressures ranging from 7 to 10 X 10 5 Pa(7 to 10 atm) to 7 to 8 MPa (70 to 80 atm) on radially restrained and unrestrained fuel segments. The average heating rates ranged from about17 to 240 J/g x s. The results indicate that while the mechanical integrity of the fuel segment was independent of the test pressure, the rapid ejection of molten fuel from pellet interfaces of unrestrained fuel, observed at the lower pressures, was delayed or suppressed at the higher pressures

  8. Phonon density of states and anharmonicity of UO2

    Science.gov (United States)

    Pang, Judy W. L.; Chernatynskiy, Aleksandr; Larson, Bennett C.; Buyers, William J. L.; Abernathy, Douglas L.; McClellan, Kenneth J.; Phillpot, Simon R.

    2014-03-01

    Phonon density of states (PDOS) measurements have been performed on polycrystalline UO2 at 295 and 1200 K using time-of-flight inelastic neutron scattering to investigate the impact of anharmonicity on the vibrational spectra and to benchmark ab initio PDOS simulations performed on this strongly correlated Mott insulator. Time-of-flight PDOS measurements include anharmonic linewidth broadening, inherently, and the factor of ˜7 enhancement of the oxygen spectrum relative to the uranium component by the increased neutron sensitivity to the oxygen-dominated optical phonon modes. The first-principles simulations of quasiharmonic PDOS spectra were neutron weighted and anharmonicity was introduced in an approximate way by convolution with wave-vector-weighted averages over our previously measured phonon linewidths for UO2, which are provided in numerical form. Comparisons between the PDOS measurements and the simulations show reasonable agreement overall, but they also reveal important areas of disagreement for both high and low temperatures. The discrepancies stem largely from a ˜10 meV compression in the overall bandwidth (energy range) of the oxygen-dominated optical phonons in the simulations. A similar linewidth-convoluted comparison performed with the PDOS spectrum of Dolling et al. obtained by shell-model fitting to their historical phonon dispersion measurements shows excellent agreement with the time-of-flight PDOS measurements reported here. In contrast, we show by comparisons of spectra in linewidth-convoluted form that recent first-principles simulations for UO2 fail to account for the PDOS spectrum determined from the measurements of Dolling et al. These results demonstrate PDOS measurements to be stringent tests for ab inito simulations of phonon physics in UO2 and they indicate further the need for advances in theory to address the lattice dynamics of UO2.

  9. IFPE/IFA-597.3, centre-line temperature, fission gas release and clad elongation at high burn-up (60-62 MWd/kg)

    International Nuclear Information System (INIS)

    Turnbull, J.A.

    2003-01-01

    Description: The fuel segments for the high burn-up integral rod behaviour test IFA-597 were taken from fuel rod 33-25065, which was irradiated in the Ringhals 1 BWR for approximately 12 years. The irradiation of this rod and its sibling rod 33-25046 was performed in two stages. During the first irradiation, 1980 to 1986, the rods were part of Ringhals assembly 6477 and an approximate rod averaged burn-up of 31 MWd/kg UO 2 was reached. The rods were then placed into fuel assembly 9902 for a second period of irradiation from 1986 to 1992. The location of the fuel rods 33-25065 and 33-25046 in this assembly were in positions 9902/D and 9902/E4 respectively. A final rod averaged burn-up of 52 MWd/kg UO 2 was achieved. The burn-up at the location of the Halden segments was estimated as 59 MWd/kg UO 2 , well beyond the formation of High Burn-up Structure (Hobs) formation at the pellet rim. At the rim, the burn-up was estimated as 130 MWd/kg UO 2 . After commercial irradiation, PIE was performed at Studsvik. Inner and outer clad oxide thickness measurements were 42 and 5 microns respectively. The measured cold rod diameter varied between 12.20 and 12.25 mm, thus only a small amount of creep-down had occurred from the original diameter of 12.25 mm. Cold gap measurements were taken by diametral compression of the clad onto the fuel. The stiffness changes twice during these measurements, the first (relocated gap) associated with the onset of pellet fragment movement, the second (compressed gap) when the fragments are together and the pellet is compressed. For these rods, the compressed diametral gap was measured as 30 microns. This is in agreement with the pellet and cladding being in contact during the final irradiation cycle, i.e., at ∼12 kW/m. FGR measurements were made after puncturing and values of 2.5%-3.3% were calculated from the extracted gas. The uncertainty is due to different methods of calculation. Ceramography showed a normal crack pattern and no evidence of

  10. Solubility of unirradiated UO2 fuel in aqueous solutions. Comparison between experimental and calculated (EQ3/6) data

    International Nuclear Information System (INIS)

    Ollila, K.

    1995-11-01

    The solubility behaviour of unirradiated UO 2 pellets was studied under oxic (air-saturated) and anoxic (N 2 ) conditions in deionized water, in sodium bicarbonate solutions with varying bicarbonate content (60 - 600 ppm), in Allard groundwater simulating granitic fresh groundwater conditions, and in bentonite water simulating the effects of bentonite on granitic fresh groundwater (25 deg C). The release of uranium was measured during static batch dissolution experiments of long duration (2-6 years). A comparison was made with the theoretical solubility data calculated with the geochemical code EQ3/6 in order to evaluate solubility (steady state) limiting factors. (orig.) (26 refs., 32 figs., 13 tabs.)

  11. The burnup dependence of light water reactor spent fuel oxidation

    International Nuclear Information System (INIS)

    Hanson, B.D.

    1998-07-01

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

  12. Advanced fuel pellet materials and designs for water cooled reactors. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    2004-10-01

    This meeting was the second IAEA meeting on this subject. The first was held in 1996 in Tokyo, Japan. They are all part of a cooperative effort through the Technical Working Group on Water Reactor Fuel Performance and Technology (TWGFPT) of IAEA, with a series of three further meetings organized by CEA, France and co-sponsored by the IAEA and OECD/NEA. In the seven years since the first meeting took place, the demands on fuel duties have increased, with higher burnup, longer fuel cycles and higher temperatures. This places additional demands on fuel performance to comply with safety requirements. Criteria relative to fuel components, i.e. pellets and fuel rod column, require limiting of fission gas release and pellet-cladding interaction (PCI). This means that fuel components should maintain the composite of rather contradictory properties from the beginning until the end of its in-pile operation. Fabrication and design tools are available to influence, and to some extent, to ensure desirable in-pile fuel properties. Discussion of these tools was one of the objectives of the meeting. The second objective was the analysis of fuel characteristics at high burnup and the third and last objective was the discussion of specific feature of MOX and urania gadolinia fuels. Sixty specialists in the field of fuel fabrication technology attended the meeting from 18 countries. Twenty-five papers were presented in five sessions covering all relevant topics from the practices and modelling of fuel fabrication technology to its optimization. Eight papers were presented in session 'Optimization of fuel fabrication technology' which all were devoted to fuel fabrication technology. They mostly treated methods for optimizing fuel manufacturing processes, but gave also a good overview on nuclear fabrication needs and capabilities in different countries. During Session 'UO 2 , MOX and UO 2 -Gd 2 O 3 pellets with additives', six papers were presented in this session, which dealt mainly

  13. Nuclear fuel behaviour modelling at high burnup and its experimental support. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    2001-07-01

    of MOX behaviour up to that for UO 2 fuel. There appears to be a good consensus on how MOX fuel performance differs from UO 2 , and on the issues that need to be addressed to achieve higher burnups. The final sessions of the TCM considered the current status of integrated fuel behaviour codes and the challenges for higher burnup modelling. The meeting provided a valuable forum for a review of the state-of-the-art. Presentations were given on a number of existing codes and others under development, covering PWR, WWER, BWR and CANDU fuel performance. Some specialised methods for specific advanced fuel types were also discussed. Recommendations on future work in the area of fission gas release; clad modelling; and MOX fuel modelling are included

  14. Densification behaviour of UO2 in six different atmospheres

    International Nuclear Information System (INIS)

    Kutty, T.R.G.; Hegde, P.V.; Khan, K.B.; Basak, U.; Pillai, S.N.; Sengupta, A.K.; Jain, G.C.; Majumdar, S.; Kamath, H.S.; Purushotham, D.S.C.

    2002-01-01

    The shrinkage behaviour of UO 2 has been studied using a dilatometer in various atmospheres of Ar, Ar-8%H 2 , vacuum, CO 2 , commercial N 2 and N 2 +1000 ppm of O 2 . The onset of shrinkage occurs at around 300-400 deg. C lower in oxidizing atmospheres such as CO 2 , commercial N 2 and N 2 +1000 ppm O 2 compared to that in reducing or inert atmospheres. Shrinkage behaviour of UO 2 is almost identical in Ar, Ar-8%H 2 and vacuum. The shrinkage in N 2 +1000 ppm O 2 begins at a lower temperature than that in the commercial N 2 . The mechanism of sintering in the reducing, inert and vacuum atmospheres is explained by diffusion of uranium vacancies and that in the oxidizing atmospheres by cluster formation

  15. Binding energy and formation heat of UO2

    International Nuclear Information System (INIS)

    Almeida, M.R. de; Veado, J.T.; Siqueira, M.L. de

    The Born-Haber cycle is utilized for the calculation of the heat of formation of UO 2 , on the assumption that the binding energy is predominantly ionic in character. The ionization potentials of U and the repulsion energy are two critical values that influence calculations. Calculations of the ionization potentials with non-relativistic Hartree-Fock-Gaspar-Kohn-Sham approximation are presented [pt

  16. Pressure-induced weak ferromagnetism in uranium dioxide, UO2

    International Nuclear Information System (INIS)

    Sakai, H; Kato, H; Tokunaga, Y; Kambe, S; Walstedt, R E; Nakamura, A; Tateiwa, N; Kobayashi, T C

    2003-01-01

    The dc magnetization of insulating UO 2 under high pressure up to ∼1 GPa has been measured using a piston-cylinder cell. Pressure-induced weak ferromagnetism appeared at low pressure (∼0.2 GPa). Both the remanent magnetization and the coercive force increase as pressure increases. This weak ferromagnetism may come from spin canting or from uncompensated moments around grain boundaries

  17. Basic evaluation on nuclear characteristics of BWR high burnup MOX fuel and core

    International Nuclear Information System (INIS)

    Nagano, M.; Sakurai, S.; Yamaguchi, H.

    1997-01-01

    MOX fuel will be used in existing commercial BWR cores as a part of reload fuels with equivalent operability, safety and economy to UO 2 fuel in Japan. The design concept should be compatible with UO 2 fuel design. High burnup UO 2 fuels are being developed and commercialized step by step. The MOX fuel planned to be introduced in around year 2000 will use the same hardware as UO 2 8 x 8 array fuel developed for a second step of UO 2 high burnup fuel. The target discharge exposure of this MOX fuel is about 33 GWd/t. And the loading fraction of MOX fuel is approximately one-third in an equilibrium core. On the other hand, it becomes necessary to minimize a number of MOX fuels and plants utilizing MOX fuel, mainly due to the fuel economy, handling cost and inspection cost in site. For the above reasons, it needed to developed a high burnup MOX fuel containing much Pu and a core with a large amount of MOX fuels. The purpose of this study is to evaluate basic nuclear fuel and core characteristics of BWR high burnup MOX fuel with batch average exposure of about 39.5 GWd/t using 9 x 9 array fuel. The loading fraction of MOX fuel in the core is within a range of about 50% to 100%. Also the influence of Pu isotopic composition fluctuations and Pu-241 decay upon nuclear characteristics are studied. (author). 3 refs, 5 figs, 3 tabs

  18. 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

  19. Automation system for production of UO2 granules

    International Nuclear Information System (INIS)

    Swaminathan, N.; Setty, C.R.P.; Banerjee, P.K.; Husnain, G.; Rao, K.C.M.; Satyanarayana, A.

    1990-01-01

    Precompaction of UO 2 powder into slugs and granulation of the slugs were used to be carried out in two different work centres involving manual loading/handling of powder and compacts which resulted in a very high level of air-borne activity. This has been simplified by integrating both the operations into one work centre on both the precompaction presses. In the present system, UO 2 powder is transferred to feed hopper through the use of high vac. feeder. The powder in metered quantities is fed into the shoe by deploying screw feeder driven by a compact hydraulic motor. The die cavity is filled with just the right quantity of powder to prevent spillage. The compacts are pushed on to the granulator through a set of guides mounted on the die platform. The granulated powder is made to pass through Vibro screen for separating the fines before collecting in a replaceable S.S. Container. This container is mounted on the final compacting press by using job crane installed on the press. The replaceable container handling facility drastically cuts down the manual handling of UO 2 granules and also eliminates spillage, air borne activity. The development and fabrication of hydraulically operated screw feeder, feed shoe, replaceable container and the job crane structure etc., were completely carried out at Nuclear Fuel Complex, Hyderabad. Paper deals in detail the design of the system developed, present operational experiences and further improvements planned. (author). 6 figs

  20. An evaluation of the influence of fuel design parameters and burnup on pellet/cladding interaction for boiling water reactor fuel rod through in-core diameter measurement

    International Nuclear Information System (INIS)

    Yanagisawa, K.

    1986-01-01

    The influence of design parameters and burning on pellet/cladding interaction (PCI) of current boiling water reactor fuel rods was studied through in-core diameter measurement. Thinner cladding and a smaller diametral gap enhanced the PCI during startup. At constant power, fuel with SiO 2 added greatly reduced PCI due to relaxation. The fuel with a small grain size greatly reduced PCI due to densification. Preirradiation of rods up to 23 MWd/kgU caused a large PCI not only in a small gap but also in a large gap rod. Relaxation and permanent deformation was small. In the power increase experiment, one rod experienced PCI failure. The spurt times of coolant radioactivity coincided well with the sudden drop of cladding axial strain and marked crack opening at the rod surface. The estimated hoop stress predicted by FEMAXI-III was 350 MPa at the failure

  1. The treatment of large quantities of high fluorin contents UO2 by ammonium double uranate (ADU) techniques

    International Nuclear Information System (INIS)

    Wang Bangwu; Chen Ying

    2010-01-01

    The paper has discussed the sinter action of UO 2 in low temperature. The study indicates the over hot part of UO 2 by the deoxidization hot of oxidation uranate mostly results in the sinter in the process of trans form ADU into UO 2 . The UO 2 settling times in kiln little influences the sinter performance of UO 2 in the same condition of high fluorin contents UO 2 returning kiln, and high fluorin contents UO 2 returning kiln does not sinter UO 2 again. Experiment on large quantities of high fluorin contents UO 2 by Ammonium Double Uranate (ADU) techniques direct returning kiln, the result shows the sinter performance of UO 2 doesn't drop in the process of high fluor in contents UO 2 direct returning kiln, and the performance of UO 2 can meet specification. (authors)

  2. Methods for assessing homogeneity in ThO2--UO2 fuels (LWBR Development Program)

    International Nuclear Information System (INIS)

    Berman, R.M.

    1978-06-01

    ThO 2 -UO 2 solid solutions fabricated as LWBR fuel pellets are examined for uniform uranium distribution by means of autoradiography. Kodak NTA plates are used. Images of inhomogeneities are 29 +- 10 microns larger in diameter than the high-urania segregations that caused them, due to the range of alpha particles in the emulsion, and an appropriate correction must be made. Photographic density is approximately linear with urania content in the region between underexposure and overexposure, but the slope of the calibration curve varies with aging and growth of alpha activity from the parasitic 232 U and its decomposition products. A calibration must therefore be performed using two known points--the average photographic density (corresponding to the average composition) and an extrapolated background (corresponding to zero urania). As part of production pellet inspection, plates are evaluated by inspectors, who count segregations by size classes. This is supplemented by microdensitometer scans of the autoradiograph and by electron probe studies of the original sample if apparent homogeneity is marginal

  3. U02 pellets surface properties and environmental conditions effects on the wet adsorption

    International Nuclear Information System (INIS)

    Junqueira, Fabio da S.; Carnaval, Joao Paulo R.

    2013-01-01

    Angra power plants fuels are made bye en riche uranium dioxide (UO 2 ) pellets which are assembled inside metal tubes. These tubes are welded and arranged in order to perform the final product, the fuel assembly. The UO 2 pellets have a specified humidity tolerance designed to comply with security and performance requirements when working under operating conditions in the reactor. This work intends to verify the pellet opened porosity and the environmental conditions (relative humidity and temperature) influence on the wet adsorption by UO 2 pellet. The work was done in 2 parts: Firstly, pallets groups from 3 opened porosity levels were tested under a fixed relative humidity, temperature and time. In the second part of the work, the most critical pallet group upon wet adsorption was tested under different relative humidity and temperature conditions, regarding design of experiments. The opened porosity and environmental conditions tests allowed the evolution of the wet adsorption by the UO 2 pallet. (author)

  4. Metallographic examination of (uth) O2 and UO2 fuel tested in power ramp conditions in triga reactor

    International Nuclear Information System (INIS)

    Ioncescu, M.; Uta, O.

    2015-01-01

    The purpose of this paper is to determine the behavior of two fuel experimental elements (EC1 and EC2), by destructive post-irradiation examination. The fuel elements were mounted inside a pattern port, one in extension of the other and irradiated in power ramp conditions in order to check their behavior. Fuel element 1 (EC1) contains (UTh)O''2 pellet, and other one (EC2) UO''2 pellet. The results of destructive post-irradiation examination are evidenced by metallographic and ceramographic analyses. The data obtained from the post-irradiation examinations are used, first to confirm the security, reliability and nuclear fuel performance, and second, for the development of CANDU fuel. The results obtained by destructive examinations regarding the integrity, sheath hydrating and oxidation as well as the structural modifications are typical for fuel elements tested in power ramp conditions. (authors)

  5. Determination of UO2 little quantity in UF4 by X-rays diffraction

    International Nuclear Information System (INIS)

    Costa, M.I.; Sato, I.M.; Imakuma, K.

    1977-01-01

    In the fluorination process, the final product UF 4 contain different levels of UO 2 as a contaminant. A routine method for quantitative analysis by x-ray diffraction has been developed. Standard curves have been plotted using mixtures of UO 2 /UF 4 with measures of intensity of (III) peak of UO 2 by the step scanning process. The integrated intensity versus UO 2 concentration curves present a linear behavior in the range from 0 to 4%. A good reprodutibility of measuring process has been observed through statistical analysis which permits to determine low fractions of UO 2 in UF 4 with +- 0,08% of accuracy [pt

  6. Production and release of the fission gas in (Th U)O2 fuel rods

    International Nuclear Information System (INIS)

    Dias, Marcio S.

    1982-06-01

    The volume, composition and release of the fission gas products were caculated for (Th, U)O 2 fuel rods. The theorectical calculations were compared with experimental results available on the literature. In ThO 2 + 5% UO 2 fuel rods it will be produced approximated 5% more fission gas as compared to UO 2 fuel rods. The fission gas composition or Xe to Kr ratio has showed a decreasing fuel brunup dependence, in opposition to that of UO 2 . Under the same fuel rod operational conditions, the (Th, U)O 2 fission gas release will be smaller as compared to UO 2 . This behaviour of (Th, U)O 2 fuel comes from smallest gas atom difusivity and higher activation energies of the processes that increase the fission gas release. (Author) [pt

  7. Behaviour of the UO2/clayey water. A spectroscopic approach

    International Nuclear Information System (INIS)

    Guilbert, S.

    2000-05-01

    This work deals with the disposal of spent nuclear fuels in deep geological layers. After three years of irradiation, these fuels are constituted of 95 % UO 2 . It is then indispensable to know the leaching behaviour of this solid because ground waters are the main agents of dispersion to biosphere of the radioelements contained in these fuels. This work includes alteration tests carried out with a device allowing to synthesize a clayey water equilibrated with a partial pressure in CO 2 in oxidizing or reducing conditions. After the tests, the solid and the solution have been characterized in order to establish a balance of the alteration. The UO 2 matrix has been characterized by XPS. The uranium in solution has been titrated by ICP-MS. In oxidizing conditions, after some weeks, the dissolution velocity of UO 2 has stabilized around 3*10 11 mol/m 2 .s. This velocity is of 4*10 12 mol/m 2 .s in a reducing medium. The uranium concentrations in the oxidized water are of about 2*10 4 mol/l after two years of leaching. After 33 days of alteration in a reducing medium, the uranium amount is of 3*10 6 mol/l. The XPS technique has revealed a superficial and progressive oxidation of the uranium(IV) and the formation of U-OH bonds in the oxidizing medium. A U(VI)/U(IV) ratio has been determined by this technique. It has stabilized around 2 in some weeks. In reducing conditions, this ratio is stable and is of about 0.5. Modeling tools have allowed to propose a class of solids potentially able to control the uranium solubility. In oxidizing conditions, the uranyl hydrates (schoepite) evolve towards uranyl silicates which are thermodynamically more stable. In reducing conditions, a control of the uranium concentration in solution by U 4 O 9 is probable. (O.M.)

  8. Development of uranium dioxide fuel pellets with addition of beryllium oxide for increasing of thermal conductivity

    International Nuclear Information System (INIS)

    Queiroz, Carolinne Mol; Ferreira, Ricardo Alberto Neto

    2011-01-01

    The CDTN - Centro de Desenvolvimento de Tecnologia Nuclear presents a project named 'Beryllium Project' viewing to increasing the thermal conductivity of UO 2 fuel pellets, increasing the lifetime of those pellets in the reactor, generating a greater economy. This increase of conductivity is obtained by means of Be O addition to the UO 2 fuel pellets, which is very used for the production of nuclear energy. The UO 2 pellets however present a thermal conductivity relatively low, generating a high temperature gradient between the center and his side surface. The addition of beryllium oxide, with higher thermal conductivity gives pellets which will present lower temperature gradient and, consequently, more durability and better utilization of energy potential of the pellet in the reactor. (author)

  9. Mechanism for transient migration of xenon in UO2

    International Nuclear Information System (INIS)

    Liu, X.-Y.; Uberuaga, B. P.; Andersson, D. A.; Stanek, C. R.; Sickafus, K. E.

    2011-01-01

    In this letter, we report recent work on atomistic modeling of diffusion migration events of the fission gas product xenon in UO 2 nuclear fuel. Under nonequilibrium conditions, Xe atoms can occupy the octahedral interstitial site, in contrast to the thermodynamically most stable uranium substitutional site. A transient migration mechanism involving Xe and two oxygen atoms is identified using basin constrained molecular dynamics employing a Buckingham type interatomic potential. This mechanism is then validated using density functional theory calculations using the nudged elastic band method. An overall reduction in the migration barrier of 1.6-2.7 eV is obtained compared to vacancy-mediated diffusion on the uranium sublattice.

  10. UO2 microspheres obtainment through the internal gelation methods

    International Nuclear Information System (INIS)

    Sterba, M.E.; Gomez Constenla, A.

    1987-01-01

    UO 2 microspheres obtainment process through the internal gelation method which allows the spheres' obtainment of uniform size is detailed herein, varying the same among 0.3 and 1.7 mm of diameter. The sintered density reaches 10.78 g/cm 3 , permitting the fuels fabrication dispersed and vibro-compacted fuels. The trichloroethylene use implementation as gelation agent is described, thus reducing the number of stages in the microspheres fabrication. At the same time, the uranium sun composition has been modified so as to be compatible with the use solvent. (Author)

  11. 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)

  12. Thermal conductivity and thermal diffusivity of solid UO2

    International Nuclear Information System (INIS)

    Fink, J.K.; Chasanov, M.G.; Leibowitz, L.

    1981-06-01

    New equations for the thermal conductivity of solid UO 2 were derived based upon a nonlinear least squares fit of the data available in the literature. In the development of these equations, consideration was given to their thermodynamic consistency with heat capacity and density and theoretical consistency with enthalpy and heat capacity. Consistent with our previous treatment of enthalpy and heat capacity, 2670 K was selected as the temperature of a phase transition. A nonlinear equation, whose terms represent contributions due to phonons and electrons, was selected for the temperature region below 2670 K. Above 2670 K, the data were fit by a linear equation

  13. Behaviour of short-lived fission products within operating UO2 fuel elements

    International Nuclear Information System (INIS)

    Hastings, I.J.; Hunt, C.E.L.; Lipsett, J.J.

    1983-01-01

    We have carried out experiments using a ''sweep gas'' technique to determine the behaviour of short-lived fission products within operating, intact UO 2 fuel elements. The Zircaloy-4-clad elements were 500 mm long and contained fuel of density 10.65-10.71 Mg/m 3 . A He-2% H 2 carrier gas swept gaseous or volatile fission products out of the operating fuel element past a gamma spectrometer for measurement. In tests at linear powers of 45 and 60 kW/m to maximum burnups of 70 MW.h/kg U, the species measured directly at the spectrometer were generally the short-lived xenons and kryptons. We did not observe iodine or bromine during normal operation. However, we have deduced the behaviour of I-133 and I-135 from the decay of Xe-133 and Xe-135 during reactor shutdowns. Plots of R/B (released/born) against lambda (decay constant) or effective lambda for all isotopes observed at 45 and 60 kW/m show that a line of slope -0.5, corresponding with diffusion kinetics, is a good fit to the measured xenon and krypton data. Our inferred release of iodine fits the same line. From this we can extrapolate to an R/B for I-131 of about 5x10 -3 . The ANS 5.4 release correlation gives calculated results in good agreement with our measurements. (author)

  14. On the behaviour of intragranular fission gas in UO2 fuel

    International Nuclear Information System (INIS)

    Loesoenen, Pekka

    2000-01-01

    Data obtained from the literature concerning the behaviour of intragranular gas in sintered LWR UO 2 fuel are reviewed comprehensively. The characteristics of single gas atoms and bubbles, as a function of irradiation time, temperature, fission rate and burn-up are described, based on the reported experimental data. The relevance of various phenomena affecting gas behaviour is evaluated. The current status of modelling of the behaviour of intragranular gas is considered in light of the present findings. Simple calculations showed that the conventional approximation for the effective diffusion coefficient does not adequately describe the gas behaviour under transient conditions, when bubble coarsening plays a key role in the release. The difference in the release fraction, compared with a more mechanistic approach, could be as large as 30%. A number of recommendations regarding possible defects in the mechanistic approach to modelling of intragranular gas are highlighted. The lack of an effective numerical method for solving the set of relevant non-linear differential equations is shown to be a serious obstacle in implementing the mechanistic models for fission gas release (FGR), in integral fuel performance codes

  15. Physical models for high burnup fuel

    International Nuclear Information System (INIS)

    Kanyukova, V.; Khoruzhii, O.; Likhanskii, V.; Solodovnikov, G.; Sorokin, A.

    2003-01-01

    In this paper some models of processes in high burnup fuel developed in Src of Russia Troitsk Institute for Innovation and Fusion Research are presented. The emphasis is on the description of the degradation of the fuel heat conductivity, radial profiles of the burnup and the plutonium accumulation, restructuring of the pellet rim, mechanical pellet-cladding interaction. The results demonstrate the possibility of rather accurate description of the behaviour of the fuel of high burnup on the base of simplified models in frame of the fuel performance code if the models are physically ground. The development of such models requires the performance of the detailed physical analysis to serve as a test for a correct choice of allowable simplifications. This approach was applied in the SRC of Russia TRINITI to develop a set of models for the WWER fuel resulting in high reliability of predictions in simulation of the high burnup fuel

  16. A molecular dynamics study of solid and liquid UO2

    International Nuclear Information System (INIS)

    Sindzingre, P.; Gillan, M.J.

    1988-01-01

    We present an extensive series of molecular dynamics simulations of UO 2 in the solid and liquid states, in which we calculate the ionic diffusion coefficients and some of the important thermodynamic quantities. The simulations are based on a rigid-ion model derived from the new shell model potentials of Jackson and co-workers and make use of recently developed constant-pressure and constant-temperature techniques. The simulations confirm that UO 2 is an oxygen superionic conductor, as suggested by recent neutron scattering experiments. The temperature of the diffuse transition to the superionic regime is in satisfactory agreement with experiment, as is the melting point of the model system. The thermal expansion coefficient, specific heat and bulk modulus for the solid agree well with experiment below about 2500 K but are less satisfactory near the melting point; we suggest that the differences may be due to the effect of electronic excitations. The volume increase on melting and thermodynamic quantities of the liquid are sensitive to details of the inter-ionic potentials and are in only fair agreement with experiment. (author)

  17. The production of sinterable UO2 from AUC

    International Nuclear Information System (INIS)

    Chang, I.S.; Do, J.B.; Choi, Y.D.; Park, M.H.; Yun, H.H.; Kim, E.H.; Kim, Y.W.

    1982-01-01

    Fluidization, feeding and discharging, and mixing of fine particles (-up to 40μ in diameter) in fluidized bed reactor has been examined. The degree of conversion has been estimated using the kinetic data differential scanning colorimetry(DSC) and thermogravimetic analysis (TGA) of ammonium uranyl carbonate (AUC) and residence time distribution data. Satisfactory operation is obtained with a sintered ceramic distributor and filters. The reactor equilvalent to approximately 1.1-1.3 stages. Thermal analysis of AUC in hydrogen atmosphere shows that the decomposition of AUC to UO 3 at 200degC is followed by reduction of UO 3 to UO 2 in two steps in the range between 400degC and 500degC and the complete conversion to UO 2 takes two minutes at 550degC. The overall conversion of above 99.5% in the fluidized bed reactor is estimated with 40 minutes of a mean particle residence time at 600degC. (Author)

  18. Irradiation effects in UO2 and CeO2

    International Nuclear Information System (INIS)

    Ye, Bei; Oaks, Aaron; Kirk, Mark; Yun, Di; Chen, Wei-Ying; Holtzman, Benjamin; Stubbins, James F.

    2013-01-01

    Single crystal CeO 2 , as a surrogate material to UO 2 , was irradiated with 500 keV xenon ions at 800 °C while being observed using in situ transmission electron microscopy (TEM). Experimental results show the formation and growth of defect clusters including dislocation loops and cavities as a function of increasing atomic displacement dose. At high dose, the dislocation loop structure evolves into an extended dislocation line structure, which appears to remain stable to the high dose levels examined in this study. A high concentration of cavities was also present in the microstructure. Despite high atomic displacement doses, the specimen remained crystalline to a cumulated dose of 5 × 10 15 ions/cm 2 , which is consistent with the known stability of the fluorite structure under high dose irradiation. Kinetic Monte Carlo calculations show that oxygen mobility is substantially higher in hypo-stoichiometric UO 2 /CeO 2 than hyper-stoichiometric systems. This result is consistent with the ability of irradiation damage to recover even at intermediate irradiation temperatures

  19. Nuclear Energy Research Initiative. Development of a Stabilized Light Water Reactor Fuel Matrix for Extended Burnup

    International Nuclear Information System (INIS)

    BD Hanson; J Abrefah; SC Marschman; SG Prussin

    2000-01-01

    The main objective of this project is to develop an advanced fuel matrix capable of achieving extended burnup while improving safety margins and reliability for present operations. In the course of this project, the authors improve understanding of the mechanism for high burnup structure (HBS) formation and attempt to design a fuel to minimize its formation. The use of soluble dopants in the UO 2 matrix to stabilize the matrix and minimize fuel-side corrosion of the cladding is the main focus

  20. Optimal rate of power increase in nuclear fuel. Pellet behaviour under dynamic conditions

    International Nuclear Information System (INIS)

    Karlsson, B.G.

    1976-05-01

    A mathematical model has been worked out for the determination of the optium power escalation rate for nuclear power plants from the view-pint of fuel integrity. The model calculates the stress and strain transients in the pellet-cladding system with rapid power increase. No burnup effects are included due to the short time scale involved. An elastic solution has been transposed to a linear viscoelastic one using the correspondence principle. The cladding has however been treated under the programme assumptions as purely elastic. The fuel material has been assumed to be completely relaxed prior to the power transient. Radial cracking is included. The UO 2 -material distortion has been assumed to be linear viscoelastic, while the dilation is assumed as elastic. The system has been treated assuming plane strain since friction between the pellet and the cladding is large with practical burnsups, and the pellet column can be regarded as infinitely long, compared to the diameter of the pellet. The results of the calculations made show that under the above assumptions the clad stress is independent of the rate of power increase in the pellet. Scince this result is in opposition to general opinion an experimental programme was performed in order to test the results of the model. These results were confirmed. The occurance of clad failures in practice is not determined purely by clad straining. Current thought pays attention to the influence of e.g. stress-corrosion phenomena as significant. The programme reported here pays no attention such-like effects, or the effects of clad creep which could be of considerable significance with local deformations. These later effects are receiving attention in work now being initiated at the Department.(author)

  1. Electrochemical characterisation of CaCl2 deficient LiCl-KCl-CaCl2 eutectic melt and electro-deoxidation of solid UO2

    Science.gov (United States)

    Sri Maha Vishnu, D.; Sanil, N.; Mohandas, K. S.; Nagarajan, K.

    2016-03-01

    The CaCl2 deficient ternary eutectic melt LiCl-KCl-CaCl2 (50.5: 44.2: 5.3 mol %) was electrochemically characterised by cyclic voltammetry and polarization techniques in the context of its probable use as the electrolyte in the electrochemical reduction of solid UO2 to uranium metal. Tungsten (cathodic polarization) and graphite (anodic polarization) working electrodes were used in these studies carried out in the temperature range 623 K-923 K. The cathodic limit of the melt was observed to be set by the deposition of Ca2+ ions followed by Li+ ions on the tungsten electrode and the anodic limit by oxidation of chloride ions on the graphite electrode (chlorine evolution). The difference between the onset potential of deposition of Ca2+ and Li+ was found to be 0.241 V at a scan rate of 20 mV/s at 623 K and the difference decreased with increase in temperature and vanished at 923 K. Polarization measurements with stainless steel (SS) cathode and graphite anode at 673 K showed the possibility of low-energy reactions occurring on the UO2 electrode in the melt. UO2 pellets were cathodically polarized at 3.9 V for 25 h to test the feasibility of electro-reduction to uranium in the melt. The surface of the pellets was found reduced to U metal.

  2. Photochemical assessment of UO2+2 complexation in Triton X-100 micellar system

    International Nuclear Information System (INIS)

    Das, S.K.; Ganguly, B.N.

    1994-01-01

    This is a report on the spectral characteristics of UO 2 +2 in the excited state in the Triton X-100 micellar medium. The downward curving of the Stern-Volmer plot explains the two kinds of populations of UO 2 +2 upon micellization. A blue shift of the quenched emission is ascribed due to the collisional encounter of UO 2 +2 with the head groups of Triton X-100. (author). 5 refs., 2 figs

  3. Fabrication and testing of ceramic UO2 fuel - I-III. Part I

    International Nuclear Information System (INIS)

    Novakovic, M.

    1961-12-01

    The task described consists of the following: fabrication of UO 2 with different granulation from uranyl nitrate by ammonia diuranate; determination of size and shape distributions of metal and ceramic powders; fabrication of sintered pressed samples UO 2 ; investigating the properties of sintered uranium dioxide dependent on the fabrication process; producing a vibrator for compacting UO 2 powder. This volume includes reports on the first two tasks

  4. Deformation behavior of UO2 at temperatures above 24000C

    International Nuclear Information System (INIS)

    Slagle, O.D.

    1978-08-01

    An experimental system was developed for measuring the high-temperature creep rates of ceramic nuclear fuels to temperatures near their melting points. The results of a series of experiments carried out on UO 2 at temperatures above 2400 0 C are reported. The strain rate was found to be proportional to the 5.7 power of the stress while activation energies ranged from 250 to 340 Kcal/mole. An expression for describing the primary creep was derived from the initial time dependence of the deformation after stress application. A technique for studying the hot pressing behavior at 2580 0 C was devised but no definitive results were obtained from the first series of experiments. An empirical relationship is proposed for calculating the creep rates at very high temperatures

  5. Sorption of Np by UO2 under repository conditions

    Science.gov (United States)

    Kazakovskaya, T. V.; Zakharova, E. V.; Haire, M. J.

    2010-03-01

    This work is a part of the joint Russian - American Program on Beneficial Use of Depleted Uranium. The production of nuclear fuels results in the accumulation of large quantities of depleted uranium (DU) in the form of uranium hexafluoride (UF6), which is converted to uranium oxides. Depleted uranium dioxide (DUO2) can be used as a component of radiation shielding and as an absorbent for migrating radionuclides that may emerge from casks containing spent nuclear fuel (SNF) that are stored for hundreds of thousands of years in high-level wastes (HLW) and SNF repositories (e.g. Yucca Mountain Project). In this case DU oxides serve as an additional engineered chemical barrier. It is known that the primary radioisotope contributor to the calculated long-term radiation dose to the public at the Yucca Mountain SNF repository site boundary is neptunium-237 (237Np). This paper describes the sorption of 237Np in various media (deionized water and J-13 solution) by DUO2. Samples of DUO2 used in this work originated from the treatment of UF6 in a reducing media to form UO2(DUO2-1 at 600°C, DUO2-2 at 700°C, and DUO2-3 at 800°C). All species of DUO2 sorb Np(V) and Np(IV) from aqueous media. Equilibrium was achieved in 24 hours for Np(V) and in 2 hours for Np(IV). Np(V) sorption is accompanied with partial reduction of Np(V) to Np(IV) and vice versa. The sorption of Np(V) onto DUO2 surfaces is irreversible. The investigations on DUO2 transformations were performed under dynamic and static conditions. Under static conditions the solubility of the DUO2 samples in J-13 solution is considerably higher than in DW. When the pre-treatment temperature is decreased, the solubility of DUO2 samples raises regardless of the media. The experiments on interaction between DUO2 and aqueous media (DW and J-13 solution) under dynamic conditions demonstrated that during 30-40 days the penetration/filtration rate of DW and J-13 solution through a thin DUO2 layer decreased dramatically, and then

  6. 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

  7. Heat transfer from internally-heated molten UO2 pools

    International Nuclear Information System (INIS)

    Stein, R.P.; Baker, L. Jr.; Gunther, W.H.; Cook, C.

    1978-01-01

    Experimental measurements of heat transfer from internally heated pools of molten UO 2 have been obtained for two cell sizes: 10 cm x 10 cm and 20 cm x 20 cm. The experiments with the large cell have supported a previous conclusion from early small data that the measured downward heat fluxes are higher than would be expected on the basis of considerations of thermal convection. A convective model underpredicts the downward heat fluxes by a factor of 2.5 to 4.5 for all but one early experiment. Arbitrary assumptions of increased thermal conductivity do not account for the discrepancy. A single model based on internal thermal radiation heat transfer is able to account for the high values. The model uses the optically thick Rosseland approximation. Because of this, it is tentatively concluded that thermal radiation plays a dominant role in controlling the heat transfer from internally heated molted fuel

  8. Preliminary study of determination of UO2 grain size using X-ray diffraction method

    International Nuclear Information System (INIS)

    Mulyana, T.; Sambodo, G. D.; Juanda, D.; Fatchatul, B.

    1998-01-01

    The determination of UO 2 grain size has accomplished using x-ray diffraction method. The UO 2 powder is obtained from sol-gel process. A copper target as radiation source in the x-ray diffractometer was used in this experiment with CμKα characteristic wavelength 1.54433 Angstrom. The result indicate that the UO 2 mean grain size on presintered (temperature 800 o C) has the value 456.8500 Angstrom and the UO 2 mean grain size on sintered (temperature 1700 o C) has value 651.4934 Angstrom

  9. UO2-7%Gd2O3 fuel process development by mechanical blending with reprocessing of waste products and usage of densification additive

    International Nuclear Information System (INIS)

    Santos, Lauro Roberto dos

    2009-01-01

    In the nuclear fuel cycle, reprocessing and storage of 'burned' fuels, either temporary or permanent, demand high investments and, in addition, can potentially generate environmental problems. A strategy to decrease these problems is to adopt measures to reduce the amount of waste generated. The usage of integrated burnable poison based on gadolinium is a measure that contributes to achieve this goal. The reason to use burnable poison is to control the neutron population in the reactor during the early life of the fresh reactor core or the beginning of each recharging fuel cycle, extending its cycle duration. Another advantage of using burnable poison is to be able to operate the reactor with higher burning rate, optimizing the usage of the fuel. The process of manufacturing UO 2 -Gd 2 O 3 integrated burnable fuel poison generates waste that, as much as possible, needs to be recycled. Blending of Gd 2 O 3 in UO 2 powder requires the usage of a special additive to achieve the final fuel pellet specified density. The objective of this work is to develop the process of obtaining UO 2 - 7% Gd 2 O 3 integrated burnable poison using densification additives, aluminum hydroxide (Al(OH)3), and reprocessing manufacturing waste products by mechanical blending. The content of 7%- Gd 2 O 3 is based on commercial PWR reactor fuels - Type Angra 2. The results show that the usage of Al(OH) 3 as an additive is a very effective choice that promotes the densification of fuel pellets with recycle up to 10%. Concentrations of 0,20 % of Al(OH) 3 were found to be the indicated amount on an 7 industrial scale, specially when the recycled products come from U 3 O 8 obtained by calcination of sintered pellets. This is particularly interesting because it is following the steps of sintering and rectifying of the pellets, which is generating the largest amounts of recycled material. (author)

  10. UO2-7%Gd2O3 fuel process development by mechanical blending with reprocessing of waste products and usage of densification additive

    International Nuclear Information System (INIS)

    Santos, Lauro Roberto dos

    2009-01-01

    In the nuclear fuel cycle, reprocessing and storage of 'burned' fuels, either temporary or permanent, demand high investments and, in addition, can potentially generate environmental problems. A strategy to decrease these problems is to adopt measures to reduce the amount of waste generated. The usage of integrated burnable poison based on gadolinium is a measure that contributes to achieve this goal. The reason to use burnable poison is to control the neutron population in the reactor during the early life of the fresh reactor core or the beginning of each recharging fuel cycle, extending its cycle duration. Another advantage of using burnable poison is to be able to operate the reactor with higher burning rate, optimizing the usage of the fuel. The process of manufacturing UO 2 -Gd 2 O 3 integrated burnable fuel poison generates waste that, as much as possible, needs to be recycled. Blending of Gd 2 O 3 in UO 2 powder requires the usage of a special additive to achieve the final fuel pellet specified density. The objective of this work is to develop the process of obtaining UO 2 - 7% Gd 2 O 3 integrated burnable poison using densification additives, aluminum hydroxide (Al(OH) 3 ), and reprocessing manufacturing waste products by mechanical blending. The content of 7%- Gd 2 O 3 is based on commercial PWR reactor fuels - Type Angra 2. The results show that the usage of Al(OH) 3 as an additive is a very effective choice that promotes the densification of fuel pellets with recycle up to 10%. Concentrations of 0,20 % of Al(OH) 3 were found to be the indicated amount on an industrial scale, specially when the recycled products come from U 3 O 8 obtained by calcination of sintered pellets. This is particularly interesting because it is following the steps of sintering and rectifying of the pellets, which is generating the largest amounts of recycled material. (author)

  11. Hot vacuum outgassing to ensure low hydrogen content in MOX fuel pellets for thermal reactors

    International Nuclear Information System (INIS)

    Majumdar, S.; Nair, M.R.; Kumar, Arun

    1983-01-01

    Hot vacuum outgassing treatment to ensure low hydrogen content in Mixed Oxide Fuel (MOX) pellets for thermal reactors has been described. Hypostoichiometric sintered MOX pellets retain more hydrogen than UO 2 pellets. The hydrogen content further increases with the addition of admixed lubricant and pore formers. However, low hydrogen content in the MOX pellets can be ensured by a hot vacuum outgassing treatment at a temperature between 773K to 823K for 2 hrs. (author)

  12. Electrochemical characterisation of CaCl2 deficient LiCl–KCl–CaCl2 eutectic melt and electro-deoxidation of solid UO2

    International Nuclear Information System (INIS)

    Sri Maha Vishnu, D.; Sanil, N.; Mohandas, K.S.; Nagarajan, K.

    2016-01-01

    The CaCl 2 deficient ternary eutectic melt LiCl–KCl–CaCl 2 (50.5: 44.2: 5.3 mol %) was electrochemically characterised by cyclic voltammetry and polarization techniques in the context of its probable use as the electrolyte in the electrochemical reduction of solid UO 2 to uranium metal. Tungsten (cathodic polarization) and graphite (anodic polarization) working electrodes were used in these studies carried out in the temperature range 623 K–923 K. The cathodic limit of the melt was observed to be set by the deposition of Ca 2+ ions followed by Li + ions on the tungsten electrode and the anodic limit by oxidation of chloride ions on the graphite electrode (chlorine evolution). The difference between the onset potential of deposition of Ca 2+ and Li + was found to be 0.241 V at a scan rate of 20 mV/s at 623 K and the difference decreased with increase in temperature and vanished at 923 K. Polarization measurements with stainless steel (SS) cathode and graphite anode at 673 K showed the possibility of low–energy reactions occurring on the UO 2 electrode in the melt. UO 2 pellets were cathodically polarized at 3.9 V for 25 h to test the feasibility of electro-reduction to uranium in the melt. The surface of the pellets was found reduced to U metal. - Highlights: • Electrochemically characterized LiCl–KCl–CaCl 2 (50.5: 44.2: 5.3 mol %) melt by CV, LSV and polarization techniques. • Ca 2+ deposits first on tungsten working electrode followed by Li + . Cl − discharges on graphite to liberate chlorine gas. • Surface of UO 2 pellet reduced to U in the melt with low carbon contamination of melt. • Slow reduction of UO 2 due to slow kinetics and low solubility of oxide ions in the low temperature melt.

  13. WIMS/PANTHER analysis of UO2/MOX cores using embedded super-cells

    International Nuclear Information System (INIS)

    Knight, M.; Bryce, P.; Hall, S.

    2012-01-01

    This paper describes a method of analysing PWR UO 2 MOX cores with WIMS/PANTHER. Embedded super-cells, run within the reactor code, are used to correct the standard methodology of using 2-group smeared data from single assembly lattice calculations. In many other codes the weakness of this standard approach has been improved for MOX by imposing a more realistic environment in the lattice code, or by improving the sophistication of the reactor code. In this approach an intermediate set of calculations is introduced, leaving both lattice and reactor calculations broadly unchanged. The essence of the approach is that the whole core is broken down into a set of 'embedded' super-cells, each extending over just four quarter assemblies, with zero leakage imposed at the assembly mid-lines. Each supercell is solved twice, first with a detailed multi-group pin-by-pin solution, and then with the standard single assembly approach. Correction factors are defined by comparing the two solutions, and these can be applied in whole core calculations. The restriction that all such calculations are modelled with zero leakage means that they are independent of each other and of the core-wide flux shape. This allows parallel pre-calculation for the entire cycle once the loading pattern has been determined, in much the same way that single assembly lattice calculations can be pre-calculated once the range of fuel types is known. Comparisons against a whole core pin-by-pin reference demonstrates that the embedding process does not introduce a significant error, even after burnup and refuelling. Comparisons against a WIMS reference demonstrate that a pin-by-pin multi-group diffusion solution is capable of capturing the main interface effects. This therefore defines a practical approach for achieving results close to lattice code accuracy, but broadly at the cost of a standard reactor calculation. (authors)

  14. Technical evaluation of the direct denitration process to obtain ceramic-grade UO2 powders using microwaves

    International Nuclear Information System (INIS)

    Lorenzo, Viviana J.; Marchi, Daniel E.; Menghini, Jorge E.

    1999-01-01

    The direct denitration process to obtain ceramic-grade UO 2 powders using microwaves has been studied and developed at laboratory scale. Conditions were given to obtain powders apt for fuel pellets fabrication within the required specifications, where mechanical treatments before pressing are not necessary. This work describes the equipment used in the process, evaluates the necessary supply and waste generation and describes the characteristics of the product obtained, as well as the conditions for its fabrication. Results show that this method allows to reduce the volume of liquid wastes generated due to their partial re-utilization, simplifying their final disposal treatment, which, in addition to their operational advantages, make this method attractive from the economical point of view. (author)

  15. Accumulation of enriched uranium UO2F2 in ultrastructure as studied by electron microscopic autoradiography

    International Nuclear Information System (INIS)

    Zhu Shoupeng; Wang Yuanchang

    1992-01-01

    A study was made on the retention of soluble enriched uranium UO 2 F 2 in ultrastructure by electron microscopic autoradiography. The early dynamic accumulation of radioactivity in the body showed that enriched uranium UO 2 F 2 was mainly localized in kidneys, especially accumulated in epithelial cells of proximal convoluted tubules leading to degeneration and necrosis of the tubules. In liver cells, enriched uranium UO 2 F 2 at first deposited in nuclei of the cells and in soluble proteins of the plasma, and later accumulated selectively in mitochondria and lysosomes. On electron microscopic autoradiographic study it was shown that the dynamic retention of radioactivity of enriched uranium UO 2 F 2 in skeleton increased steadily through the time period of exposure. Enriched uranium UO 2 F 2 chiefly deposited in nuclei and mitochondria of osteoblasts as well as of osteoclasts

  16. 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.

  17. 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.

  18. Preliminary Results on a Contact between 4 kg of Molten UO2 and Liquid Sodium

    International Nuclear Information System (INIS)

    Amblard, M.

    1976-01-01

    The CORECT II Experiment consists in simulating the penetration of sodium into an assembly when the fuel is molten. In other words, it is a shock-tube type of experiment with dimensions representative of a full-scale assembly. the experiment consists in dropping a 100 litre column of sodium onto partially molten UO 2 . The following measurements are carried out in transient regime: - sodium velocity in the column; - pressure in the interaction chamber; - pressures at the bottom and at the top of a 5 m tube; - pressure in the argon blanket. The experimental parameters are: - the mass of UO 2 involved (about 4 or 7 kg of 80% molten UO 2 ); - the initial temperature of the sodium (up to 700 deg. C); - the pressure of the residual gas in the interaction chamber during the fall of the sodium; - the dimensions of the interaction chamber and the sodium supply tube; - the form of contact between the UO 2 and the sodium (the sodium may fall on partially liquid and settled UO 2 or on UO 2 pre-dispersed by forced trapping of sodium). To date, 6 tests have been performed. These tests have always resulted in fine fragmentation without any violent interaction. Since no knowledge is available on the change of grain size distribution with time, on the temperature of grain formation, and on the grain movement in the sodium, it is very difficult to interpret these UO 2 -Na tests. We intend to carry out more severe interaction tests on this experimental set-up, by eliminating as much as possible the non-condensable gas which cushions the mechanical impact of the sodium on the UO 2 (tests have shown that by strongly de-pressurizing the liquid UO 2 the fuel could be dispersed by boiling, and this effect should also improve the possibilities of a liquid/liquid contact). - by injecting a little sodium into the UO 2 to facilitate its dispersion in the coolant

  19. Fabrication of chamfered uranium-plutonium mixed carbide pellets

    International Nuclear Information System (INIS)

    Arai, Yasuo; Iwai, Takashi; Shiozawa, Kenichi; Handa, Muneo

    1985-10-01

    Chamfered uranium-plutonium mixed carbide pellets for high burnup irradiation test in JMTR were fabricated in glove boxes with purified argon gas. The size of die and punch in a press was decided from pellet densities and dimensions including the angle of chamfered parts. No chip or crack caused by adopting chamfered pellets was found in both pressing and sintering stages. In addition to mixed carbide pellets, uranium carbide pellets used as insulators were also successfully fabricated. (author)

  20. Hybrid pellets: an improved concept for fabrication of nuclear fuel

    International Nuclear Information System (INIS)

    Matthews, R.B.; Hart, P.E.

    1979-09-01

    The feasibility of fabricating fuel pellets using gel-derived microspheres as press feed was evaluated. By using gel-derived microspheres as press feed, the potential exists for eliminating dusty operations like milling, slugging, and granulation, from the pelleting process. The free-flowing character of the spheres should also result in limited dust generation during powder transport and pressing operations. The results of this study clearly demonstrate that fuel pellets can be successfully fabricated on a laboratory scale using UO 2 gel microspheres as press feed. Under moderate sintering conditions, 1,500 0 C for 4 h in Ar-4% H 2 , UO 2 pellets with densities up to 96% TD were fabricated. A range of pellet microstructures and densities were achieved depending on sphere forming and calcining conditions. Based on these results, a set of necessary sphere properties are suggested: O/U less than 2.20, crystallite size less than 500 A, specific surface area greater than 8 m 2 /g, and sphere size 200 and 400 μm. Preliminary attempts to fabricate ThO 2 and ThO 2 -UO 2 pellets using microspheres were unsuccessful because the requisite sphere properties were not achieved. Areas requiring additional development include: demonstration of the process on scaled-up equipment suitable for use in a remote fuel fabrication facility and evaluation of the irradiation performance of pellet fuels from gel-spheres

  1. Development of high burnup nuclear fuel technology

    International Nuclear Information System (INIS)

    Suk, Ho Chun; Kang, Young Hwan; Jung, Jin Gone; Hwang, Won; Park, Zoo Hwan; Ryu, Woo Seog; Kim, Bong Goo; Kim, Il Gone

    1987-04-01

    The objectives of the project are mainly to develope both design and manufacturing technologies for 600 MWe-CANDU-PHWR-type high burnup nuclear fuel, and secondly to build up the foundation of PWR high burnup nuclear fuel technology on the basis of KAERI technology localized upon the standard 600 MWe-CANDU- PHWR nuclear fuel. So, as in the first stage, the goal of the program in the last one year was set up mainly to establish the concept of the nuclear fuel pellet design and manufacturing. The economic incentives for high burnup nuclear fuel technology development are improvement of fuel utilization, backend costs plant operation, etc. Forming the most important incentives of fuel cycle costs reduction and improvement of power operation, etc., the development of high burnup nuclear fuel technology and also the research on the incore fuel management and safety and technologies are necessary in this country

  2. Study of effect of dynamic mill treatment on the compaction and sintering of UO2-5wt%CeO2

    International Nuclear Information System (INIS)

    Na, S. H.; Kim, S. H.; Kim, Y. K.; Lee, Y. W.; Yoo, M. J.

    2002-01-01

    Some properties (apparent density, green density, sintered density and grain size) of the simulated mixed oxide(UO 2 -5wt%CeO 2 ) prepared by using the dynamic mill newly developed were investigated. As the dynamic milling time increases, these properties increase. However, the increases of sintered density and grain size were saturated above 2 hrs and 4 hrs of dynamic milling treatment, respectively. It appeared that the dynamic mill has a similar capability to that of other milling methods, and hence can manufacture the pellets having desired properties by controlling milling time

  3. Fabrication, irradiation and post-irradiation examinations of MO2 and UO2 sphere-pac and UO2 pellet fuel pins irradiated in a PWR loop

    International Nuclear Information System (INIS)

    Linde, A. van der; Lucas Luijckx, H.J.B.; Verheugen, J.H.N.

    1981-04-01

    Three fuel pin bundles, R-109/1, 2 and 3, were irradiated in a PWR loop in the HFR at Petten during respectively 131, 57 and 57 effective full power days at average powers of approximately 39 kW.m -1 and at peak powers of approximately 60 kW.m -1 . The results of the post-irradiation examinations of these fuel bundles are presented. (Auth.)

  4. Production method of burnable poison incorporated fuel pellet by coating

    International Nuclear Information System (INIS)

    Naito, Naoyoshi.

    1993-01-01

    A cylindrical member is formed with an organic material which is melted, decomposed or evaporated by heating. Such organic materials include polyethylene and polyvinyl alcohol, for example. A predetermined amount of burnable poisons are homogeneously incorporated in the cylindrical member by a means, such as melting before fabricating it into a cylindrical shape. UO 2 fuel pellets are inserted to the cylindrical member and heated, to scatter only the organic materials, so that non-volatile burnable poisons are homogeneously left on the surface of the pellets. It is preferred that the cylindrical member having pellets inserted therein is inserted to a cladding tube and applied with a heat treatment. With such procedures, a UO 2 pellet is coated with burnable poisons by a convenient and compact device. In addition, grinding step after the coating is unnecessary. (I.N.)

  5. Increased fuel burn-up and fuel cycle equilibrium

    International Nuclear Information System (INIS)

    Debes, M.

    2001-01-01

    Improvement of nuclear competitiveness will rely mainly on increased fuel performance, with higher burn-up, and reactors sustained life. Regarding spent fuel management, the EDF current policy relies on UO 2 fuel reprocessing (around 850 MTHM/year at La Hague) and MOX recycling to ensure plutonium flux adequacy (around 100 MTHM/year, with an electricity production equivalent to 30 TWh). This policy enables to reuse fuel material, while maintaining global kWh economy with existing facilities. It goes along with current perspective to increase fuel burn-up up to 57 GWday/t mean in 2010. The following presentation describes the consequences of higher fuel burn-up on fuel cycle and waste management and implementation of a long term and global equilibrium for decades in spent fuel management resulting from this strategy. (author)

  6. Design of the ME-I powder mixer. Report of the design and construction of the powder mixer, according to the safety requisites of the pellets fabrication process

    International Nuclear Information System (INIS)

    Mariano H, E.

    1991-03-01

    In order to fulfill the requirements of preparation of powder of UO 2 ; according to procedure P-M-PP-01, rev. 0, for the process of production of pellets, it was designed and manufactured a powders mixer to incorporate the lubricant one (zinc stearate) to the powder of UO 2 . This equipment allows to mix the powder of UO 2 evenly with the one zinc stearate, without forming considerable quantities of fine of UO 2 , besides a sure control for the operators of the process and an easy access to the mixer to inspect the mixture. (Author)

  7. Effect of metallic iron on the oxidative dissolution of UO2 doped with a radioactive alpha emitter in synthetic Callovian-Oxfordian groundwater

    Science.gov (United States)

    Odorowski, Mélina; Jegou, Christophe; De Windt, Laurent; Broudic, Véronique; Jouan, Gauthier; Peuget, Sylvain; Martin, Christelle

    2017-12-01

    In the hypothesis of direct disposal of spent fuel in a geological nuclear waste repository, interactions between the fuel mainly composed of UO2 and its environment must be understood. The dissolution rate of the UO2 matrix, which depends on the redox conditions on the fuel surface, will have a major impact on the release of radionuclides into the environment. The reducing conditions expected for a geological disposal situation would appear to be favorable as regards the solubility and stability of the UO2 matrix, but may be disturbed on the surface of irradiated fuel. In particular, the local redox conditions will result from a competition between the radiolysis effects of water under alpha irradiation (simultaneously producing oxidizing species like H2O2, hydrogen peroxide, and reducing species like H2, hydrogen) and those of redox active species from the environment. In particular, Fe2+, a strongly reducing aqueous species coming from the corrosion of the iron canister or from the host rock, could influence the dissolution of the fuel matrix. The effect of iron on the oxidative dissolution of UO2 was thus investigated under the conditions of the French disposal site, a Callovian-Oxfordian clay formation chosen by the French National Radioactive Waste Management Agency (Andra), here tested under alpha irradiation. For this study, UO2 fuel pellets doped with a radioactive alpha emitter (238/239Pu) were leached in synthetic Callovian-Oxfordian groundwater (representative of the French waste disposal site groundwater) in the presence of a metallic iron foil to simulate the steel canister. The pellets had varying levels of alpha activity, in order to modulate the concentrations of species produced by water radiolysis on the surface and to simulate the activity of aged spent fuel after 50 and 10,000 years of alpha radioactivity decay. The experimental data showed that whatever the sample alpha radioactivity, the presence of iron inhibits the oxidizing dissolution of

  8. Uranium migration in spark plasma sintered W/UO2 CERMETS

    Science.gov (United States)

    Tucker, Dennis S.; Wu, Yaqiao; Burns, Jatuporn

    2018-03-01

    W/UO2 CERMET samples were sintered in a Spark Plasma Sintering (SPS) furnace at various temperature under vacuum and pressure. High Resolution Transmission Electron Microscopy (HRTEM) with Energy Dispersive Spectroscopy (EDS) was performed on the samples to determine interface structures and uranium diffusion from the UO2 particles into the tungsten matrix. Local Electrode Atom Probe (LEAP) was also performed to determine stoichiometry of the UO2 particles. It was seen that uranium diffused approximately 10-15 nm into the tungsten matrix. This is explained in terms of production of oxygen vacancies and Fick's law of diffusion.

  9. A study on improvement of UO2 powder production process for high sintered density

    International Nuclear Information System (INIS)

    Park, Jin Hoh; Hwang, Sung Tae; Jun, Kwan Sik; Choi, Yoon Dong; Choi, Jong Hyun; Lee, Kyoo Il; Kim, Tae Joon; Jung, Kyung Chae; Kim, Kwang Lak; Kwon, Sang Woon; Kim, Byung Hoh; Hong, Soon Bok

    1995-01-01

    Various conversion processes were reviewed from the viewpoint of manufacturing cost, product quality and liquid waste. The MDD process was selected a suitable target process for the good quality of UO 2 powder and the recycling availability of nitric acid. The MDD process consists of two steps, double salt preparation [(NH 4 ) 2 UO 2 (NO 3 ) 4 ] from uranyl nitrate solution and thermal decomposition/reduction to UO 2 powder. The reaction mechanism and properties for the intermediates were analyzed to define the proposed operational conditions of the process. The conceptual process was proposed and experimental facility was designed and installed. 12 figs, 7 tabs, 7 refs. (Author)

  10. Study by electronic structure calculations of the radiation damage in the UO2 nuclear fuel: behaviour of the point defects and fission gases

    International Nuclear Information System (INIS)

    Vathonne, Emerson

    2014-01-01

    Uranium dioxide (UO 2 ) is worldwide the most widely used fuel in nuclear plants in the world and in particular in pressurized water reactors (PWR). In-pile the fission of uranium nuclei creates fission products and point defects in the fuel. The understanding of the evolution of these radiation damages requires a multi-scale modelling approach of the nuclear fuel, from the scale of the pellet to the atomic scale. We used an electronic structure calculation method based on the density functional theory (DFT) to model radiation damage in UO 2 at the atomic scale. A Hubbard-type Coulomb interaction term is added to the standard DFT formalism to take into account the strong correlations of the 5f electrons in UO 2 . This method is used to study point defects with various charge states and the incorporation and diffusion of krypton in uranium dioxide. This study allowed us to obtain essential data for higher scale models but also to interpret experimental results. In parallel of this study, three ways to improve the state of the art of electronic structure calculations of UO 2 have been explored: the consideration of the spin-orbit coupling neglected in current point defect calculations, the application of functionals allowing one to take into account the non-local interactions such as van der Waals interactions important for rare gases and the use of the Dynamical Mean Field Theory combined to the DFT method in order to take into account the dynamical effects in the 5f electron correlations. (author) [fr

  11. Corrosion studies with high burnup light water reactor fuel. Release of nuclides into simulated groundwater during accumulated contact time of up to two years

    International Nuclear Information System (INIS)

    Zwicky, Hans-Urs; Low, Jeanett; Ekeroth, Ella

    2011-03-01

    In the framework of comprehensive research work supporting the development of a Swedish concept for the disposal of highly radioactive waste and spent fuel, Studsvik has performed a significant number of spent fuel corrosion studies under a variety of different conditions. These experiments, performed between 1990 and 2002, covered a burnup range from 27 to 49 MWd/kgU, which was typical for fuel to be disposed at that time. As part of this work, the so called Series 11 tests were performed under oxidising conditions in synthetic groundwater with fuel samples from a rod irradiated in the Ringhals 1 Boiling Water Reactor (BWR). In the meantime, Swedish utilities tend to increase the discharge burnup of fuel operated in their reactors. This means that knowledge of spent fuel corrosion performance has to be extended to higher burnup as well. Therefore, a series of experiments has been started at Studsvik, aiming at extending the data base acquired in the Series 11 corrosion tests to higher burnup fuel. Fuel burnup leads to complex and significant changes in the composition and properties of the fuel. The transformed microstructure, which is referred to as the high burnup structure or rim structure in the outer region of the fuel, consists of small grains of submicron size and a high concentration of pores of typical diameter 1 to 2 μm. This structure forms in UO 2 fuel at a local burnup above 50 MWd/kgU, as long as the temperature is below 1,000-1,100 deg C. The high burnup at the pellet periphery is the consequence of plutonium build-up by neutron capture in 238 U followed by fission of the formed plutonium. The amount of fission products in the fuel increases more or less linearly with burnup, in contrast to alpha emitting actinides that increase above average. As burnup across a spent fuel pellet is not uniform, but increases towards the periphery, the radiation field is also larger at the pellet surface. At the same time, it is easier for water to access the

  12. Molybdenum-UO2 cermet irradiation at 1145 K.

    Science.gov (United States)

    Mcdonald, G.

    1971-01-01

    Two molybdenum-uranium dioxide cermet fuel pins with molybdenum clad were fission-heated in a forced-convection helium coolant for sufficient time to achieve 5.3% burnup. The cermet core contained 20 wt % of 93.2% enriched uranium dioxide. The results were as follows: there was no visible change in the appearance of the molybdenum clad during irradiation; the maximum increase in diameter of the fuel pins was 0.8%; there was no migration of uranium dioxide along grain boundaries and no evident interaction between molybdenum and uranium dioxide; and, finally, approximately 12% of the fission gas formed was released from the cermet core into the gas plenum.

  13. Measurements of density and of thermal expansion coefficient of sodium tetraborate (borax)-UO2 and of sodium metaborate-UO2 solutions

    International Nuclear Information System (INIS)

    Dalle Donne, M.; Dorner, S.

    1980-12-01

    Measurements have been performed of the density and volumetric thermal expansion coefficient of liquid sodium tetraborate (borax) and of sodium metaborate both pure and with two different amounts of UO 2 dissolved in each. These data are required for the design of core-catchers based on sodium borates. The measurements have been performed with the buoyancy method in the temperature range from 850 0 C to 1325 0 C. The data for the pure borax and for the sodium metaborate agree reasonably well with the data from the literature, giving confidence that the measurements are correct and the new data for the salts with UO 2 are reliable. (orig.) [de

  14. 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

  15. Development of automation and remotisation systems for fabrication of (Th-233U)O2 MOX fuel for AHWR

    International Nuclear Information System (INIS)

    Saraswat, Anupam; Danny, K.M.; Chakraborty, S.; Somayajulu, P.S.; Kumar, Arun; Mittal, R.; Prasad, R.S.; Mahule, K.N.; Panda, S.; Jayarajan, K.

    2011-01-01

    To meet the ever increasing power requirement of India, country is planning to utilize its large thorium reserves for the third stage of nuclear power program based on Thorium-Uranium 233 fuel in A.H.W.R. Although there are many advantages of (Th- 233 U)O 2 fuel cycle, presence of radiological hazards due to the presence of 1000-2000 ppm level of 232 U in the 233 U fuel and inertness of ThO 2 makes handling and fabrication of fuel difficult. The associated high alpha and gamma activity demands high level of automation and remote handling in alpha tight hot cells. To demonstrate automation and remotisation in (Th- 233 U)O 2 fuel fabrication, a mock up facility is being set up at BARC. This facility shall develop automation systems required for remote fuel fabrication in a simulated hot cell environment. There are many innovative schemes and systems being developed like integrated powder pellet system, remote viewing system for hot cell application etc. Low visibility inside the hot cell has always been a problem for the operator. To overcome this problem a remote viewing system has been developed by which entire hot cell area can be scanned with the use of a joystick and the display can be seen on a LCD monitor. The viewing system is made up of radiation resistant optics which can work even in high gamma fields. It consists of objective end assembly which is used to scan the hot cell area with the help of prism doublets and drive mechanism for capturing full 360 deg solid angle view. There is a Galilean telescope and focusing system used for focusing images of distant objects. Drive mechanism can be controlled by the joystick available to the operator. System has a high resolution CCD display and camera which gives a clear display of objects lying inside the hot cell area. Integrated powder pellet system is being developed for fabrication of MOX pellets from feed powder. This will be automated system which will take input in the form of MOX powder and convert it

  16. 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.

  17. The influence of moisture on air oxidation of UO2: Calculations and observations

    International Nuclear Information System (INIS)

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

    1993-01-01

    Phase relationships among solids in the UO 2 -O 2 -H 2 O system at 25, 100, and 200C and pressures to 2 MPa have been calculated from critically evaluated thermodynamic data. Stability limits of the solids are expressed in terms of oxygen and water partial pressures at each temperature. The results are then discussed in terms of known UO 2 oxidation reactions and uranium mineralogy. Particular attention is paid to UO 3 hydrates, some of which are shown to be stable phases in air at very low relative humidities (down to ∼0.1% at 25C). This is relevant to fuel storage because of the very high molar volumes of these phases, relative to UO 2 , and consequent potential for damage to defected fuel assemblies. Comparison of the calculated phase relationships with observed UO 2 oxidation behavior helps to identify those phase interconversions that are kinetically constrained

  18. Etching of UO2 in NF3 RF Plasma Glow Discharge

    International Nuclear Information System (INIS)

    John M. Veilleux

    1999-01-01

    A series of room temperature, low pressure (10.8 to 40 Pa), low power (25 to 210 W) RF plasma glow discharge experiments with UO 2 were conducted to demonstrate that plasma treatment is a viable method for decontaminating UO 2 from stainless steel substrates. Experiments were conducted using NF 3 gas to decontaminate depleted uranium dioxide from stainless-steel substrates. Depleted UO 2 samples each containing 129.4 Bq were prepared from 100 microliter solutions of uranyl nitrate hexahydrate solution. The amorphous UO 2 in the samples had a relatively low density of 4.8 gm/cm 3 . Counting of the depleted UO 2 on the substrate following plasma immersion was performed using liquid scintillation counting with alpha/beta discrimination due to the presence of confounding beta emitting daughter products, 234 Th and 234 Pa. The alpha emission peak from each sample was integrated using a gaussian and first order polynomial fit to improve quantification. The uncertainties in the experimental measurement of the etched material were estimated at about ± 2%. Results demonstrated that UO 2 can be completely removed from stainless-steel substrates after several minutes processing at under 200 W. At 180 W and 32.7 Pa gas pressure, over 99% of all UO 2 in the samples was removed in just 17 minutes. The initial etch rate in the experiments ranged from 0.2 to 7.4 microm/min. Etching increased with the plasma absorbed power and feed gas pressure in the range of 10.8 to 40 Pa. A different pressure effect on UO 2 etching was also noted below 50 W in which etching increased up to a maximum pressure, approximately23 Pa, then decreased with further increases in pressure

  19. 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)

  20. 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

  1. Simulation of the irradiation-induced micro-thermo-mechanical behaviors evolution in ADS nuclear fuel pellets

    Science.gov (United States)

    Ding, Shurong; Zhao, Yunmei; Wan, Jibo; Gong, Xin; Wang, Canglong; Yang, Lei; Huo, Yongzhong

    2013-11-01

    An Accelerator Driven System (ADS) is dedicated to Minor Actinides (MA) transmutation. The fuels for ADS are highly innovative, which are composite fuel pellets with the fuel particles containing MA phases dispersed in a MgO or Mo matrix. Assuming that the fuel particles are distributed periodically in the MgO matrix, a three-dimensional finite element model is developed. The three-dimensional incremental large-deformation constitutive relations for the fuel particles and matrix are separately built, and a method is accordingly constructed to implement simulation of the micro-thermo-mechanical behaviors evolution. Evolutions of the temperature and mechanical fields are given and discussed. With irradiation creep included in the MgO matrix constitutive relation, the conclusions can be drawn as that (1) irradiation creep has a remarkable effect on the mechanical behaviors evolution in the matrix; (2) irradiation creep plays an important role in the damage mechanism interpretation of ceramic matrix fuel pellets. Thermal conductivity The thermal conductivity model is adopted as KUO2 = K0·FD·FP·FM·FR, which was proposed by Lucuta et al. [10] to adapt to the high burnup conditions with consideration of the effects of temperature, burnup, porosity and fission products. K0 is the thermal conductivity of fully dense un-irradiated UO2, as Eq. (1) in W/m K; FD, FP are the adjust factors reflecting the effects of dissolved and precipitated fission products; FM and FR are factors due to porosity and irradiation effects. The adopted thermal conductivity varies with temperature and burnup, which expresses its degradation with burnup, with the terms as k0={1}/{0.0375+2.165×10-4T}+{4.715×109}/{T2}exp-{16361}/{T} FD={1.09}/{B3.265}+{0.0643}/{√{B}}√{T}artan{1}/{1.09/B3.265}+{0.0643}/{√{B}}√{T} FP=1+0.019B/3-0.019B{1}/{1+exp(1200-T100)} FM={1-P}/{1+(s-1)P} FR=1-{0.2}/{1+expT-90080} Thermal expansion The engineering strain of thermal expansion [11] is given as {ΔL}/{L0

  2. 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)

  3. Analysis of gas flow measurements from the IFA-633 UO2/MOX comparison test

    International Nuclear Information System (INIS)

    Rossiter, Glyn

    2005-01-01

    The release rate to birth rate ratio (R/B) results from the gas flow measurements performed during the joint programme irradiation of the IFA-633 UO 2 /MOX comparison test have been analysed using both classical and fractal methodologies. Possible calculational procedures for precursor enhancement factors and rod average diffusion coefficients were considered and suitable procedures were then implemented. The surface area to volume ratio (S/V) and recoil R/B values generated using the two methodologies have been compared to each other and to results obtained for other Halden Project gas flow rigs (IFAs 504, 558, 563, 569 and 655). The merits of the methodologies have then been discussed. It was found that the trends in the classical recoil R/B and in the fractal S/V for the shortest lived isotopes were in better agreement with the expected S/V behaviour than the trends in the classical S/V and in the fractal S/V for the longer lived isotopes. The beginning of life (BOL) S/V versus temperature behaviour for both IFA-633 and IFA-655 has been investigated and has been found to be more consistent with expectation when the fractal methodology is used. The peak fuel temperature versus rod average burnup behaviour of the IFA-633 fuel rods has been examined in order to investigate whether there is any correlation between the S/V results and the extent of periods during which the Halden (or Vitanza) threshold for significant fission gas release was exceeded. The behaviour was more consistent with the trends in the classical recoil R/B and fractal S/V for the shortest lived isotopes than with the trends in the classical S/V and the fractal S/V for the longer lived isotopes. The analysis of the through-life and BOL S/V and recoil R/B results generated using the classical and fractal methodologies has shown that the behaviour of the classical recoil R/B is difficult to explain. This is evidence that the classical recoil R/B results contain a diffusional release component

  4. 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

  5. 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)

  6. Development and verification of Monte Carlo burnup calculation system

    International Nuclear Information System (INIS)

    Ando, Yoshihira; Yoshioka, Kenichi; Mitsuhashi, Ishi; Sakurada, Koichi; Sakurai, Shungo

    2003-01-01

    Monte Carlo burnup calculation code system has been developed to evaluate accurate various quantities required in the backend field. From the Actinide Research in a Nuclear Element (ARIANE) program, by using, the measured nuclide compositions of fuel rods in the fuel assemblies irradiated in the commercial Netherlands BWR, the analyses have been performed for the code system verification. The code system developed in this paper has been verified through analysis for MOX and UO2 fuel rods. This system enables to reduce large margin assumed in the present criticality analysis for LWR spent fuels. (J.P.N.)

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

    International Nuclear Information System (INIS)

    Forsyth, R.

    1987-03-01

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

  8. Leaching patterns and secondary phase formation during unsaturated leaching of UO2 at 90 degrees C

    International Nuclear Information System (INIS)

    Wronkiewicz, D.J.; Bates, J.K.; Gerding, T.J.; Veleckis, E.; Tani, B.S.

    1991-11-01

    Experiments are being conducted that examine the reaction of UO 2 with dripping oxygenated ground water at 90 degrees C. The experiments are designed to identify secondary phases formed during UO 2 alteration, evaluate parameters controlling U release, and act as scoping tests for studies with spent fuel. This study is the first of its kind that examines the alteration of UO 2 under unsaturated conditions expected to exist at the proposed Yucca Mountain repository site. Results suggest the UO 2 matrix will readily react within a few months after being exposed to simulated Yucca Mountain conditions. A pulse of rapid U release, combined with the formation of dehydrated schoepite on the UO 2 surface, characterizes the reaction between one to two years. Rapid dissolution of intergrain boundaries and spallation of UO 2 granules appears to be responsible for much of the U released. Differential release of the UO 2 granules may be responsible for much of the variation observed between duplicate experiments. Less than 5 wt % of the released U remains in solution or in a suspended form, while the remaining settles out of solution as fine particles or is reprecipitated as secondary phases. Subsequent to the pulse period, U release rates decline and a more stable assemblage of uranyl silicate phases are formed by incorporating cations from the ground water leachant. Uranophane, boltwoodite, and sklodowskite appear as the final solubility limiting phases that form in these tests. This observed paragenetic sequence (from uraninite to schoepite-type phases to uranyl silicates) is identical to those observed in weathered zones of natural uraninite occurrences. The combined results indicate that the release of radionuclides from spent fuel may not be limited by U solubility constraints, but that spallation of particulate matter may be an important, if not the dominant release mechanism affecting release

  9. 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)

  10. Study on the retention of enriched UO2F2 in the mouse and its radiogenotoxicological effects

    International Nuclear Information System (INIS)

    Hu Qiyue; Zhu Shoupeng

    1991-06-01

    The study on toxicological effects of enriched UO 2 F 2 was undertaken in purebred BALB/c male mice to examine: (a) the retention in body; (b) the testicular clearance; (c) the effect of sperm abnormality; (d) the effect of chromosomal aberrations in spermatogonia and primary spermatocytes; and (e) the effect of DNA damage in germ cells in various spermiogenic stages. Results show that enriched UO 2 F 2 mainly deposited in the kidneys, then the skeleton and liver. The amount of enriched UO 2 F 2 depositing in other tissues was small. Enriched UO 2 F 2 was similar to the natural uranium in transference and retention in the body. The testis had efficient clearance of enriched UO 2 F 2 . Enriched UO 2 F 2 could result in sperm abnormality. Even with the same treating does but at different treating time the rates of sperm abnormality were different. Enriched UO 2 F 2 could result in chromosomal aberrations in spermatogonia and primary spermatocytes. The important type of aberrations in spermatogonia was break. For primary spermatocytes the most significant aberration was multivalents. Enriched UO 2 F 2 could also result in DNA breakage in germ cells. The sensitivity of mouse germ cells at various stages to enriched UO 2 F 2 was different. There was a linear relationship between the amount of sperm DNA eluted and enriched UO 2 F 2 dose

  11. Method of manufacturing nuclear fuel pellet

    International Nuclear Information System (INIS)

    Oguma, Masaomi; Masuda, Hiroshi; Hirai, Mutsumi; Tanabe, Isami; Yuda, Ryoichi.

    1989-01-01

    In a method of manufacturing nuclear fuel pellets by compression molding an oxide powder of nuclear fuel material followed by sintering, a metal nuclear material is mixed with an oxide powder of the nuclear fuel material. As the metal nuclear fuel material, whisker or wire-like fine wire or granules of metal uranium can be used effectively. As a result, a fuel pellet in which the metal nuclear fuel is disposed in a network-like manner can be obtained. The pellet shows a great effect of preventing thermal stress destruction of pellets upon increase of fuel rod power as compared with conventional pellets. Further, the metal nuclear fuel material acts as an oxygen getter to suppress the increase of O/M ratio of the pellets. Further, it is possible to reduce the swelling of pellet at high burn-up degree. (T.M.)

  12. Choosing the optimum burnup

    International Nuclear Information System (INIS)

    Geller, L.; Goldstein, L.; Franks, W.A.

    1986-01-01

    This paper reviews some of the considerations utilities must evaluate when going to higher discharge burnups. The advantages and disadvantages of higher discharge burnups are described, as well as a consistent approach for evaluating optimum discharge burnup and its comparison to current practice. When an analysis is performed over the life of the plant, the design of the terminal cycles has significant impact on the lifetime savings from higher burnups. Designs for high burnup cycles have a greater average inventory value in the core. As one goes to higher burnup, there is a greater likelihood of discarding a larger value in unused fuel unless the terminal cycles are designed carefully. This effect can be large enough in some cases to wipe out the lifetime cost savings relative to operating with a higher discharge burnup cycle

  13. Near Surface Stoichiometry in UO2: A Density Functional Theory Study

    Directory of Open Access Journals (Sweden)

    Jianguo Yu

    2015-01-01

    Full Text Available The mechanisms of oxygen stoichiometry variation in UO2 at different temperature and oxygen partial pressure are important for understanding the dynamics of microstructure in these crystals. However, very limited experimental studies have been performed to understand the atomic structure of UO2 near surface and defect effects of near surface on stoichiometry in which the system can exchange atoms with the external reservoir. In this study, the near (110 surface relaxation and stoichiometry in UO2 have been studied with density functional theory (DFT calculations. On the basis of the point-defect model (PDM, a general expression for the near surface stoichiometric variation is derived by using DFT total-energy calculations and atomistic thermodynamics, in an attempt to pin down the mechanisms of oxygen exchange between the gas environment and defected UO2. By using the derived expression, it is observed that, under poor oxygen conditions, the stoichiometry of near surface is switched from hyperstoichiometric at 300 K with a depth around 3 nm to near-stoichiometric at 1000 K and hypostoichiometric at 2000 K. Furthermore, at very poor oxygen concentrations and high temperatures, our results also suggest that the bulk of the UO2 prefers to be hypostoichiometric, although the surface is near-stoichiometric.

  14. Identification of secondary phases formed during unsaturated reaction of UO2 with EJ-13 water

    International Nuclear Information System (INIS)

    Bates, J.K.; Tani, B.S.; Veleckis, E.

    1989-01-01

    A set of experiments, wherein UO 2 has been contacted by dripping water, has been conducted over a period of 182.5 weeks. The experiments are being conducted to develop procedures to study spent fuel reaction under unsaturated conditions that are expected to exist over the lifetime of the proposed Yucca Mountain repository site. One half of the experiments have been terminated, while one half are ongoing. Analyses of solutions that have dripped from the reacted UO 2 have been performed for all experiments, while the reacted UO 2 surfaces have been examined for the terminated experiments. A pulse of uranium release from the UO 2 solid, combined with the formation of schoepite on the surface of the UO 2 , was observed between 39 and 96 weeks of reaction. Thereafter, the uranium release decreased and a second set of secondary phases was observed. The latter phases incorporated cations from the EJ-13 water and included boltwoodite, uranophane, sklodowskite, compreignacite, and schoepite. The experiments are continuing to monitor whether additional changes in solution chemistry or secondary phase formation occurs. 6 refs., 2 figs., 2 tabs

  15. Fuel pellet fracture and relocation

    International Nuclear Information System (INIS)

    Walton, L.A.; Husser, D.L.

    1983-01-01

    The model used to describe fuel pellet fracture and relocation is an important feature of a fuel performance computer code. This model becomes especially important if the computer code is principally to be used for the evaluation of pellet clad interaction. The fracture and relocation model being developed for the B and W fuel performance code FUMAC was derived from an extensive data base. Cross sections of irradiated fuel rods were photographically magnified and measured to determine the configuration of the fragments of the fractured fuel pellets. Data, representing a wide range of LWR fuel designs and as-manufactured mechanical configurations, were catalogued and systematically reduced and then correlated as a function of the likely independent variables. These correlations define the key phenomenological behavior patterns which the relocation model must duplicate and indicate which mechanistic approaches are viable explanations of this behavior. The data base covers the burnup range from approximately one to 35 GWd/mtU and linear heat rates from less than 100 to nearly 700 W/Cm. This paper presents the correlated data base and the methods used to derive and interpret it. It was determined from this data base that pellet cracking is initially both power level and burnup dependent but tends to saturate eventually with continued steady irradiation. Fuel pellet relocation was found to be much more extensive than would be deduced from thermal considerations alone. Even at very low burnups fuel fragments were found to move outward until restrained by the cladding. The results also suggest that changes in internal resistance to heat flow within the pellets due to the opening of cracks may be as important as peripheral gap changes to the thermal modeler. The transient response and thermal implications of this model are recommended as primary areas for future investigation

  16. Thermal reactions of uranium metal, UO 2, U 3O 8, UF 4, and UO 2F 2 with NF 3 to produce UF 6

    Science.gov (United States)

    McNamara, Bruce; Scheele, Randall; Kozelisky, Anne; Edwards, Matthew

    2009-11-01

    This paper demonstrates that NF 3 fluorinates uranium metal, UO 2, UF 4, UO 3, U 3O 8, and UO 2F 2·2H 2O to produce the volatile UF 6 at temperatures between 100 and 550 °C. Thermogravimetric and differential thermal analysis reaction profiles are described that reflect changes in the uranium fluorination/oxidation state, physiochemical effects, and instances of discrete chemical speciation. Large differences in the onset temperatures for each system investigated implicate changes in mode of the NF 3 gas-solid surface interaction. These studies also demonstrate that NF 3 is a potential replacement fluorinating agent in the existing nuclear fuel cycle and in actinide volatility reprocessing.

  17. Density, thermal expansion coefficient and viscosity of sodium tetraborate (borax)-UO2 and of sodium metaborate-UO2 solutions at high temperatures

    International Nuclear Information System (INIS)

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

    1983-01-01

    Measurements have been performed of the density, of the volumetric thermal expansion coefficient and of the viscosity of liquid sodium tetraborate (borax) and of sodium metaborate both pure and with two different amounts of UO 2 dissolved in each. The viscosity measurements have been performed for the solution of sodium tetraborate with UO 2 and CeO 2 , and with CeO 2 only as well. These data are required for the design of core-catchers based on sodium borates. The density measurements have been performed with the buoyancy method in the temperature range from 825 0 C to 1300 0 C, the viscosity measurements in the temperature range 700-1250 0 C with a modified Haake viscosity balance. The balance was previously calibrated at ambient temperature with a standard calibration liquid and at high temperatures, with data for pure borax available from the literature. (orig.)

  18. On the correlation between fuel structure and mechanical properties of UO2

    International Nuclear Information System (INIS)

    Blank, H.; Mandler, R.; Matzke, H.; Routbort, J.; Werner, P.

    1983-01-01

    The relation between the structure of a UO 2 fuel and its mechanical properties are discussed and illustrated for particular types of UO 2 by measurements of fracture surface energy, hardness, fracture stress and compressive deformation at 1870 and 1970 K. This gives the background for treating the question whether it is possible to find a simple experimental method for correlating the mechanical properties of UO 2 before irradiation with those after various irradiation histories. Hardness measurements might be such a method if combined with a detailed structural analysis and sufficient knowledge about the irradiation history. However, for a meaningful interpretation of the data the presently available 'classical' methods of fracture mechanics are inadequate and, furthermore, sufficient additional (not yet available) information on the relations between mechanical properties and structural details are required. (author)

  19. 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.)

  20. 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

  1. Reducing the stoichiometric excess of HF in the hydrofluorination of UO2

    International Nuclear Information System (INIS)

    Zhao Jun; Qiu Lufu; Zhong Xing; Xu Heqing

    1989-11-01

    In a fluidized bed, UO 2 obtained from the decomposition-reduction of AUC (Ammonium Uranyl Carbonate) was fed to absorb HF remaining in the exhaust gas of UF 4 production process. In the case of 60% conversion of UO 2 and the reaction temperature in the region of 300 deg C, HF remaining in the exhaust gas in absorbing fluidized bed was less than 7 ∼ 8% (w/w), i.e. apparent stoichiometric excess of HF had reduced to 0% more or less. Hence, with the high hydrofluorination reactivity of UO 2 obtained from the decomposition-reduction of AUC, it is possible to reduce evidently the stoichiometric excess of HF in the hydrofluorination process by two fluidized beds in series in which solids move against the gas flow

  2. Vapor deposition of large area NpO2 and UO2 deposits

    International Nuclear Information System (INIS)

    Adair, H.L.; Gibson, J.R.; Kobisk, E.H.; Dailey, J.M.

    1976-01-01

    Deposition of NpO 2 and UO 2 thin films over an area of 7.5 to 10 cm diam has become a routine operation in preparation of fission chamber plates. Vacuum evaporation or electroplating has been used for this purpose. The ''paint brush'' technique has been used as well; however, uniformity requirements normally eliminate this procedure. Vapor deposition in vacuum appears to be the most suitable technique for preparing NpO 2 and UO 2 deposits of >200 cm 2 . This paper describes the procedures used in preparing uniform large area deposits of NpO 2 (approximately 300 cm 2 ) and UO 2 (approximately 2000 cm 2 ) by vacuum evaporation using electron bombardment heating and several substrate motion and heating methods to achieve uniformity and adhesion

  3. Synthesis and investigation of uranyl molybdate UO2MoO4

    International Nuclear Information System (INIS)

    Nagai, Takayuki; Sato, Nobuaki; Kitawaki, Shin-ichi; Uehara, Akihiro; Fujii, Toshiyuki; Yamana, Hajimu; Myochin, Munetaka

    2013-01-01

    In order to examine easily synthetic conditions of uranyl molybdate, UO 2 MoO 4 , used for the reprocessing process study of spent nuclear oxide fuels in alkaline molybdate melts, the uranium molybdate compounds were produced from U 3 O 8 powder and anhydrous MoO 3 reagent. The results of having investigated them in solid state by using X-ray diffractometry and Raman spectrometry, it was confirmed that UO 2 MoO 4 could be synthesized by heating mixed powder of U 3 O 8 and MoO 3 with stoichiometric mole ratio at 770 °C for 4 h under air atmosphere. Moreover, adding this UO 2 MoO 4 into Li 2 MoO 4 -Na 2 MoO 4 eutectic melt, most of the dissolved uranium species in the melt were observed as hexa–valent uranyl ions by absorption spectrophotometry

  4. Thermal properties of UO2 from density functional theory: role of strong correlations

    International Nuclear Information System (INIS)

    Panigrahi, Puspamitra; Kaur Gurpreet; Valsakumar, M.C.

    2011-01-01

    We report a study of ground state magnetic structure of Uranium-dioxide (UO 2 ) using ab initio calculations employing PAW pseudopotentials and Dudarev's version of GGA+U formalism as implemented in VASP to take into account the strong on-site Coulomb correlation among the localized Uranium-5f electrons. By choosing the value of the Hubbard parameter U eff to be 4.0 eV, we have confirmed the experimental observation that the ground state of UO 2 is an insulator with an anti-ferromagnetic (AFM) ordering. We study systematically the ground state structural, electronic, and magnetic properties of UO 2 and focus on the structure sensitive thermal properties such as specific heat, thermal expansion and comment on the calculation of thermal conductivity. (author)

  5. Technological aspects concerning the production procedures of UO2-Gd2O3 nuclear fuel

    International Nuclear Information System (INIS)

    Durazzo, Michelangelo; Riella, Humberto Gracher

    2007-01-01

    The direct incorporation of Gd 2 O 3 powder into UO 2 powder by dry mechanical blending is the most attractive process for producing UO 2 -Gd 2 O 3 nuclear fuel. However, previous experimental results by our group indicated that pore formation due to the Kirkendall effect delays densification and, consequently, diminishes the final density of this type of nuclear fuel. Considering this mechanism as responsible for the poor sintering behavior of UO 2 -Gd 2 O 3 fuel prepared by the mechanical blending method, it was possible to propose, discuss and, in certain cases, preliminarily test feasible adjustments in fabrication procedures that would minimize, or even totally compensate, the negative effects of pore formation due to the Kirkendall effect. This work presents these considerations. (author)

  6. Modeling of UO2 aqueous dissolution over a wide range of conditions

    International Nuclear Information System (INIS)

    Steward, S.A.; Weed, H.C.

    1993-11-01

    Previously it was not possible to predict reliably the rate at which spent fuel would react with groundwater because of conflicting data in the literature. The dissolution of the UO 2 spent fuel matrix is a necessary step for aqueous release of radioactive fission products. Statistical experimental design was used to plan a set of UO 2 dissolution experiments to examine systematically the effects of temperature (25--75C), dissolved oxygen (0.002--0.2 atm overpressure), pH (8--10) and carbonate (2-200x10 -4 molar) concentrations on UO 2 dissolution. The average uranium dissolution rate was 4.3 mg/m 2 /day. The regression fit of the data indicate an Arrhenius type activation energy of 8750 cal/mol and a half-power dependence on dissolved oxygen in the simulated groundwater

  7. Determination of the UO2-ZrO2-BaO equilibrium diagram

    International Nuclear Information System (INIS)

    Paschoal, J.O.A.; Kleykanp, H.; Thuemmler, F.

    1984-01-01

    It is determined the equilibrium diagram of UO 2 - ZrO 2 - BaO to interpret and predict changes in the chemical properties of ceramic (oxide) nuclear fuels during irradiation. The isothermal section of the system at 1700 0 C was determined experimentally, utilizing the techniques of ceramography, X-ray diffraction analysis, microprobe analysis and differential thermal analysis. The solid solubility limits at 1700 0 C between UO 2 and ZrO 2 , UO 2 and BaO, ZrO 2 and BaO, ZrO 2 and BaO and BaUO 3 and BaZrO 3 is presented. The influence of oxygen potential in relation to the different phases is discussed and the phase diagram of the system presented. (M.C.K.) [pt

  8. Radiolytic modelling of spent fuel oxidative dissolution mechanism. Calibration against UO2 dynamic leaching experiments

    International Nuclear Information System (INIS)

    Merino, J.; Cera, E.; Bruno, J.; Quinones, J.; Casas, I.; Clarens, F.; Gimenez, J.; Pablo, J. de; Rovira, M.; Martinez-Esparza, A.

    2005-01-01

    Calibration and testing are inherent aspects of any modelling exercise and consequently they are key issues in developing a model for the oxidative dissolution of spent fuel. In the present work we present the outcome of the calibration process for the kinetic constants of a UO 2 oxidative dissolution mechanism developed for using in a radiolytic model. Experimental data obtained in dynamic leaching experiments of unirradiated UO 2 has been used for this purpose. The iterative calibration process has provided some insight into the detailed mechanism taking place in the alteration of UO 2 , particularly the role of · OH radicals and their interaction with the carbonate system. The results show that, although more simulations are needed for testing in different experimental systems, the calibrated oxidative dissolution mechanism could be included in radiolytic models to gain confidence in the prediction of the long-term alteration rate of the spent fuel under repository conditions

  9. Fish pelleting

    African Journals Online (AJOL)

    PUBLICATIONS1

    fish meal pelletizing machine utilized 4kg of ingredients to produce 3.77kg pellets at an effi- ciency of .... Design and fabrication of fish meal pellet processing machine ... 53 ... horsepower for effective torque application on .... two edges were tacked with a spot weld to hold ... then welded on to the shaft making sure that the.

  10. Irradiation behaviour of UO2/Mo porous cermets for thermionic converters

    International Nuclear Information System (INIS)

    Stora, J.P.; Kauffmann, Y.

    1975-01-01

    Two types of UO 2 Mo porous cernets have been fabricated and irradiated in a Cythere irradiation device. The first cermet is constituted by little bits of dense fuel in which the two constituants are finely dispersed. The whole open porosity is located between the granules. This type of cermet is called breche (33.4vol%UO 2 , 51vol%Mo, 14.8vol%porosity). At the end of the irradiation the burn up was 19000MWd/t(U) and neither swelling of the cermet nor deformation of the can were noted. On the contrary, a shrinkage of the emitter was observed attributed to a fuel densification under irradiation. The second type of cermet is called macrogranule (36vol%UO 2 , 49vol%Mo 15vol%porosity). UO 2 granules of 0.07cm mean diameter are dispersed in the molybdenum matrix. The porosity is regularly distributed all around the UO 2 kernels. The post irradiation metrology shows that the emitter is fairly stable. Only a slight ovalisation of about 0.5% was noted, but the granules of UO 2 were redistributed inside the molybdenum matrix, overlapping the metallic cavity by a condensation-evaporation process. The matrix has crept into the central void and consequently the volume has grown and the whole porosity has increased from about 15% to about 23%. This creeping is due to the fission gas pressure in the molybdenum cavities after 3000 hours of irradiation. In conclusion two types of cermets have shown good behaviour under irradiation and should allow lifetimes of several thousand hours of operation for thermionic fuel elements [fr

  11. Thermal diffusivity measurements between 0 0C and 2000 0C: application to UO2

    International Nuclear Information System (INIS)

    Van Craeynest, J.C.; Weilbacher, J.C.; Lallement, R.

    1969-01-01

    We have built two types of apparatus to measure the thermal diffusivity of ceramic fuels. The first apparatus, based on Angstrom's method, operates between 0 deg. C and 1000 deg. C. Satisfactory results have been obtained for iron, nickel and molybdenum. The other apparatus, based on Cowan's method, operates between 1000 deg. C and 2000 deg. C on thin slabs. The thermal conductivity of UO 2 has been measured from 0 deg. C to 2000 deg. C. There is a good agreement between our results and the well known values for UO 2 . (authors) [fr

  12. Surface modelling on heavy atom crystalline compounds: HfO2 and UO2 fluorite structures

    International Nuclear Information System (INIS)

    Evarestov, Robert; Bandura, Andrei; Blokhin, Eugeny

    2009-01-01

    The study of the bulk and surface properties of cubic (fluorite structure) HfO 2 and UO 2 was performed using the hybrid Hartree-Fock density functional theory linear combination of atomic orbitals simulations via the CRYSTAL06 computer code. The Stuttgart small-core pseudopotentials and corresponding basis sets were used for the core-valence interactions. The influence of relativistic effects on the structure and properties of the systems was studied. It was found that surface properties of Mott-Hubbard dielectric UO 2 differ from those found for other metal oxides with the closed-shell configuration of d-electrons

  13. High density, uniformly distributed W/UO2 for use in Nuclear Thermal Propulsion

    Science.gov (United States)

    Tucker, Dennis S.; Barnes, Marvin W.; Hone, Lance; Cook, Steven

    2017-04-01

    An inexpensive, quick method has been developed to obtain uniform distributions of UO2 particles in a tungsten matrix utilizing 0.5 wt percent low density polyethylene. Powders were sintered in a Spark Plasma Sintering (SPS) furnace at 1600 °C, 1700 °C, 1750 °C, 1800 °C and 1850 °C using a modified sintering profile. This resulted in a uniform distribution of UO2 particles in a tungsten matrix with high densities, reaching 99.46% of theoretical for the sample sintered at 1850 °C. The powder process is described and the results of this study are given below.

  14. 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)

  15. A study on entrapment: splashing of liquid UO2 over small sodium volumes

    International Nuclear Information System (INIS)

    Schins, H.; Klein, K.; Jorzik, E.

    1978-01-01

    Three experiments were done in which each time more than 1 kg of liqid UO 2 was splashed over a plate in which different stainless steel cups full of sodium were contained. No whatsoever indication of entrapment of sodium could be found. The pressure tests indicate up to 1,8 atm and they relax in some second time. These pressure diagrams have been satisfactorily analysed as indicating a sodium vapor formation. In the third experiment e.g. the quenching of 800 g UO 2 in 720 g of sodium in a closed volume of argon of 225 l will provide the requested maximum pressure

  16. Homogenization in powder compacts of UO2-PuO2

    International Nuclear Information System (INIS)

    Verma, R.

    1979-01-01

    The homogenization kinetics in mixed UO 2 -PuO 2 compacts have been studied by adopting a concentric core-shell model of diffusion. An equation relating the extent of homogenization expressed in terms of the fraction of UO 2 remaining undissolved and the time of annealing has been derived. From the equation, the periods required at different annealing temperatures to attain a specified level of homogenization have been calculated. These calculated homogenization times have been found to be in fair agreement with the experimentally observed homogenization times. The derived relationship has also been shown to satisfactorily predict homogenization in Cu-Ni powder compacts. (Auth.)

  17. Modifier free supercritical fluid extraction of uranium from sintered UO2, soil and ore samples

    International Nuclear Information System (INIS)

    Kanekar, A.S.; Pathak, P.N.; Acharya, R.; Mohapatra, P.K.; Manchanda, V.K.

    2011-01-01

    Direct extraction of uranium from different samples viz. sintered UO 2 , soil and ores was carried out by modifier free supercritical fluid using tri-n-butyl phosphate-nitric acid (TBP-HNO 3 ) adduct as extractant. These studies showed that pre-equilibration with more concentrated nitric acid helps in better dissolution and extraction of uranium from sintered UO 2 samples. Modifier free supercritical fluid extraction appears attractive with respect to minimization of secondary wastes. This method resulted 80-100% extraction of uranium from different soil/ore samples. The results were confirmed by performing neutron activation analysis of original (before extraction) and residue (after extraction) samples. (author)

  18. TOPICAL REPORT ON ACTINIDE-ONLY BURNUP CREDIT FOR PWR SPENT NUCLEAR FUEL PACKAGES

    International Nuclear Information System (INIS)

    DOE

    1997-01-01

    A methodology for performing and applying nuclear criticality safety calculations, for PWR spent nuclear fuel (SNF) packages with actinide-only burnup credit, is described. The changes in the U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241 concentration with burnup are used in burnup credit criticality analyses. No credit for fission product neutron absorbers is taken. The methodology consists of five major steps. (1) Validate a computer code system to calculate isotopic concentrations of SNF created during burnup in the reactor core and subsequent decay. A set of chemical assay benchmarks is presented for this purpose as well as a method for assessing the calculational bias and uncertainty, and conservative correction factors for each isotope. (2) Validate a computer code system to predict the subcritical multiplication factor, k eff , of a spent nuclear fuel package. Fifty-seven UO 2 , UO 2 /Gd 2 O 3 , and UO 2 /PuO 2 critical experiments have been selected to cover anticipated conditions of SNF. The method uses an upper safety limit on k eff (which can be a function of the trending parameters) such that the biased k eff , when increased for the uncertainty is less than 0.95. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. Three bounding axial profiles have been established to assure the ''end effect'' is accounted for conservatively. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). Burnup credit loading curves show the minimum burnup required for a given initial enrichment. The utility burnup record is compared to this requirement after the utility accounts for the uncertainty in its record. Separate curves may be generated for each assembly design, various minimum cooling times and burnable absorber histories. (5) Verify that SNF assemblies meet the package loading criteria and confirm proper assembly selection

  19. Study of production of fuel pellets for a reactor

    International Nuclear Information System (INIS)

    Mendes, Luiz F.F.; Conti, Thadeu N.

    2017-01-01

    Nowadays the electrical energy was been used much on society. A method for getting electricity is through nuclear power plants, this power plant uses fission that occurs inside the UO 2 pellets to generate thermal energy that will be transform into electric. The pellets production was made from enriched UF 6 uses some techniques of reprocessing UF 6 gas to UO 2 powder. This reprocessing process done by wet route (Ammonium Diuranate ADU or Ammonium Uranium Carbonate AUC) or by dry route (Fluidized bed or GECO). With getting of UO 2 powder is forwarded to metallurgy where this powder is compacted in cylindrical matrix so that powder take the desired shape, this green pellets are full of the empty spaces (porosity) for this it is sent to the sintering. The sintering consists of a joint of these particles of powders by means of the heating of this green pellets, coming arrive the melting temperature, the UO 2 molecules melting each other so decrease the porosity and increase the density. For the production of fuel pellets the process all most used is wed route by means the AUC ,this process arrive created for replace the ADU because the AUC is a process where less rework for the pore geometry is required compared to DUA. The fluidized bed process is more used in small samples however, for a large amount it becomes unfeasible, moreover the dry route process require more robust materials because of the generation of HF that is highly corrosive and cannot used the UNH (uranyl nitrate hexahydrate) used for recycle materials discarded in manufacturing. (author)

  20. Lattice cell burnup calculation

    International Nuclear Information System (INIS)

    Pop-Jordanov, J.

    1977-01-01

    Accurate burnup prediction is a key item for design and operation of a power reactor. It should supply information on isotopic changes at each point in the reactor core and the consequences of these changes on the reactivity, power distribution, kinetic characters, control rod patterns, fuel cycles and operating strategy. A basic stage in the burnup prediction is the lattice cell burnup calculation. This series of lectures attempts to give a review of the general principles and calculational methods developed and applied in this area of burnup physics

  1. 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)

  2. Behavior of UO2 and FISSIUM in sodium vapor atmosphere at temperatures up to 28000C

    International Nuclear Information System (INIS)

    Feuerstein, H.; Oschinski, J.

    1986-11-01

    In case of a HCDA a rubble bed of fuel debris may form under a sodium pool and reach high temperatures. An experimental technique was developed to study the behavior of fuel and fission products in out-of-pile tests in a sodium vapor atmosphere. Evaporation rates of UO 2 were measured up to 2800 0 C. The evaporation was found to be a complex process, depending on temperature and the 'active' surface. Evaporation restructures the surface of the samples, however no new 'active' surface is formed. UO 2 forms sometimes well shaped crystals and curious erosion products. The efficiency of the used condenser/filter lines was higher than 99.99%. In case of a HCDA all the evaporated substances will condense in the soidum pool. Thermal reduction of the UO 2 reduces the oxygen potential of the system. The final composition at 2500 0 C was found to be UO 1.95 . The only influence of the sodium vapor was found for the diffusion of UO 2 into the thoria of the crucible. Compared with experiments in an atmosphere of pure argon, the diffusion rate was reduced. (orig.) [de

  3. 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)

  4. Prediction of minimum UO2 particle size based on thermal stress initiated fracture model

    International Nuclear Information System (INIS)

    Corradini, M.

    1976-08-01

    An analytic study was employed to determine the minimum UO 2 particle size that could survive fragmentation induced by thermal stresses in a UO 2 -Na Fuel Coolant Interaction (FCI). A brittle fracture mechanics approach was the basis of the study whereby stress intensity factors K/sub I/ were compared to the fracture toughness K/sub IC/ to determine if the particle could fracture. Solid and liquid UO 2 droplets were considered each with two possible interface contact conditions; perfect wetting by the sodium or a finite heat transfer coefficient. The analysis indicated that particles below the range of 50 microns in radius could survive a UO 2 -Na fuel coolant interaction under the most severe temperature conditions without thermal stress fragmentation. Environmental conditions of the fuel-coolant interaction were varied to determine the effects upon K/sub I/ and possible fragmentation. The underlying assumptions of the analysis were investigated in light of the analytic results. It was concluded that the analytic study seemed to verify the experimental observations as to the range of the minimum particle size due to thermal stress fragmentation by FCI. However the method used when the results are viewed in light of the basic assumptions indicates that the analysis is crude at best, and can be viewed as only a rough order of magnitude analysis. The basic complexities in fracture mechanics make further investigation in this area interesting but not necessarily fruitful for the immediate future

  5. 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

  6. Interaction between UO2 kernel and pyrocarbon coating in irradiated and unirradiated HTR fuel particles

    International Nuclear Information System (INIS)

    Drago, A.; Klersy, R.; Simoni, O.; Schrader, K.H.

    1975-08-01

    Experimental observations on unidirectional UO 2 kernel migration in TRISO type coated particle fuels are reported. An analysis of the experimental results on the basis of data and models from the literature is reported. The stoichiometric composition of the kernel is considered the main parameter that, associated with a temperature gradient, controls the unidirectional kernel migration

  7. The reactive solid-gas flow of a fluidized bed for UO2 conversion

    International Nuclear Information System (INIS)

    Juanico, L.E.

    1991-01-01

    The reactive solid-gas flow of a fluidized bed for UO 2 conversion was modeled. The sedimentation-reaction process was treated using the drift-flux equations. Also, the associated pressure transient due to the reaction gas release was analyzed. An experiment was carried out to compare the results, and pressure transient was numerically simulated, reaching interesting conclusions. (Author) [es

  8. Characterization of hydrogen, nitrogen, oxygen, carbon and sulfur in nuclear fuel (UO2) and cladding nuclear rod materials

    International Nuclear Information System (INIS)

    Crewe, Maria Teresa I.; Lopes, Paula Corain; Moura, Sergio C.; Sampaio, Jessica A.G.; Bustillos, Oscar V.

    2011-01-01

    The importance of Hydrogen, Nitrogen, Oxygen, Carbon and Sulfur gases analysis in nuclear fuels such as UO 2 , U 3 O 8 , U 3 Si 2 and in the fuel cladding such as Zircaloy, is a well known as a quality control in nuclear industry. In UO 2 pellets, the Hydrogen molecule fragilizes the metal lattice causing the material cracking. In Zircaloy material the H2 molecules cause the boiling of the cladding. Other gases like Nitrogen, Oxygen, Carbon and Sulfur affect in the lattice structure change. In this way these chemical compounds have to be measure within specify parameters, these measurement are part of the quality control of the nuclear industry. The analytical procedure has to be well established by a convention of the quality assurance. Therefore, the Oxygen, Carbon, Sulfur and Hydrogen are measured by infrared absorption (IR) and the nitrogen will be measured by thermal conductivity (TC). The gas/metal analyzer made by LECO Co. model TCHEN-600 is Hydrogen, Oxygen and Nitrogen analyzer in a variety of metals, refractory and other inorganic materials, using the principle of fusion by inert gas, infrared and thermo-coupled detector. The Carbon and Sulfur compounds are measure by LECO Co. model CS-400. A sample is first weighed and placed in a high purity graphite crucible and is casted on a stream of helium gas, enough to release the oxygen, nitrogen and hydrogen. During the fusion, the oxygen present in the sample combines with the carbon crucible to form carbon monoxide. Then, the nitrogen present in the sample is analyzed and released as molecular nitrogen and the hydrogen is released as gas. The hydrogen gas is measured by infrared absorption, and the sample gases pass through a trap of copper oxide which converts CO to CO 2 and hydrogen into water. The gases enter the cell where infrared water content is then converted making the measurement of total hydrogen present in the sample. The Hydrogen detection limits for the nuclear fuel is 1 μg/g for the Nitrogen

  9. Technological and licensing challenges for high burnup fuel

    International Nuclear Information System (INIS)

    Gross, H.; Urban, P.; Fenzlein, C.

    2002-01-01

    Deregulation of electricity markets is driving electricity prices downward as well in the U.S. as in Europe. As a consequence high burnup fuel will be demanded by utilities using either the storage or the reprocessing option. At a minimum, burnups consistent with the current political enrichment limit of 5 w/o will be required for both markets.Significant progress has been achieved in the past by Siemens in meeting the demands of utilities for increased fuel burnup. The technological challenges posed by the increased burnup are mainly related to the corrosion and hydrogen pickup of the clad, the high burnup properties of the fuel and the dimensional changes of the fuel assembly structure. Clad materials with increased corrosion resistance appropriate for high burnup have been developed. The high burnup behaviour of the fuel has been extensively investigated and the decrease of thermal conductivity with burnup, the rim effect of the pellet and the increase of fission gas release with burnup can be described, with good accuracy, in fuel rod computer codes. Advanced statistical design methods have been developed and introduced. Materials with increased corrosion resistance are also helpful controlling the dimensional changes of the fuel assembly structure. In summary, most of the questions about the fuel operational behaviour and reliability in the high burnup range have been solved - some of them are still in the process of verification - or the solutions are visible. This fact is largely acknowledged by regulators too. The main licensing challenges for high burnup fuel are currently seen for accident condition analyses, especially for RIA and LOCA. (author)

  10. Study of the experimental parameters for the determination of Ca, Cr, Cu, Fe, Mn and Ni on nuclear grade UO2 by X-ray fluorescence technique

    International Nuclear Information System (INIS)

    Salvador, V.L.R.

    1982-01-01

    An analytical method for the simultaneous determinations of low concentrations of Ca, Cr, Cu, Fe, Mn and Ni on the nuclear grade UO 2 by X-ray fluorescence technique, without the use of chemical treatment, is described. The optimization of the experimental conditions was established on the X-ray fluorescence spectrometer and a low limit of detection (4 - 7 μg/gU) was achieved which satisfies the requirement in the nuclear fuel specification. The samples were prepared in the form of double layer pressed pellets using boric acid as a binding agent. The characteristic first order K sub(α) line intensity of each minor component was measured and the values of its concentrations were deduced using respective standard calibrations curves. The precision, accuracy and acceptability of the method were determined for all elements. The values of the precision are in the range of 2 - 10% and the accuracy are lower than 7%. (Author) [pt

  11. Extended burnup with SEU fuel in Atucha-1 NPP

    International Nuclear Information System (INIS)

    Alvarez, L.; Casario, J.; Fink, J.; Perez, R.; Higa, M.

    2002-01-01

    Atucha-1 is a Pressurized Heavy Water Reactor originally fuelled with natural uranium. Fuel Assemblies consist of 36 fuel rods and the active length is 5300 mm. The total length of the fuel assembly is about 6 m. The average discharge burnup of natural UO 2 fuel is 5900 MWd/tU. After the deregulation of the Argentine electricity market there was an important incentive to reduce the impact of fuel cost on the cost of generation. To keep the competitiveness of the nuclear energy against another sources of electricity it was necessary to reduce the cost of the nuclear fuel. With this objective a program to introduce SEU (0.85 % 235 U) fuel in Atucha-1 was launched in 1993. As a result of this program the average SEU fuel discharge burnup increased to more than 11000 MWd/tU. The first SEU fuels were introduced in Atucha-1 in 1995 and, in the present stage of the program, 71% of core positions are loaded with this type of fuel. This paper describes key aspects of Atucha-1 fuel design and their relevance limiting the burnup extension and shows relevant data regarding the SEU in-reactor performance. At the present time 125 SEU Fuel Assemblies have been irradiated without failures associated with the extended burnup or unfavorable influences on the operation of the power station. (author)

  12. 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

  13. 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)

  14. Cs2SeO4-UO2SeO4-H2O system at 25 deg C

    International Nuclear Information System (INIS)

    Serezhkina, L.B.; Serezhkin, V.N.

    1987-01-01

    Using the method of isothermal solubility at 25 deg C the interaction of cesium and uranyl selenates in aqueous solution is studied. Formation of congruently soluble Cs 2 UO 2 (SeO 4 ) 2 x2H 2 O and Cs 2 (UO 2 ) 2 x(SeO 4 ) 3 is ascertained, their crystallographic characteristics being determined

  15. Fuel pellet

    International Nuclear Information System (INIS)

    Hayashi, K.

    1980-01-01

    Fuel pellet for insertion into a cladding tube in order to form a fuel element or a fuel rod. The fuel pellet has got a belt-like projection around its essentially cylindrical lateral circumferential surface. The upper and lower edges in vertical direction of this belt-like projection are wave-shaped. The projection is made of the same material as the bulk pellet. Both are made in one piece. (orig.) [de

  16. Determination of gas residues in uranium dioxide pellets

    International Nuclear Information System (INIS)

    Riella, H.G.

    1978-01-01

    The measurement of low amounts of residual gases, excluding water, in ceramic grade uranium dioxide pellets, using high temperature vacuum extraction technique, is dealt with. The high temperature extraction gas analysis apparatus was designed and assembled for sequential analysis of up to eight uranium dioxide pellets by run. The system consists of three major units, namely outgassing unit, transfer unit and analytical unit. The whole system is evacuated to a final pressure of less then 10 -5 torr. A weighed pellet is transfered into the outgassing unit for subsequent dropping into a Platinum-Rhodium crucible which is heated inductively up to 1600 0 C during 30 minutes. The released gases are imediately transfered from the outgassing to analytical unit passing through a cold trap at -95 0 C to remove water vapor. The gases are transfered to previously calibrated volumetric bulb where the total pressure and temperature are determined. An estimate of the gas content in the pellets at STP condition is obtained from the measured volume, pressure and temperature of the gas mixture by applying ideal gases equation. Analysis to two lots (fourteen samples) of uranium dioxide pellets by the method described here indicated a mean gas content of 0,060cm 3 /g UO 2 . The lower limit of this technique is 0,03cm 3 /g UO 2 (STP). The time required for the analysis of eight pellets is about 9 hours [pt

  17. Microscopic appearance analysis of raw material used for the production of sintered UO2 by scanning electron microscope

    International Nuclear Information System (INIS)

    Liu feiming

    1992-01-01

    The paper describes the microscopic appearance of UO 2 , U 3 O 8 , ADU and AUC powders used for the production of sintered UO 2 slug of nuclear fuel component of PWR. The characteristic analysis of the microscopic appearance observed by scanning electron microscope shows that the quality and finished product rate of sintered UO 2 depend on the appearance characteristic of the active Uo 2 powder, such as grade size and its distribution, spherulitized extent, surface condition and heap model etc.. The addition of U 3 O 8 to the UO 2 powder improves significantly the quality and the finished product rate. The mechanism of this effect is discussed on the basis of the microscopic appearance characteristic for two kinds of powder

  18. Method for adding additional isotopes to actinide-only burnup credit

    International Nuclear Information System (INIS)

    Lancaster, D.B.; Fuentes, E.; Kang, C.

    1998-01-01

    The Topical Report on Actinide-Only Burnup Credit for Pressurized Water Reactor Spent Nuclear Fuel Packages requires computer code validation to be performed against a benchmark set of chemical assays for isotopic concentration and against a benchmark set of critical experiments for package criticality. Both sets contain all the isotopes included in the methodology. The chemical assays used include the uranium and plutonium isotopes, while the critical experiments were composed of UO 2 or MOX rods, covering the isotopes in the actinide only approach. Since other isotopes are not included in the validation benchmark sets, it would be necessary to justify both the content and worth of any additional isotope for which burnup credit is to be taken (i.e., both the concentration and criticality effect of each particular isotope must be validated). A method is proposed here that can be used for any number of additional isotopes. As does the actinide-only burnup credit methodology, this method makes use of chemical assay data to establish the conservatism in the prediction of each isotope's concentration. Criticality validation is also performed using a benchmark set of UO 2 and MOX critical experiments, where the additional isotopes are validated using worth experiments to conservatively account for any uncertainty in their cross sections. The remaining requirements (analysis and modeling parameters, loading criteria generation, and physical implementation and controls) are performed exactly as described in the actinide-only burnup credit methodology. This report provides insight into each particular requirement in the new methodology

  19. Procedure for the fabrication of ceramic fuel pellets with an adjustable structure

    International Nuclear Information System (INIS)

    Henke, M.; Klemm, U.; Sobek, D.

    1986-01-01

    The invention concerns a procedure for the fabrication of ceramic fuel pellets of UO 2 , PuO 2 , ThO 2 and their mixtures with an adjustable structure. Before or during the milling the particle shaped fuel pellets have been added polyethylenglycol in a 20 - 60 % aqueous solution with an amount of 0.5 - 2.0 % in weight. This additive has an effect on a controlled pore formation and grain growth advancement

  20. Comparative density functional study of the complexes [UO2(CO3)3]4- and [(UO2)3(CO3)6]6- in aqueous solution.

    Science.gov (United States)

    Schlosser, Florian; Moskaleva, Lyudmila V; Kremleva, Alena; Krüger, Sven; Rösch, Notker

    2010-06-28

    With a relativistic all-electron density functional method, we studied two anionic uranium(VI) carbonate complexes that are important for uranium speciation and transport in aqueous medium, the mononuclear tris(carbonato) complex [UO(2)(CO(3))(3)](4-) and the trinuclear hexa(carbonato) complex [(UO(2))(3)(CO(3))(6)](6-). Focusing on the structures in solution, we applied for the first time a full solvation treatment to these complexes. We approximated short-range effects by explicit aqua ligands and described long-range electrostatic interactions via a polarizable continuum model. Structures and vibrational frequencies of "gas-phase" models with explicit aqua ligands agree best with experiment. This is accidental because the continuum model of the solvent to some extent overestimates the electrostatic interactions of these highly anionic systems with the bulk solvent. The calculated free energy change when three mono-nuclear complexes associate to the trinuclear complex, agrees well with experiment and supports the formation of the latter species upon acidification of a uranyl carbonate solution.

  1. Standard specification for sintered gadolinium oxide-uranium dioxide pellets

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2008-01-01

    1.1 This specification is for finished sintered gadolinium oxide-uranium dioxide pellets for use in light-water reactors. It applies to gadolinium oxide-uranium dioxide pellets containing uranium of any 235U concentration and any concentration of gadolinium oxide. 1.2 This specification recognizes the presence of reprocessed uranium in the fuel cycle and consequently defines isotopic limits for gadolinium oxide-uranium dioxide pellets made from commercial grade UO2. Such commercial grade UO2 is defined so that, regarding fuel design and manufacture, the product is essentially equivalent to that made from unirradiated uranium. UO2 falling outside these limits cannot necessarily be regarded as equivalent and may thus need special provisions at the fuel fabrication plant or in the fuel design. 1.3 This specification does not include (1) provisions for preventing criticality accidents or (2) requirements for health and safety. Observance of this specification does not relieve the user of the obligation to be aw...

  2. The role of grain boundary fission gases in high burn-up fuel under reactivity initiated accident conditions

    International Nuclear Information System (INIS)

    Lemoine, F.; Papin, J.; Frizonnet, J.M.; Cazalis, B.; Rigat, H.

    2002-01-01

    In the frame of reactivity-initiated accidents (RIA) studies, the CABRI REP-Na programme is currently performed, focused on high burn-up UO 2 and MOX fuel behaviour. From 1993 to 1998, seven tests were performed with UO 2 fuel and three with MOX fuel. In all these tests, particular attention has been devoted to the role of fission gases in transient fuel behaviour and in clad loading mechanisms. From the analysis of experimental results, some basic phenomena were identified and a better understanding of the transient fission gas behaviour was obtained in relation to the fuel and clad thermo-mechanical evolution in RIA, but also to the initial state of the fuel before the transient. A high burn-up effect linked to the increasing part of grain boundary gases is clearly evidenced in the final gas release, which would also significantly contribute to the clad loading mechanisms. (authors)

  3. Electrochemical studies of the effect of H2 on UO2 dissolution

    International Nuclear Information System (INIS)

    King, F.; Shoesmith, D.W.

    2004-09-01

    This report summarises evidence for the effect of H 2 on the oxidation and dissolution of UO 2 derived from electrochemical studies. In the presence of γ-radiation or with SIMFUEL electrodes containing ε-particles, the corrosion potential (E CORR ) of UO 2 is observed to be suppressed in the presence of H 2 by up to several hundred milli volts. This effect has been observed at room temperature with 5 MPa H 2 (in the case of γ-irradiated solutions) and at 60 deg C with a H 2 partial pressure of only 0.002-0.014 MPa H 2 with the SIMFUEL electrode. The suppression of E CORR in the presence of H 2 indicates that the degree of surface oxidation and the rate of dissolution of UO 2 is lower in the presence of H 2 .The precise mechanism of the effect of H 2 is unclear at this time. The mechanism appears to involve a surface heterogeneous process, rather than a homogeneous solution process. Under some circumstances the value of E CORR approaches the equilibrium potential for the H 2 /H + couple, suggesting galvanic coupling between sites on which this electrochemical process is catalysed and the rest of the UO 2 surface. It is also possible that H* radical species, either produced radiolytically from H 2 O or by dissociation of H 2 on ε-particles or surface-active UO 2+x sites, reduce oxidised U(V)/U(VI) surface states to U(IV). The effect of H 2 on reducing the degree of surface oxidation is only partially reversible, since surfaces reduced in H 2 atmospheres (re-)oxidise more slowly and to a lesser degree than surfaces not previously exposed to H 2 . Homogeneous reactions between dissolved H 2 and either oxidants or dissolved U(VI) cannot explain the observed effects.Regardless of the precise mechanism, the suppression of the degree of surface oxidation results in lower UO 2 dissolution rates in the presence of H 2 . Application of an electro-chemical dissolution model to the observed E CORR values suggests that the fractional dissolution rate of used fuel in the

  4. Criticality experiments with low enriched UO2 fuel rods in water containing dissolved gadolinium

    International Nuclear Information System (INIS)

    Bierman, S.R.; Murphy, E.S.; Clayton, E.D.; Keay, R.T.

    1984-02-01

    The results obtained in a criticality experiments program performed for British Nuclear Fuels, Ltd. (BNFL) under contract with the United States Department of Energy (USDOE) are presented in this report along with a complete description of the experiments. The experiments involved low enriched UO 2 and PuO 2 -UO 2 fuel rods in water containing dissolved gadolinium, and are in direct support of BNFL plans to use soluble compounds of the neutron poison gadolinium as a primary criticality safeguard in the reprocessing of low enriched nuclear fuels. The experiments were designed primarily to provide data for validating a calculation method being developed for BNFL design and safety assessments, and to obtain data for the use of gadolinium as a neutron poison in nuclear chemical plant operations - particularly fuel dissolution. The experiments program covers a wide range of neutron moderation (near optimum to very under-moderated) and a wide range of gadolinium concentration (zero to about 2.5 g Gd/l). The measurements provide critical and subcritical k/sub eff/ data (1 greater than or equal to k/sub eff/ greater than or equal to 0.87) on fuel-water assemblies of UO 2 rods at two enrichments (2.35 wt % and 4.31 wt % 235 U) and on mixed fuel-water assemblies of UO 2 and PuO 2 -UO 2 rods containing 4.31 wt % 235 U and 2 wt % PuO 2 in natural UO 2 respectively. Critical size of the lattices was determined with water containing no gadolinium and with water containing dissolved gadolinium nitrate. Pulsed neutron source measurements were performed to determine subcritical k/sub eff/ values as additional amounts of gadolinium were successively dissolved in the water of each critical assembly. Fission rate measurements in 235 U using solid state track recorders were made in each of the three unpoisoned critical assemblies, and in the near-optimum moderated and the close-packed poisoned assemblies of this fuel

  5. High Burnup Effects Program

    International Nuclear Information System (INIS)

    Barner, J.O.; Cunningham, M.E.; Freshley, M.D.; Lanning, D.D.

    1990-04-01

    This is the final report of the High Burnup Effects Program (HBEP). It has been prepared to present a summary, with conclusions, of the HBEP. The HBEP was an international, group-sponsored research program managed by Battelle, Pacific Northwest Laboratories (BNW). The principal objective of the HBEP was to obtain well-characterized data related to fission gas release (FGR) for light water reactor (LWR) fuel irradiated to high burnup levels. The HBEP was organized into three tasks as follows: Task 1 -- high burnup effects evaluations; Task 2 -- fission gas sampling; and Task 3 -- parameter effects study. During the course of the HBEP, a program that extended over 10 years, 82 fuel rods from a variety of sources were characterized, irradiated, and then examined in detail after irradiation. The study of fission gas release at high burnup levels was the principal objective of the program and it may be concluded that no significant enhancement of fission gas release at high burnup levels was observed for the examined rods. The rim effect, an as yet unquantified contributor to athermal fission gas release, was concluded to be the one truly high-burnup effect. Though burnup enhancement of fission gas release was observed to be low, a full understanding of the rim region and rim effect has not yet emerged and this may be a potential area of further research. 25 refs., 23 figs., 4 tabs

  6. 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.

  7. Grain Size and Phase Purity Characterization of U3Si2 Pellet Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Hoggan, Rita E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Tolman, Kevin R. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Cappia, Fabiola [Idaho National Lab. (INL), Idaho Falls, ID (United States); Wagner, Adrian R. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Harp, Jason M. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2018-05-01

    Characterization of U3Si2 fresh fuel pellets is important for quality assurance and validation of the finished product. Grain size measurement methods, phase identification methods using scanning electron microscopes equipped with energy dispersive spectroscopy and x-ray diffraction, and phase quantification methods via image analysis have been developed and implemented on U3Si2 pellet samples. A wide variety of samples have been characterized including representative pellets from an initial irradiation experiment, and samples produced using optimized methods to enhance phase purity from an extended fabrication effort. The average grain size for initial pellets was between 16 and 18 µm. The typical average grain size for pellets from the extended fabrication was between 20 and 30 µm with some samples exhibiting irregular grain growth. Pellets from the latter half of extended fabrication had a bimodal grain size distribution consisting of coarsened grains (>80 µm) surrounded by the typical (20-30 µm) grain structure around the surface. Phases identified in initial uranium silicide pellets included: U3Si2 as the main phase composing about 80 vol. %, Si rich phases (USi and U5Si4) composing about 13 vol. %, and UO2 composing about 5 vol. %. Initial batches from the extended U3Si2 pellet fabrication had similar phases and phase quantities. The latter half of the extended fabrication pellet batches did not contain Si rich phases, and had between 1-5% UO2: achieving U3Si2 phase purity between 95 vol. % and 98 vol. % U3Si2. The amount of UO2 in sintered U3Si2 pellets is correlated to the length of time between U3Si2 powder fabrication and pellet formation. These measurements provide information necessary to optimize fabrication efforts and a baseline for future work on this fuel compound.

  8. Fuel-pellet-fabrication experience using direct-denitration-recycle-PuO2-coprecipitated mixed oxide

    International Nuclear Information System (INIS)

    Rasmussen, D.E.; Schaus, P.S.

    1980-01-01

    The fuel pellet fabrication experience described in this paper involved three different feed powders: coprecipitated PuO 2 -UO 2 which was flash calcined in a fluidized bed; co-direct denitrated PuO 2 -UO 2 ; and direct denitrated LWR recycle PuO 2 which was mechanically blended with natural UO 2 . The objectives of this paper are twofold; first, to demonstrate that acceptable quality fuel pellets were fabricated using feed powders manufactured by processes other than the conventional oxalate process; and second, to highlight some pellet fabrication difficulties experienced with the direct denitration LWR recycle PuO 2 feed material, which did not produce acceptable pellets. The direct denitration LWR recycle PuO 2 was available as a by-product and was not specifically produced for use in fuel pellet fabrication. Nevertheless, its characteristics and pellet fabrication behavior serve to re-emphasize the importance of continued process development involving both powder suppliers and fuel fabricators to close the fuel cycle in the future

  9. Thermal conductivity of fully dense unirradiated UO2: A new formulation from experimental results between 100 deg. C and 2500 deg. C, and associated fundamental properties

    International Nuclear Information System (INIS)

    Delette, G.; Charles, M.

    1997-01-01

    The various contributions to the thermal conductivity of UO 2 are first reviewed: contribution from phonons is preponderant up to 1600 deg. C; radiative contribution is negligible in the case of a polycrystalline sample, and is unable to account for the increase in conductivity observed above 1600 deg. C; electronic contribution, which seems patently to be responsible for this increase, is efficient from 1400 deg. C (electron-hole pairs treated as small polarons). Given the lack of decisive arguments on their actual efficiency, it was decided that, for temperatures above 2400 deg. C, neither a possible decrease in the electronic contribution due to vacancies, nor an additional possible contribution from Frenkel pairs would be described. We do not go therefore beyond the above-mentioned electronic contribution. In the light of these considerations, the law established by CEA for the thermal conductivity of unirradiated UO 2 , on the basis of a homogeneous set of measurements between 100 deg. C and 2500 deg. C, has been revised. A least-square method applied to the above measurements was used to derive simultaneously the four adjustable constants of the law which is henceforth proposed: λ = 1/A + BT+C/T 2 exp(-W/kT). The values of A, B, C, and W obtained, have the advantage of giving a better account of the high temperature results compared to the Martin recommendations (which has, by the way, the same physical form as above). Furthermore, this new law allows an extrapolation towards the melting point which is better founded than the earlier CEA law, while still respecting the value of the integrated thermal conductivity up to melting. Finally, various burnup effects (impurities, stoichiometry, etc.) can be formulated in physical terms. Moreover, the values of various fundamental properties connected with the derived constants have been determined: Debye temperature, Grueneisen parameter, Mott-Hubbard energy, electrical conductivity. These values are discussed and

  10. Dynamic leaching studies of 48 MWd/kgU UO2 commercial spent nuclear fuel under oxic conditions

    Science.gov (United States)

    Serrano-Purroy, D.; Casas, I.; González-Robles, E.; Glatz, J. P.; Wegen, D. H.; Clarens, F.; Giménez, J.; de Pablo, J.; Martínez-Esparza, A.

    2013-03-01

    The leaching of a high-burn-up spent nuclear fuel (48 MWd/KgU) has been studied in a carbonate-containing solution and under oxic conditions using a Continuously Stirred Tank Flow-Through Reactor (CSTR). Two samples of the fuel, one prepared from the centre of the pellet (labelled CORE) and another one from the fuel pellet periphery, enriched with the so-called High Burn-Up Structure (HBS, labelled OUT) have been used.For uranium and actinides, the results showed that U, Np, Am and Cm gave very similar normalized dissolution rates, while Pu showed slower dissolution rates for both samples. In addition, dissolution rates were consistently two to four times lower for OUT sample compared to CORE sample.Considering the fission products release the main results are that Y, Tc, La and Nd dissolved very similar to uranium; while Cs, Sr, Mo and Rb have up to 10 times higher dissolution rates. Rh, Ru and Zr seemed to have lower dissolution rates than uranium. The lowest dissolution rates were found for OUT sample.Three different contributions were detected on uranium release, modelled and attributed to oxidation layer, fines and matrix release.

  11. Critique of the Board-Hall model for thermal detonations in UO2--Na systems

    International Nuclear Information System (INIS)

    Williams, D.C.

    1976-01-01

    The Board--Hall model for detonating thermal explosions is reviewed and some criticisms are offered in terms of its application to UO 2 -Na systems. The basic concept of a detonation-like thermal explosion is probably valid provided certain fundamental conditions can be met; however, Board and Hall's arguments as to just how these conditions can be met in UO 2 -Na mixtures appear to contain serious flaws. Even as given, the model itself predicts that a very large triggering event is needed to initiate the process. More importantly, the model for shock-induced fragmentation greatly overestimates the tendency for such fragmentation to occur. The shock-dispersive effects of mixtures are ignored. Altogether, the model's deficiencies imply that, as given, it is not applicable to LMFBR accident analysis; nonetheless, one cannot completely rule out the possibility of meeting the fundamental conditions for detonation by other mechanisms

  12. Evaluation of melting point of UO2 by molecular dynamics simulation

    International Nuclear Information System (INIS)

    Arima, Tatsumi; Idemitsu, Kazuya; Inagaki, Yaohiro; Tsujita, Yuichi; Kinoshita, Motoyasu; Yakub, Eugene

    2009-01-01

    The melting point of UO 2 has been evaluated by molecular dynamics simulation (MD) in terms of interatomic potential, pressure and Schottky defect concentration. The Born-Mayer-Huggins potentials with or without a Morse potential were explored in the present study. Two-phase simulation whose supercell at the initial state consisted of solid and liquid phases gave the melting point comparable to the experimental data using the potential proposed by Yakub. The heat of fusion was determined by the difference in enthalpy at the melting point. In addition, MD calculations showed that the melting point increased with pressure applied to the system. Thus, the Clausius-Clapeyron equation was verified. Furthermore, MD calculations clarified that an addition of Schottky defects, which generated the local disorder in the UO 2 crystal, lowered the melting point.

  13. 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

  14. The preparation of UO2 powder: effect of ammonium uranate properties

    International Nuclear Information System (INIS)

    Woolfrey, J.L.

    1978-01-01

    Ammonium uranate (AU) powders were precipitated from a uranyl nitrate solution with gaseous ammonia. The decomposition of the powders in hydrogen was studied to determine those properties of AU which affect the decomposition reactions and influence the properties of the final UO 2 powder. The thermal decomposition was affected by the initial composition (ammonia and nitrate content) and the morphology of the AU powders. The amount of self-reduction increased with increasing combined ammonia content and decreased with increasing nitrate content. The specific surface area of the decomposed powder increased with increasing total ammonia content and initial surface area of the precursor AU powder. Thermal treatment of the decomposed powder can be used to modify such effects and, in commercial powder production, is used to control the properties of the final UO 2 powder. (Auth.)

  15. Large-scale production of UO2 kernels by sol–gel process at INET

    International Nuclear Information System (INIS)

    Hao, Shaochang; Ma, Jingtao; Zhao, Xingyu; Wang, Yang; Zhou, Xiangwen; Deng, Changsheng

    2014-01-01

    In order to supply elements (300,000 elements per year) for the Chinese pebble bed modular high temperature gas cooled reactor (HTR-PM), it is necessary to scale up the production of UO 2 kernels to 3–6 kgU per batch. The sol–gel process for preparation of UO 2 kernels have been improved and optimized at Institute of Nuclear and New Energy Technology (INET), Tsinghua University, PR China, and a whole set of facility was designed and constructed based on the process. This report briefly describes the main steps of the process, the key equipment and the production capacities of every step. Six batches of kernels for scale-up verification and four batches of kernels for fuel elements for in-pile irradiation tests have been successfully produced, respectively. The quality of the produced kernels meets the design requirements. The production capacity of the process reaches 3–6 kgU per batch

  16. A Review of Fragmentation Models Relative to Molten UO2 Breakup when Quenched in Sodium Coolant

    International Nuclear Information System (INIS)

    Cronenberg, A.W.; Grolmes, M.A.

    1976-01-01

    An important aspect of the fuel-coolant interaction problem relative to liquid metal fast breeder reactor (LMFBR) safety analysis is the fragmentation of molten oxide fuel during contact with liquid sodium coolant. A proper description of the kinetics of such an event requires an understanding of the breakup process and an estimate of the size and dispersion of such finely divided fuel in coolant. In recent years, considerable interest has centered on the problem of determining the nature of such fragmentation. In this paper, both analytic and experimental studies pertaining to such breakup are reviewed in light of recent developments in the understanding of heat transfer and solidification phenomena during quenching of UO 2 in sodium. A more extensive review of this subject can be found in Ref. 1. In conclusion: As discussed, a number of models have been proposed in an attempt to understand the nature of the UO 2 fragmentation process. The four principle mechanisms considered likely to cause such fragmentation (impact forces, boiling, violent gas release, and shell solidification) have been developed to the point where comparative analysis is possible. In addition, recent developments in the understanding of the physics of oxide fuel behavior in sodium coolant (boiling regime criteria, vapor nucleation theories, and prediction of solidification kinetics enable us to asses whether or not the various model assumptions are realistic. In view of this knowledge the following conclusions are made. For the case of hydrodynamic influence on fragmentation, it can be said that although the disruptive forces of impact and viscous drag may contribute to breakup, their effects are not controlling with respect to high temperature materials, including UO 2 -sodium. With respect to the vapor bubble growth and collapse mechanism it was shown that for sodium quenching, where coolant contact may, be expected (as opposed to water), the thermodynamic work potential of the bubble is

  17. Fission gas release and grain growth in THO2-UO2 fuel irradiated at high temperature

    International Nuclear Information System (INIS)

    Goldberg, I.; Waldman, L.A.; Giovengo, J.F.; Campbell, W.R.

    1979-01-01

    Data are presented on fission gas release and grain growth in ThO 2 -UO 2 fuels irradiated as part of the LWBR fuel element development program. These data for rods that experienced peak linear power outputs ranging from 15 to 22 KW/ft supplement fission gas release data previously reported for 51 rods containing ThO 2 and ThO 2 -UO 2 fuel irradiated at peak linear powers predominantly below 14 KW/ft. Fission gas release was relatively high (up to 15.0 percent) for the rods operated at high power in contrast to the relatively low fission gas release (0.1 to 5.2 percent) measured for the rods operated at lower power. Metallographic examination revealed extensive equiaxed grain growth in the fuel at the high power axial locations of the three rods

  18. Preparation of UO2 dense spherical particles by sol-gel technique

    International Nuclear Information System (INIS)

    Urbanek, V.; Dolezal, J.

    1977-01-01

    The results of the basic research and development of processes of preparation of dense UO 2 spherical particles by sol-gel technique are presented. Attention was paid to the study of chemistry of internal gelation step in the uranylnitrate-urea-hexamethylentetramine system. The existence regions of several stable gels with different properties were established in connection with variable ratio of basic gel's components and the appropriate ''Phase diagrams'' were drawn. From these diagrams, two of the most interesting types of uranyl gels were chosen for the subsequent thermal processing which included drying, reduction and sintering. The detailed studies of each step of the whole process enabled preparation of UO 2 dense spheres with well defined microstructure

  19. UO2 corrosion in high surface-area-to-volume batch experiments

    International Nuclear Information System (INIS)

    Bates, J. K.; Finch, R. J.; Hanchar, J. M.; Wolf, S. F.

    1997-01-01

    Unsaturated drip tests have been used to investigate the alteration of unirradiated UO 2 and spent UO 2 fuel in an unsaturated environment such as may be expected in the proposed repository at Yucca Mountain. In these tests, simulated groundwater is periodically injected onto a sample at 90 C in a steel vessel. The solids react with the dripping groundwater and water condensed on surfaces to form a suite of U(VI) alteration phases. Solution chemistry is determined from leachate at the bottom of each vessel after the leachate stops interacting with the solids. A more detailed knowledge of the compositional evolution of the leachate is desirable. By providing just enough water to maintain a thin film of water on a small quantity of fuel in batch experiments, we can more closely monitor the compositional changes to the water as it reacts to form alteration phases

  20. New interpretation on formation of UO2 Post-Accident Heat Removal particulate in sodium

    International Nuclear Information System (INIS)

    Schins, H.

    1986-01-01

    A comparative experimental study on quenching in sodium of four molten fuel materials, UO 2 Al 2 P 3 , Cu and stainless steel, is presented. Experimental results like temperatures, pressures, particle shapes, particle size distributions, crack patterns and crystal grain sizes are given and interpreted. These fuel-coolant interactions (FCI) can be understood as all being characterized by transition boiling of sodium. The fuel is first fragmented by the sodium vapor bubble growth and collapse process. These particulates have smooth surfaces. The two materials, UO 2 and Al 2 O 3 , are fragmented further by a delayed mechanism which is thermal stress shrinkage cracking. Delayed particles are fragments of larger ones. Furthermore, attention is drawn to the theoretical results which show that pure FCI-particulate is significantly finer

  1. Burnup performance of OTTO cycle pebble bed reactors with ROX fuel

    International Nuclear Information System (INIS)

    Ho, Hai Quan; Obara, Toru

    2015-01-01

    Highlights: • A 300 MW t Small Pebble Bed Reactor with Rock-like oxide fuel is proposed. • Using ROX fuel can achieve high discharged burnup of spent fuel. • High geological stability can be expected in direct disposal of the spent ROX fuel. • The Pebble Bed Reactor with ROX fuel can be critical at steady state operation. • All the reactor designs have a negative temperature coefficient. - Abstract: A pebble bed high-temperature gas-cooled reactor (PBR) with rock-like oxide (ROX) fuel was designed to achieve high discharged burnup and improve the integrity of the spent fuel in geological disposal. The MCPBR code with a JENDL-4.0 library, which developed the analysis of the Once-Through-Then-Out (OTTO) cycle in PBR, was used to perform the criticality and burnup analysis. Burnup calculations for eight cases were carried out for both ROX fuel and a UO 2 fuel reactor with different heavy-metal loading conditions. The effective multiplication factor of all cases approximately equalled unity in the equilibrium condition. The ROX fuel reactor showed lower FIFA than the UO 2 fuel reactor at the same heavy-metal loading, about 5–15%. However, the power peaking factor and maximum power per fuel ball in the ROX fuel core were lower than that of UO 2 fuel core. This effect makes it possible to compensate for the lower-FIFA disadvantage in a ROX fuel core. All reactor designs had a negative temperature coefficient that is needed for the passive safety features of a pebble bed reactor

  2. Increase of thermal conductivity of uranium dioxide nuclear fuel pellets with beryllium oxide addition

    International Nuclear Information System (INIS)

    Camarano, D.M.; Mansur, F.A.; Santos, A.M.M. dos; Ferraz, W.B.

    2016-01-01

    The UO_2 fuel is one of the most used nuclear fuel in thermal reactors and has many advantages such as high melting point, chemical compatibility with cladding, etc. However, its thermal conductivity is relatively low, which leads to a premature degradation of the fuel pellets due to a high radial temperature gradient during reactor operation. An alternative to avoid this problem is to increase the thermal conductivity of the fuel pellets, by adding beryllium oxide (BeO). Pellets of UO_2 and UO_2-BeO were obtained from a homogenized mixture of powders of UO_2 and BeO, containing 2% and 3% by weight of BeO and sintering at 1750 °C for 3 h under H_2 atmosphere after uniaxial pressing at 400 MPa. The pellet densities were obtained by xylol penetration-immersion method and the thermal diffusivity, specific heat and thermal conductivity were determined according to ASTM E-1461 at room temperature (25 deg C) and 100 deg C. The thermal diffusivity measurements were carried out employing the laser flash method. The thermal conductivity obtained at 25 deg C showed an increase with the addition of 2% and 3% of BeO corresponding to 19% and 28%, respectively. As for the measurements carried out at 100 deg C, there was an increase in the thermal conductivity for the same BeO contents of 20% and 31%. These values as a percentage of increased conductivity were obtained in relation to the UO_2 pellets. (author)

  3. 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.

  4. Investigation of the CANLUB/sheath interface in CANDU fuel at extended burnup by XPS and SEM/WDX

    International Nuclear Information System (INIS)

    Hocking, W.H.; Behnke, R.; Duclos, A.M.; Gerwing, A.F.; Chan, P.K.

    1997-01-01

    A systematic investigation of the fuel-sheath interface in CANDU fuel as a function of extended burnup has been undertaken by XPS and SEM/WDX analysis. Adherent deposits of UO 2 and fission products, including Cs, Ba, Rb, I, Te, Cd and possibly Ru, have been routinely identified on CANLUB coated and bare Zircaloy surfaces. Some trends in the distribution and chemistry of key fission products have begun to emerge. Several potential mechanisms for degradation of the CANLUB graphite layer at high burnup have been practically excluded. New evidence of carbon relocation within the fuel element and limited reaction with excess oxygen has also been obtained. (author)

  5. 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

  6. 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)

  7. 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

  8. The effect of UO2 density on fission product gas release and sheath expansion

    International Nuclear Information System (INIS)

    Notley, M.J.F.; MacEwan, J.R.

    1965-03-01

    The effect of UO 2 density on fission product gas release and sheath expansion has been determined in an irradiation experiment in which the performance of fuel elements with densities between 10.42 and 10.74 g/cm 3 was compared at ∫λdθ values of 39 and 42 W/cm. The elements were irradiated as clusters of four in a pressurized water loop, hence their irradiation histories were identical. Fission product gas release and the extend of grain growth were greater for the lower density elements. Both effects can be attributed solely to the variation of the thermal conductivity of the fuel with the fractional porosity p, if λ p λ [1 - (2.6 ± 0.8) p] where λ is the thermal conductivity of fully dense UO 2 and λ p is that of the porous UO 2 . This expression is in agreement with laboratory findings. A correlation between the extent of grain growth in the UO 2 and the fractional gas release was found to exist in this test and was shown to apply in a large number of other fuel irradiations. Diametral sheath strain was lower for the low density fuel elements than for those of high density, although the former were deduced to have operated with higher central temperatures. It is supposed that the thermal expansion of the fuel can be partially accommodated by elimination of some of the original porosity. The data are consistent with the assumption that approximately half the porosity in the region of the fuel undergoing grain growth is eliminated. (author)

  9. Determination of fluoride content in UO2F2 and ADUF solution by ion selective electrode

    International Nuclear Information System (INIS)

    Samanta, Papu; Kumar, Pradeep; Bagchi, A.C.

    2017-01-01

    During production of uranium metal powder, liquid solution UO 2 F 2 and ADUF containing high content of fluoride gets generated. Fluoride being corrosive in nature, fluorides concentration needs to determined. Ion selective electrode, LaF 3 (Eu) crystal, has been used. Uranium was found to interfere with fluoride analysis. Study was carried out to selectively remove uranium by solvent extraction employing D2EHPA+Cyanex 923 and TBP in dodecane. The TBP was found effective to remove uranium. (author)

  10. Cyclic process for re-use of waste water generated during the production of UO2

    International Nuclear Information System (INIS)

    Crossley, T.J.

    1976-01-01

    The process is described whereby waste water produced during the hydrolysis and ammonium hydroxide treatment of UF 6 to produce ammonium diuranate is recycled for reuse. The solution containing large amounts of ammonia and fluorides and trace amounts of uranium is first treated with lime to precipitate the fluoride. The ammonia is distilled off and recycled to UO 2 F 2 treatment vessel. The CaF 2 precipitate is separated by centrifugation and the aqueous portion is passed through cationic exchange beds

  11. Temperature distribution on fuel rods: a study on the effect of eccentricity in the position of UO2 pellets

    International Nuclear Information System (INIS)

    Gaspar Junior, Joao Carlos Aguiar

    2010-01-01

    This work proposes the development of a method of solving equations of heat transfer applied in fuel rods using the finite element method, in order to evaluate the performance and safety of the nuclear system. Was prepared in a Fortran program to evaluate the equations governing the problem, the boundary conditions and apply the properties of materials on a steady state. This program uses the mesh generation input and graphical output generated by the program GID. The method was validated against the analytical solution found in the book Todreas and Kazimi with error less than 0.2% and with respect to the improved analytical solution of Nijsing for axisymmetry rod and eccentricity rod with error less than a 3.6%. Applications have been developed with the use of correlations for properties with the temperature dependence of resolution axisymmetry rod and the resolution of a rod with eccentricity. The method developed, should it be implemented, would allow the assessment of fuel rods in the given situations and other scenarios, as well as adding a tool of substantial value in the analysis of rods. (author)

  12. 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.

  13. 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

  14. The determination of UO2 and UF4 in fused fluoride salts

    International Nuclear Information System (INIS)

    Batiste, D.J.; Lee, D.A.

    1989-01-01

    The determination of uranium oxide solubilities in fused fluoride salts is important in the electrolytic preparation of uranium metal. This project was initiated to develop a method for the determination of UO 2 separately from UF 4 in UF 4 -CaF 2 -LiF fused salts. Previous methods used for the determination of UO 2 in fused fluoride salts involved inert gas fusions where oxygen was liberated as CO 2 , and hydrofluorination where oxygen was released as H 2 O; but the special equipment used for these procedures was no longer available. These methods assumed that all of the oxygen liberated was due to UO 2 and does not consider impurities from reagents and other oxygen sources that amount to a bias of approximately 0.3 wt %. This titrimetric method eliminates the bias by selectively extracting the UF 4 with a Na 2 EDTA-H 3 BO 3 solution. The remaining uranium oxide residue is treated and titrated gravimetrically to a potentiometric endpoint with NBS standard K 2 Cr 2 O 7 . An aliquot of the Na 2 EDTA-H 3 BO 3 extract is also titrated gravimetrically to a potentiometric endpoint, this uranium component is determined and calculated as UF 4 . 4 refs., 2 figs., 2 tabs

  15. Electronic structure analysis of UO2 by X-ray absorption spectroscopy

    International Nuclear Information System (INIS)

    Ozkendir, O.M.

    2009-01-01

    Full text: Due to the essential role of Actinides in nuclear science and technology, electronic and structural investigations of actinide compounds attract major interest in science. Electronic structure of actinide compounds have important properties due to narrow 5f states which play key role in bonding with anions. The properties of Uranium has been a subject of enduring interest due to its being a major importance as a nuclear fuel and is the highest numbered element which can be found naturally on earth. UO 2 forms as a secondary uranyl group occurred during metamictization of uranium oxide compounds [1].Uranium oxide thin films have been investigated by X-ray Absorption Fine Structure spectroscopy (XAFS) [2]. The full multiple scattering approach has been applied to the calculation of U L3 edge spectra of UO 2 . The calculations are based on different choices of one electron potentials according to Uranium coordinations by using the real space multiple scattering method FEFF 8.2 code [3,4]. U L3-edge absorption spectrum in UO 2 is compared with U L3-edges in USiO 4 and UTe which are chosen due to their different electronic and chemical structures.We have found prominent changes in the XANES spectra of Uranium oxide thin films due to valency properties. Such observed changes are explained by considering the structural, electronic and spectroscopic properties. (author)

  16. 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

  17. Kinetics of UO2(s) dissolution under reducing conditions: Numerical modelling

    International Nuclear Information System (INIS)

    Puigdomenech, I.; Casas, I.; Bruno, J.

    1990-05-01

    A numerical model is presented that describes the dissolution and precipitation of UO 2 (s) under reducing conditions. For aqueous solutions with pH>4, main reaction is: UO 2 (s)+2H 2 O↔U(OH) 4 (aq). The rate constant for the precipitation reaction is found to be log(k p )=-1.2±0.2 h -1 m -2 , while the value for the rate constant of the dissolution reaction is log(k d )=-9.0±0.2 mol/(1 h m 2 ). Most of the experiments reported in the literature show a fast initial dissolution of a surface film of hexavalent uranium oxide. Making the assumption that the chemical composition of the surface coating is U 3 O 7 (s), we have derived a mechanism for this process, and its rate constants have been obtained. The influence of HCO 3 - and CO 3 2- on the mechanism of dissolution and precipitation of UO 2 (s) is still unclear. From the solubility measurements reported, one may conclude that the identity of the aqueous complexes in solution is not well known. Therefore it is not possible to make a mechanistic interpretation of the kinetic data in carbonate medium. (orig.)

  18. The influence of porosity on the thermal conductivity of irradiated UO2 fuel

    International Nuclear Information System (INIS)

    Bakker, K.; Kwast, H.; Cordfunke, E.H.P.

    1994-12-01

    The influence of porosity on the thermal conductivity of irradiated UO 2 fuel has been determined with the Finite Element Method (FEM). Light-microscopy photographs were made of the fuel. The pore shape and the pore distribution are entered in the FEM program from these photographs. The two dimensional (2D) thermal conductivity in the plane of the photograph is obtained from the FEM calculations. The 2D thermal conductivity, that has no physical meaning itself, is the lower limit of the three dimensional (3D) thermal conductivity. For three well defined pore shapes the relation is determined between the 2D thermal conductivity and the 3D thermal conductivity. From these computations a simple relation is obtained that transfers the 2D thermal conductivity into the 3D thermal conductivity, independent of the pore shape. The influence of porosity on the 3D thermal conductivity of irradiated UO 2 fuel and UO 2 fuel doped with Nb 2 O 5 was computed with the FEM. (orig.)

  19. Theoretical comparative study of the industrial fabrication routes for UO2 powder

    International Nuclear Information System (INIS)

    Gonzaga, Reinaldo; Goncalves, Joao da Silva

    2008-01-01

    UO 2 powder is produced in an industrial scale by different fabrication routes, divided into dry and wet routes, or a combination of both. The wet processes most often used industrially are the ADU and AUC processes, whose names originate in the intermediate precipitate obtained during powder fabrication, Ammonium Diuranate and Ammonium Uranil Carbonate. Considering the dry processes, the most widely used ones are the DC (Dry Conversion) and IDR (Integrate Dry Route) process. As to the differences and peculiarities among the fabrication routes, each has marked advantages and disadvantages that are of extreme importance when it comes to selecting and establishing a UO 2 powder production plant based on a particular fabrication route. Among the important factors of comparison to be considered are the product quality characteristics, production capability, quantity of waste, operating costs of each process with raw material, labor, etc. This paper is intended to make a theoretical comparison between wet and dry processes for UO 2 powder fabrication, taking as the basis the previously mentioned factors of comparison. (author)

  20. 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

  1. In-situ TEM observation of nano-void formation in UO2 under irradiation

    Science.gov (United States)

    Sabathier, C.; Martin, G.; Michel, A.; Carlot, G.; Maillard, S.; Bachelet, C.; Fortuna, F.; Kaitasov, O.; Oliviero, E.; Garcia, P.

    2014-05-01

    Transmission electron microscopy (TEM) observations of UO2 polycrystals irradiated in situ with 4 MeV Au ions were performed at room temperature (RT) to better understand the mechanisms of cavity and ultimately fission products nucleation in UO2. Experiments were carried out at the JANNuS Orsay facility that enables in situ ion irradiations inside the microscope to be carried out. The majority of 4 MeV gold ions were transmitted through the thin foil, and the induced radiation defects were investigated by TEM. Observations showed that nano-void formation occurs at ambient temperature in UO2 thin foils irradiated with energetic heavy ions under an essentially nuclear energy loss regime. The diameter and density of nano-objects were measured as a function of the gold irradiation dose at RT. A previous paper has also revealed a similar nano-object population after a Xe implantation performed at 390 keV at 870 K. The nano-object density was modelled using simple concepts derived from Classical Molecular Dynamics simulations. The results are in good agreement, which suggests a mechanism of heterogeneous nucleation induced by energetic cascade overlaps. This indicates that nano-void formation mechanism is controlled by radiation damage. Such nanovoids are likely to act as sinks for mobile fission products during reactor operation.

  2. 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

  3. Appropriate burnup measurements for transportation burnup credit

    International Nuclear Information System (INIS)

    Lancaster, D.; Fuentes, E.

    1997-01-01

    This paper addresses two of the measurement specifications used in analyzing spent fuel packages to gain burnup credit. The philosophy and calculation of rejection criteria and measurement accuracy are discussed. Any assembly for which the declared measured value and reactor record value deviate by more than 10% will be rejected. Measurement accuracy requirements are established for dependent and independent systems. The requirements have been tested and are achievable, ensuring safe operation without extra cost. 6 refs

  4. High burnup models in computer code fair

    Energy Technology Data Exchange (ETDEWEB)

    Dutta, B K; Swami Prasad, P; Kushwaha, H S; Mahajan, S C; Kakodar, A [Bhabha Atomic Research Centre, Bombay (India)

    1997-08-01

    An advanced fuel analysis code FAIR has been developed for analyzing the behavior of fuel rods of water cooled reactors under severe power transients and high burnups. The code is capable of analyzing fuel pins of both collapsible clad, as in PHWR and free standing clad as in LWR. The main emphasis in the development of this code is on evaluating the fuel performance at extended burnups and modelling of the fuel rods for advanced fuel cycles. For this purpose, a number of suitable models have been incorporated in FAIR. For modelling the fission gas release, three different models are implemented, namely Physically based mechanistic model, the standard ANS 5.4 model and the Halden model. Similarly the pellet thermal conductivity can be modelled by the MATPRO equation, the SIMFUEL relation or the Halden equation. The flux distribution across the pellet is modelled by using the model RADAR. For modelling pellet clad interaction (PCMI)/ stress corrosion cracking (SCC) induced failure of sheath, necessary routines are provided in FAIR. The validation of the code FAIR is based on the analysis of fuel rods of EPRI project ``Light water reactor fuel rod modelling code evaluation`` and also the analytical simulation of threshold power ramp criteria of fuel rods of pressurized heavy water reactors. In the present work, a study is carried out by analysing three CRP-FUMEX rods to show the effect of various combinations of fission gas release models and pellet conductivity models, on the fuel analysis parameters. The satisfactory performance of FAIR may be concluded through these case studies. (author). 12 refs, 5 figs.

  5. High burnup models in computer code fair

    International Nuclear Information System (INIS)

    Dutta, B.K.; Swami Prasad, P.; Kushwaha, H.S.; Mahajan, S.C.; Kakodar, A.

    1997-01-01

    An advanced fuel analysis code FAIR has been developed for analyzing the behavior of fuel rods of water cooled reactors under severe power transients and high burnups. The code is capable of analyzing fuel pins of both collapsible clad, as in PHWR and free standing clad as in LWR. The main emphasis in the development of this code is on evaluating the fuel performance at extended burnups and modelling of the fuel rods for advanced fuel cycles. For this purpose, a number of suitable models have been incorporated in FAIR. For modelling the fission gas release, three different models are implemented, namely Physically based mechanistic model, the standard ANS 5.4 model and the Halden model. Similarly the pellet thermal conductivity can be modelled by the MATPRO equation, the SIMFUEL relation or the Halden equation. The flux distribution across the pellet is modelled by using the model RADAR. For modelling pellet clad interaction (PCMI)/ stress corrosion cracking (SCC) induced failure of sheath, necessary routines are provided in FAIR. The validation of the code FAIR is based on the analysis of fuel rods of EPRI project ''Light water reactor fuel rod modelling code evaluation'' and also the analytical simulation of threshold power ramp criteria of fuel rods of pressurized heavy water reactors. In the present work, a study is carried out by analysing three CRP-FUMEX rods to show the effect of various combinations of fission gas release models and pellet conductivity models, on the fuel analysis parameters. The satisfactory performance of FAIR may be concluded through these case studies. (author). 12 refs, 5 figs

  6. Preparations of high density (Th,U)O2 pellets

    International Nuclear Information System (INIS)

    Akabori, Mitsuo; Ikawa, Katsuichi

    1986-07-01

    Preparations of high density and homogeneous (Th,U)O 2 pellets by a powder metallurgy method were examined. (Th,U)O 2 powders were prepared by calcining coprecipitates of ammonium uranate and thorium hydroxide derived from nitrates and mixed sols, and by calcining mixed oxalates precipitated from nitrates. (Th,U)O 2 pellets were characterized with respect to sinterability, lattice parameter, microstructure, homogeneity and stoichiometry. Sintering atmospheres had a significant effect upon all the properties of the derived pellets. The sinterability of (Th,U)O 2 was most favourable in oxidizing and reducing atmospheres for ThO 2 -rich and UO 2 -rich compositions, respectively, and can be enhanced by presence of water vapour in sintering atmospheres. In addition, highly homogeneous (Th,U)O 2 pellets with 99 % in theoretical density were derived from the sol powders. (author)

  7. Numerical solution of stiff burnup equation with short half lived nuclides by the Krylov subspace method

    International Nuclear Information System (INIS)

    Yamamoto, Akio; Tatsumi, Masahiro; Sugimura, Naoki

    2007-01-01

    The Krylov subspace method is applied to solve nuclide burnup equations used for lattice physics calculations. The Krylov method is an efficient approach for solving ordinary differential equations with stiff nature such as the nuclide burnup with short lived nuclides. Some mathematical fundamentals of the Krylov subspace method and its application to burnup equations are discussed. Verification calculations are carried out in a PWR pin-cell geometry with UO 2 fuel. A detailed burnup chain that includes 193 fission products and 28 heavy nuclides is used in the verification calculations. Shortest half life found in the present burnup chain is approximately 30 s ( 106 Rh). Therefore, conventional methods (e.g., the Taylor series expansion with scaling and squaring) tend to require longer computation time due to numerical stiffness. Comparison with other numerical methods (e.g., the 4-th order Runge-Kutta-Gill) reveals that the Krylov subspace method can provide accurate solution for a detailed burnup chain used in the present study with short computation time. (author)

  8. Development Status of a CVD System to Deposit Tungsten onto UO2 Powder via the WCI6 Process

    Science.gov (United States)

    Mireles, O. R.; Kimberlin, A.; Broadway, J.; Hickman, R.

    2014-01-01

    Nuclear Thermal Propulsion (NTP) is under development for deep space exploration. NTP's high specific impulse (> 850 second) enables a large range of destinations, shorter trip durations, and improved reliability. W-60vol%UO2 CERMET fuel development efforts emphasize fabrication, performance testing and process optimization to meet service life requirements. Fuel elements must be able to survive operation in excess of 2850 K, exposure to flowing hydrogen (H2), vibration, acoustic, and radiation conditions. CTE mismatch between W and UO2 result in high thermal stresses and lead to mechanical failure as a result UO2 reduction by hot hydrogen (H2) [1]. Improved powder metallurgy fabrication process control and mitigated fuel loss can be attained by coating UO2 starting powders within a layer of high density tungsten [2]. This paper discusses the advances of a fluidized bed chemical vapor deposition (CVD) system that utilizes the H2-WCl6 reduction process.

  9. Consequences of the increase of burnup on the fuel

    International Nuclear Information System (INIS)

    Melin, P.; Lavoine, O.; Houdaille, B.

    1986-04-01

    The examinations carried out on the FRAGEMA fuel of EDF reactors show its good behavior in service. The results of research and development programs developed by EDF, FGA and the CEA show that this fuel can be irradiated up to a high burnup, and allow to point out the axies of research to improve still the performance of the product in a more and more soliciting environment (increase of power and burnup coupled with load following). Among the solutions considered, there are the design and fabrication adjustments (geometry, initial pressurization), more fundamental changes concerning fuel cans and fuel pellets, which need still research and development programs [fr

  10. Fission gas release from fuel at high burnup

    International Nuclear Information System (INIS)

    Meyer, R.O.; Beyer, C.E.; Voglewede, J.C.

    1978-03-01

    The release of fission gas from fuel pellets at high burnup is reviewed in the context of the safety analysis performed for reactor license applications. Licensing actions are described that were taken to correct deficient gas release models used in these safety analyses. A correction function, which was developed by the Nuclear Regulatory Commission staff and its consultants, is presented. Related information, which includes some previously unpublished data, is also summarized. The report thus provides guidance for the analysis of high burnup gas release in licensing situations

  11. Determination of trace elements in ceramic uranium dioxide pellets powders CRMs by ICP-AES

    International Nuclear Information System (INIS)

    Liu Husheng; Li Jun

    1997-01-01

    The 237-quaternary ammonium extraction resin chromatography is used to the separation of 6 trace elements in ceramic uranium dioxide pellets powders, which are used as certified reference materials (CRMs). The sample is dissolved in 6.5 mol/L HNO 3 and uranium is separated by chromatographic column. the 6 trace elements Al, Ba, Co, Ta, Ti and V contained in the elutriant are determined by using ICP directly reading spectrometer. For a 300 mg sample, the lowest determinable concentration of impurities in ceramic UO 2 pellets powders CRMs is (0.016-0.250) x 10 -6 . The relative standard deviation is less than 7.5%. The proposed method provides excellent and accurate analytical data for the ceramic UO 2 pellets powders samples (CRMs)

  12. The MgSeO4-UO2SeO4-H2O system at 25 deg C

    International Nuclear Information System (INIS)

    Serezhkina, L.B.; Serezhkin, V.N.

    1984-01-01

    The method of isothermal solubility at 25 deg C has been used to study MgSeO 4 -UO 2 SeO 4 -H 2 O system.