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80 articles found

L0167 – Preparation of morphology controlled Th1?xUxO2 sintered pellets from low-temperature precursors

Dense sintered samples of Th1 ? xUxO2 solid solutions were prepared from the initial precipitation of oxalate precursors through two different wet chemical routes, based either on the direct precipitation of the cations or on the use of hydrothermal method. For both low-temperature precursors, the specific surface area was followed versus the heating temperature and the influence of the conversion step on the oxide powder reactivity was evidenced since it allowed to obtain reactive surfaces in the range of 15–45 m2 g?1 without any additional grinding step. From dilatometric studies, the operating conditions required for the complete densification of the Th1 ? xUxO2 pellets were set to a heat treatment of 3 h at 1500 °C. In these conditions, the density of the samples lies between 94% and 99% of the calculated value whatever the preparation method chosen which appeared very promising compared to the results already reported under inert atmosphere. The initial precipitation of low-temperature precursors thus allowed to lower the sintering temperature by about 100 °C while the use of hydrothermal conditions significantly improved the cationic distribution in the sintered samples, as shown from EPMA statistical experiments.
N. Hingant, N. Clavier, N. Dacheux, S. Hubert, N. Barré, R. Podor, L. Aranda, Powder Technology 208 (2011) 454–460

L0161 – Densification behaviour and sintering kinetics of ThO2–4%UO2 pellet

ThO2–?4% 233UO2 fuel will be the driver fuel for the forthcoming Advanced Heavy Water Reactor (AHWR) in India. Densification behaviour such as shrinkage and shrinkage rates of the green pellets of ThO2–4wt.% UO2 (natural ‘U’) fabricated by Coated Agglomerate Pelletization (CAP) process were studied using a vertical dilatometer at different heating rates. Activation energy of sintering, ‘Q’, was estimated in the initial stages of sintering by continuous rate of heating (CRH) technique as proposed by ‘Wang and Rishi Raj’ and ‘Young and Cutler’. The sintering mechanism was identified to be as the grain boundary diffusion (GBD) and the average ‘Q’ value obtained by these two methods were found to be 350 ± 16 kJ/mole and 358 ± 5 kJ/mole, respectively.
Joydipta Banerjee, T.R.G. Kutty, Arun Kumar, H.S. Kamath, Srikumar Banerjee, Journal of Nuclear Materials 408 (2011) 224–230

L0155 – Thermophysical properties of U2Mo intermetallic

The current paper has investigated specific heat capacity, coefficient of thermal expansion and phase transition temperatures of U2Mo intermetallic. The transformation of U2Mo phase to bcc (?-uranium) phase occurred at 853 K. The coefficient of thermal expansion has been determined in the temperature range 423–873 K to be 14.65 × 10?6 K?1. The specific heat data showed a smooth curve up to 773 K and above it a change in trend to follow a ?-shape was observed. The specific heat values of U2Mo have been found to be lower than that calculated from the additivity rule. This study provides first time data on U2Mo. The data obtained from this investigation were compared to available literature on U–Mo alloys
T.R.G. Kutty, Smruti Dash, Joydipta Banerjee, Santu Kaity, Arun Kumar, C.B. Basak, Journal of Nuclear Materials 420 (2012) 193–197

L0153 – Combining in situ HT-ESEM observations and dilatometry: An original and fast way to the sintering map of ThO2

The sintering of ThO2 pellets prepared from the initial precipitation of thorium oxalate was investigated by the means of HT-ESEM observations and dilatometric measurements. On the one hand, the use of environmental microscope allowed the in situ observation of the pellet behaviour during heat treatments between 1250 °C and 1400 °C. Subsequent image analysis led to the determination of local (i.e. at the grain scale) and global (i.e. at the pellet scale) kinetic parameters. Particularly, the average grain size was plotted versus the holding time for the three considered temperatures. On the other hand, analogous experiments were performed by dilatometry and led to monitor both linear shrinkage and density of the samples. On this basis, the combination of the two sets of data allowed us to establish for the first time a sintering map for ThO2. This latter clearly evidenced two different zones driven by densification (d ? 92%) then by grain growth (d ? 92%) which can be used efficiently to monitor the final microstructure of sintered thorium oxide.
N. Clavier, R. Podor, L. Deliere, J. Ravaux, N. Dacheux, Materials Chemistry and Physics Volume 137, Issue 3, 15 January 2013, Pages 742–749

L0137 – Thermal Expansion of Simulated Fuels with Dissolved Fission Products in a UO2 Matrix

As a part of the DUPIC (direct use of spent PWR fuel in CANDU reactors) fuel development program, the thermal expansion of simulated spent fuel pellets with dissolved fission products has been studied by using a thermo-mechanical analyzer (TMA) in the temperature range from 298K to 1773K to investigate the effects of fission products forming solid solutions in a UO2 matrix on the thermal expansions. Simulated fuels with an equivalent burn-up of (30 to 120) GWd/tU were used in this study. The linear thermal expansions of the simulated fuel pellets were higher than that of UO2, and the difference between these fuel pellets and UO2 increased monotonically with temperature. For the temperature range from 298K to 1773K, the values of the average linear thermal expansion coefficients for UO2 and simulated fuels with an equivalent burn-up of (30, 60, and 120) GWd/tU are 1.19 × 10?5 K?1, 1.22 × 10?5 K?1, 1.26 × 10?5 K?1, and 1.32 × 10?5 K?1, respectively.
K. H. Kang, S. H. Na, C. J. Park, Y. H. Kim, K. C. Song, S. H. Lee, S. W. Kim, Int J Thermophys (2009) 30, 1386–1395

L0085 – Thermal expansion of near stoichiometric (U,Er)O2 solid solutions

Thermal expansions of (U,Er)O2 solid solutions were investigated between room temperature and 1673 K by using a thermo-mechanical analyzer. Lattice parameters of the (U,Er)O2 pellets are lower than that of UO2 and they decrease as the Er contents increase. The linear thermal expansion and average thermal expansion coefficients of (U,Er)O2 are higher than that of UO2. For the temperature range from room temperature to 1673 K, the average thermal expansion coefficient values for UO2 and (U0.8Er0.2)O2 are 10.94x10^(-6) and 11.42x10^(-6) K-1, respectively.
S-H. Kim, H-S. Kim, Y-W. Lee, D-S. Sohn, D-S. Suhr, Materials Letters 60 (2006) 1480-1483

L0079 – The effects of the physical states of a simulated fission product on the linear thermal expansion of (U0.924Ce0.076)O2

The linear thermal expansions of an (U0.924Ce0.076)O2 pellet, doped a simulated fission product (Nd2O3 or Ru), were measured from room temperature to 1673K in a flowing argon atmosphere using TMA. Nd2O3 and Ru represent the physical states of a fission product, a dissolved oxide and a metallic precipitate, respectively. Using the measured data, the mean coefficients of a linear thermal expansion was obtained as a function of the temperature, and the effects of the physical states of a simulated fission product on the thermal expansion were investigated. In the case of the Nd2O3 forming a dissolved oxide, the thermal expansion of the sample increased and the increment was proportional to the Nd contents, because the melting point of the Nd2O3 was lower than that of UO2 and although the metallic precipitate hardly affected the crystal structure, the linear thermal expansion also increased with an increasing Ru contents.
D-J. Kim, Y-S. Kim, J-H. Yang, S-H. Kim, Y-W. Lee, H-S. Kim, Journal of Alloys and Compounds 421 (2006) 151-155

L0065 – Effects of solid fission products forming dissolved oxide (Nd) and metallic precipitate (Ru) on the thermal conductivity of uranium base oxide fuel

The effects of solid fission products on the thermal conductivity of uranium base oxide nuclear fuel were experimentally investigated. Neodymium (Nd) and ruthenium (Ru) were added to represent the physical states of solid fission products such as 'dissolved oxide' and 'metallic precipitate', respectively. Thermal conductivity was determined on the basis of the thermal diffusivity, density and specific heat values. The effects of the additives on the thermal conductivity were quantified in the form of the thermal resistivity equation - the reciprocal of the phonon conduction equation - which was determined from the measured data. It is concluded that the thermal conductivity of the irradiated nuclear fuel is affected by both the 'dissolved oxide' and the 'metallic precipitate', however, the effects are in the opposite direction and the 'dissolved oxide' influences the thermal conductivity more significantly than that of the 'metallic precipitate'.
D-J. Kim, J-H. Yang, J-H. Kim, Y-W. Rhee, K-W. Kang, K-S. Kim, K-W. Song, Thermochimica Acta 455 (2007) 123-128

L0047 – Variation of the lattice parameter and thermal expansion coefficient of (U,Dy)O2 as a function of DyO1.5 content

Thermal expansions of (U,Dy)O2 solid solutions were investigated between room temperature and 1673K by using a thermo-mechanical analyzer. The lattice parameter of (U,Dy)O2 pellets is lower than that of UO2 and it decreases as Dy content increases. The linear thermal expansion and average thermal expansion coefficients of (U,Dy)O2 are higher than that of UO2. For the temperature range from room temperature to 1673 K, the average thermal expansion coefficient values for UO2 and (U0.8Dy0.2)O2 are 10.97x10^(-6) and 11.37x10^(-6) K-1, respectively.
S-H. Kim, H-S. Kim, Y-W Lee, D-S. Sohn, D-S. Suhr, Journal of Alloys and Compounds 407 (2006) 263-267

L0025 – Experimental assessment of thermophysical properties of (Pu,Zr)N

The thermophysical properties of plutonium-zirconium nitride (0-25 at.% Pu), namely heat capacity, thermal conductivity and thermal expansion were measured on pellets produced thanks to the well established pellet pressing route. The experimental results obtained on ZrN were consistent with the existing literature data. The thermal properties measurements reported for (Pu0.25Zr0.75)N were in rather good agreement with predictive values calculated from ZrN and PuN available data.
V. Basini, J.P. Ottaviani, J.C. Richaud, M. Streit, F. Ingold, Journal of Nuclear Materials 344 (2005) 186-190

L0018 – The linear thermal expansion and the thermal diffusivity measurements for near-stoichiometric (U, Ce)O2 solid solutions

The thermal diffusivities of near-stoichiometric (U, Ce)O2 solid solutions containing CeO2 up to 22 mol% were investigated in the temperature range of 298-1273 K using the laser flash method. Also, linear thermal expansion measurements were performed in the temperature range of 298-1673 K using a thermomechanical analysis. The thermal conductivities were determined by a calculation of the thermal diffusivity, the density and the specific heat. The thermal conductivities of the tested samples could be expressed as a function of the temperature by the phonon conduction equation k = (A + BT)-1. The thermal conductivity decreased gradually with an increasing Ce content. This was attributable to the increasing lattice defect thermal resistance caused by the U4+, Ce4+ and O2- ions as phonon scattering centers.
D-J. Kim, Y-S. Kim, S-H. Kim, J-H. Kim, J-H. Yang, Y-W. Lee and H-S. Kim, Thermochimica Acta 441 (2006) 127-131

E0336 – Phase equilibria in the FeO1+x–UO2–ZrO2 system in the FeO1+x-enriched domain

Experimental results of the investigation of the FeO1+x–UO2–ZrO2 system in neutral atmosphere are presented. The ternary eutectic position and the composition of the phases crystallized at this point have been determined. The phase diagram is constructed for the FeO1+x-enriched region and the onset melting temperature of 1310 °C probably represents a local minimum and so will be a determining factor in this system and its application to safety studies in nuclear reactors.
V.I. Almjashev, M. Barrachin, S.V. Bechta, D. Bottomley, F. Defoort, M. Fischer, V.V. Gusarov, S. Hellmann, V.B. Khabensky, E.V. Krushinov, D.B. Lopukh, L.P. Mezentseva, A. Miassoedov, Yu.B. Petrov, S.A. Vitol, Journal of Nuclear Materials 400 (2010) 119–126

E0149 – Barium borosilicate glass – a potential matrix for immobilization of sulfate bearing high-level radioactive liquid waste

Borosilicate glass formulations adopted worldwide for immobilization of high-level radioactive liquid waste (HLW) is not suitable for sulphate bearing HLW, because of its low solubility in such glass. A suitable glass matrix based on barium borosilicate has been developed for immobilization of sulphate bearing HLW. Various compositions based on different glass formulations were made to examine compatibility with waste oxide with around 10 wt% sulfate content. The vitrified waste product obtained from barium borosilicate glass matrix was extensively evaluated for its characteristic properties like homogeneity, chemical durability, glass transition temperature, thermal conductivity, impact strength, etc. using appropriate techniques. Process parameters like melt viscosity and pour temperature were also determined. It is found that SB-44 glass composition (SiO2: 30.5 wt%, B2O3: 20.0 wt%, Na2O: 9.5 wt% and BaO: 19.0 wt%) can be safely loaded with 21 wt% waste oxide without any phase separation. The other product qualities of SB-44 waste glass are also found to be on a par with internationally adopted waste glass matrices. This formulation has been successfully implemented in plant scale.
C.P. Kaushik, R.K. Mishra, P. Sengupta, Amar Kumar, D. Das, G.B. Kale, Kanwar Raj, Journal of Nuclear Materials 358 (2006) 129-138

E0101 – Determination of specific surface area of uranium oxide powders using differential thermal analysis technique

The two step oxidation of UO(2+x) and reduction of U3Os powders observed during Differential Thermal Analysis (DTA) has been exploited to determine their Speeifie Surface Areas (SSAs). The results obtained by this method have been compared with the Braunauer, Emmett and Teller (BET) method and are found to be in good agreement in the SSA range of 2-4 m2/gm in the ease of UO(2+x) obtained from ADU route and 4-8 m2/gm in the ease of AUC route. A precision of ± 0.1 m2/gm is obtained. The maximum temperature of oxidation and reduction of these oxides are dependent upon their preparative routes such as Ammonium Diuranate (ADU) and Ammonium Uranyl Carbonate (AUC).
Y. Balaji Rao, R.B. Yadav, R. Narazana Swamy, B. Gopalan, S. Syamsundar, Journal of Thermal Analysis and Calorimetry 44 (1995) 1439-1448

B3360 – The system Yb-Ru-O: High temperature studies on the oxides Yb2Ru2O7(s) and Yb3RuO7(s)

The standard molar Gibbs free energy of formation of Yb2Ru2O7(s) and Yb3RuO7(s) was determined using an oxide solid-state electrochemical cell wherein calcia-stabilized-zirconia (CSZ) was used as an electrolyte. The standard molar Gibbs free energy of formation of Yb2Ru2O7(s) and Yb3RuO7(s) from elements in their standard state was calculated by the least squares regression analysis of the data obtained in the present study and can be given respectively by: {?fG°(Yb2Ru2O7, s)/(kJ mol?1) ± 1.88} = ?2450.4 + 0.6025·(T/K) and {?fG°(Yb3RuO7, s)/(kJ mol?1) ± 2.36} = ?3109.6 + 0.6202·(T/K). Standard molar heat capacity C°p,m(T) of Yb2Ru2O7(s), was measured using a heat flux type differential scanning calorimeter (DSC) in the temperature range from 307 K to 845 K. The heat capacity of Yb2Ru2O7(s) was used along with the data obtained from the oxide electrochemical cell to calculate the standard enthalpy of formation of the compound Yb2Ru2O7(s), from elements at 298.15 K.
Aparna Banerjee, Ziley Singh Chaudhary, Materials Chemistry and Physics 142 (2013) 12-16

B3359 – Solid oxide electrochemical cell and differential scanning calorimetry used for thermodynamic measurements of the ternary oxides: Nd2RuO5(s) and Nd2Ru2O7(s)

The standard molar Gibbs free energy of formation of Nd2RuO5(s) and Nd2Ru2O7(s) was determined using an oxide solid-state electrochemical cell wherein calcia stabilized zirconia (CSZ) was used as an electrolyte. The standard molar Gibbs free energy of formation of Nd2RuO5(s) and Nd2Ru2O7(s) from elements in their standard state was calculated by the least squares regression analysis of the data obtained in the present study and can be given respectively by: {?fGo(Nd2RuO5, s)/(kJ mol?1) ± 1.55} = ?2130.29 + 0.4786·(T/K); {?fGo(Nd2Ru2O7, s)/(kJ mol?1) ± 2.2} = ?2501.1 + 0.692·(T/K). Standard molar heat capacity Cop,m(T) of Nd2Ru2O7(s) was measured using a heat flux type differential scanning calorimeter (DSC) in two different temperature ranges, from 127 K to 299 K and 307 K to 745 K. The heat capacity of Nd2Ru2O7(s) was used along with the data obtained from the oxide electrochemical cell to calculate the standard enthalpy and entropy of formation of the compound from elements at 298.15 K.
Aparna Banerjee, Ziley Singh Chaudhary, Materials Chemistry and Physics 138 (2013) 417-422

B3355 – System Er–Ru–O: High temperature study of the heavy rare earth pyrochlore Er2Ru2O7(s) by electrochemical cell and differential scanning calorimeter

The Gibbs free energy of formation of Er2Ru2O7(s) has been determined using solid-state electrochemical technique employing oxide ion conducting electrolyte. The reversible electromotive force (e.m.f.) of the following solid-state electrochemical cell has been measured. Standard molar heat capacity C°p,m(T) of Er2Ru2O7(s) was measured using a heat flux type differential scanning calorimeter (DSC) in two different temperature ranges, from 129 K to 296 K and 307 K to 845 K. The heat capacity in the higher temperature range was fitted into a polynomial expression.
Aparna Banerjee, Solid State Ionics 253 (2013) 70–75

B3345 – Thermodynamic assessment of the (LiF + UF3) and (NaF + UF3) systems

In this article new thermodynamic descriptions of the (LiF + UF3) and the (NaF + UF3) systems based on new calorimetric data are presented. The experimental investigations were carried out using differential scanning calorimetry (DSC) to obtain phase transition temperatures between (0 and 60) mol% of UF3. On the basis of these measurements a thermodynamic assessment of the (LixU1-x)F3-2x and (NaxU1-x)F3-2x liquid solutions, using the quasi-chemical model was achieved. UF3 undergoes a disproportionation at higher temperatures to UF4 and U metal which had a significant impact on the transition temperatures in the (LiF + UF3) system. For this reason the influence of the presence of up to 15 mol% of UF4 in this system was calculated and compared to the experimental results. The final thermodynamic description was corrected for the pure binary system.
M. Beilmann, O. Beneš, R.J.M. Konings, Th. Fanghänel, J. Chem. Thermodynamics 57 (2013) 22–31

B3149 – Experimental study of uranium carbide pyrophoricity

Mixed plutonium and uranium monocarbide (UPuC) is considered as a possible fuel material for future nuclear gas fast reactors. Its safe handling is currently a major concern, because inflammation of this material under the shape of fine powders is easy and highly exothermic (pyrophoricity) even under ambient temperature and partial pressure of oxygen inferior to 0.2 bar. CEA Marcoule is implied in both experimental and numerical studies on the UC powder oxidation exothermic reaction. Experimental tests consist in determining the influence of various parameters (gas composition, heating ramp, specific surface of powders) on the sample inflammation temperature. Two kinds of analytical apparatus are used: The differential thermal analysis (DTA) and the differential scanning calorimetry (DSC) coupled to the thermo gravimetric analysis (TGA). These apparatus are also linked to a gas mass spectrometer to follow the composition of combustion chamber gases. Results obtained with small quantities revealed that UC powder is highly reactive in air in the temperature range of 150–250 °C and showed a strong dependence between powder height in crucibles and inflammation temperature.
C. Berthinier, S. Coullomb, C. Rado, E. Blanquet, R. Boichot, C. Chatillon, Powder Technology 208 (2011) 312–317

B3065 – Microstructural and thermophysical properties of U–6 wt.%Zr alloy for fast reactor application

The microstructural and high temperature behavior of U–6 wt.%Zr alloy has been investigated in this study. U–6 wt.%Zr alloy sample for this study was prepared by following injection casting route. The thermophysical properties like coefficient of thermal expansion, specific heat, thermal conductivity of the above alloy were determined. The hot-hardness data of the U–6 wt.%Zr alloy was also generated from room temperature to 973 K. Apart from that, the fuel-clad chemical compatibility with T91 grade steel was also studied by diffusion couple experiment. No studies have been reported on U–6 wt.%Zr alloy. This paper aims at filling up the gap on characterization and thermophysical property evaluation of U–6 wt.%Zr alloy.
Santu Kaity, Joydipta Banerjee, M.R. Nair, K. Ravi, Smruti Dash, T.R.G. Kutty, Arun Kumar, R.P. Singh, Journal of Nuclear Materials 427 (2012) 1–11

B3046 – Solubility and partitioning of minor-actinides and lanthanides in alumino-borosilicate nuclear glass

We prepared a heterogeneous SiO2-B2O3-Na2O-Al2O3-CaO-La2O3 glass containing 11.76 wt.% AmO2. The composition of this heterogeneous material was determined from that of non-active glasses in which Am was fully substituted by Nd. For trivalent ion concentrations above the solubility limit, crystal formation occurred in the form of apatite-like silicates, this in both active and non-active glasses. Additions of Am+3 and Nd3+ led to depolymerization of the silicate network which is consistent with the modifier role expected from these cations. Thus, we can compare the incorporation mechanisms of neodymium and americium in the glassy network and within the apatite-like crystalline phases
Abdessamad Kidari, Magali Magnin, Richard Caraballo, Magaly Tribet, Franck Doreau, Sylvain Peuget, Jean-Luc Dussossoy, Isabelle Bardez-Giboirea, Christophe Jégou, Procedia Chemistry 7 ( 2012 ) 554 – 558

B3019 – Etude thermodynamique de la sorption de l’uranyle sur la monazite et la magnétite

mon travail concerne l’étude des phénomènes de sorption intervenant entre un ion radiotoxique « modèle » en solution aqueuse, l’ion uranyle UO2 2+ et deux substrats minéraux : la monazite, composé méthodologique, système relativement simple sur lequel une étude structurale à température ambiante a déjà été réalisée et la magnétite, dépôt poreux issu de la corrosion des composants en acier au carbone du poste d'eau du circuit secondaire des REP et produit de corrosion des containers métalliques destinés au stockage des déchets radioactifs en formation géologique profonde
Olivia Felix, Thèse Institut de Physique Nucléaire d'Orsay, Juillet 2012

B2973 – High temperature thermal properties of UFe2. Measurement by drop calorimetry and modelling using quasi-harmonic Debye–Grüneisen formalism

High temperature drop calorimetry measurements of enthalpy increment have been made on phase pure UFe2 compound for temperatures up to 1,473 K. The measured data agreed well with the earlier reported data that are available only up to 1,000 K. An integrated modelling of the present heat capacity data has been made using the framework of quasi-harmonic Debye-Grüneisen formalism, the input parameters for which have been obtained after a critical appraisal of the available thermophysical data. The modelling yielded fully consistent estimates of both heat capacity and volume thermal expansivity, for the complete temperature range, 0–1,273 K.
Arun Kumar Rai, Subramanian Raju, J Therm Anal Calorim, 2012

B2951 – Thermodynamic investigation of the LiF–ThF4 system

A thermodynamic investigation of the LiF–ThF4 system is presented in this study. The enthalpy of mixing of the (Li,Th)Fx liquid solution was measured for the first time using a method designed for conventional DSC technique. To verify the possibility to measure the mixing enthalpy with the used calorimeter, the known LiF–KF system was first investigated and compared to the literature data. After the successful test, this technique was applied to investigate the LiF–ThF4 system and obtained novel results are presented in this study. Furthermore, using the DSC technique, new equilibrium data of the LiF–ThF4 phase diagram were measured, confirming the stability of the LiThF5 phase. The intermediate compound Li3ThF7 was synthesized and its enthalpy of fusion was determined. Considering the new experimental data, the LiF–ThF4 system was re-optimized using a quasi chemical model for the description of excess Gibbs parameters of the liquid solution.
E. Capelli, O. Beneš, M. Beilmann, R.J.M. Konings, J. Chem. Thermodynamics 58 (2013) 110–116

B2914 – Preparation of poly(methyl acrylate)/phosphate/composites and its possible use as storage medium for radioactive isotopes

The distribution of 137Cs, 152Eu, 238U, and 85Sr in a solid/aqueous system (poly(methyl acrylate)/phosphate/composite in contact with groundwater, was investigated using c-Spectrometry and flourometry. The results were compared with earlier results with mineral phosphate in the solid phase. The effect of contact time, pH and the concentration of concurrent element were studied. The ability of the prepared polymer composites to keep the studied radioisotopes in the solid phase is much higher than mineral phosphate. The used polymer composites have been prepared consisting of natural phosphate powder and the monomer methyl acrylate using gamma irradiation. The yield of polymerization was followed upwith respect to the irradiation dose using thermogravimetric analyzer (TGA). A thermomechanical analyzer (TMA) was used to locate the area of the glass transition temperatures (Tg) using the mode with alternative variable force; the mode with constant force was used to determine the Tg of the pure polymer and the polymer composite prepared at the same irradiation dose. The Tg of the pure poly(methyl acrylate) is 13 ± 3 °C, and the Tg of poly(methyl acrylate)/phosphate/composites is 8 ± 3 °C. The Tg were also determined using the DSC technique, and similar values were found.
Z. Ajji, O. Alhassanieh, J Radioanal Nucl Chem (2011) 287, 69–75

B2896 – High-temperature X-ray diffraction and specific heat studies on GdAlO3, Gd3Al5O12 and Gd4Al2O9

Gadolinium aluminates, GdAlO3, Gd3Al5O12 and Gd4Al2O9 were synthesized by the solution combustion method. Very fine particles in the nanoparticle range of 10–20nm could be prepared by this method as evidenced by surface area measurement by multipoint BET method. Thermal studies on these compounds were carried out using high-temperature X-ray diffraction (HT-XRD) and differential scanning calorimetry (DSC) methods. The thermal expansion coefficients of GdAlO3, Gd3Al5O12 and Gd4Al2O9 were calculated from the lattice parameter data and specific heats were calculated from DSC data. The lattice parameters of GdAlO3 and Gd3Al5O12 were found to increase linearly with temperature whereas Gd4Al2O9 did not show a linear trend. The specific heats of these compounds show an increasing trend with increase in aluminum atom fraction. Based on the thermodynamic data available in the literature and the specific heat data obtained in this study, oxygen potential diagram was constructed at 1000 K.
Satyajeet Chaudhury, S.C. Parida, K.T. Pillai, K.D. Singh Mudher, Journal of Solid State Chemistry 180 (2007) 2393–2399

B2723 – Thermophysical characterization of ZrN and (Zr,Pu)N

Nitride compounds are currently considered as possible fuels for advanced nuclear reactors. Among the relevant properties to be investigated, thermophysical quantities such as heat capacity and thermal diffusivity/conductivity are key to ensure safe operation of the fuel in the reactor. Both physical and chemical phenomenological experimental activities are included in this effort. The experimental procedures were optimizedto limit oxidation ofmaterials in presence of even small quantities of oxygen and moisture, especially during the measurements at high temperature. Heat capacity and thermal diffusivity/conductivity were measured for ZrN and (Zr,Pu)N compounds. A consistent set of results was obtained from low- and high-temperature Cp measurements for both materials. The excess Cp of solution was estimated. Increasing trends with temperature were observed for the thermal diffusivity and conductivity of both materials in the temperature range considered, confirming the metallic behavior of this class of materials. The presence of Pu causes an almost constant decrease of the thermal conductivity compared to the pure ZrN matrix.
A. Ciriello, V.V. Rondinella?, D. Staicu, J. Somers, O. Benes, R. Jardin ,D. Bouëxière, F.Wastin, E. Coliena, Journal of Alloys and Compounds 473 (2009)265-271

B2722 – Kinetic studies of the ?–? phase transition in the Zr1%Nb cladding for nuclear reactors

The a–b phase transition of a zirconium alloy doped with 1 mol% of niobium (E110 alloy) is investigated using differential scanning calorimetry. The onset and endpoints of the transition are identified from the measured heat flow signal and from the integration of the observed peak the extent of the a–b phase change is calculated as a function of temperature. The experiment has been performed at different heating rates and a shift of the onset temperature with increasing heating rate was observed. From the dataset the equilibrium transition curve has been extrapolated and compared with other types of zirconiumbased cladding materials.
O. Beneš, P. Van Uffelen, J. van de Laar, Cs. Gy}ori, R.J.M. Konings, Z. Hózer, Journal of Nuclear Materials 414 (2011) 88–91

B2721 – High temperature heat capacity of PuPO4 monazite-analogue

The enthalpy increments of PuPO4 have been measured using drop calorimetry in the temperature range from 530 K to 1386 K. The heat capacity was derived from the obtained data and compared with heat capacity data obtained directly from differential scanning calorimeter measured in this study from 400 K to 1400 K. The recommended heat capacity of PuPO4 was determined based on both techniques
Ondrej Beneš, Karin Popa, Vivien Reuscher, Alessandro Zappia, Dragos Staicu, Rudy J.M. Konings, Journal of Nuclear Materials 418 (2011) 182–185

B2713 – A comprehensive study of the heat capacity of CsF from T = 5 K to T = 1400 K

In this study, we present new experimental heat capacity data of solid and liquid phase of CsF covering the temperature range from 5 K to 1400 K. The low temperature data were obtained by adiabatic calorimetry and compared with the theoretical heat capacity computed from ab initio harmonic crystal approximation. The absolute entropy at T = 298.15 K determined based on the presented results is 93:60 J Kÿ1 molÿ1. The high temperature heat capacity was measured using a drop calorimeter and the data for the solid phase revealed significant excess properties. The data for the liquid phase were correlated with the molecular dynamic simulation obtaining a good agreement
O. Beneš, R.J.M. Konings, D. Sedmidubsky, M. Beilmann, O.S. Valu, E. Capelli, M. Salanne, S. Nichenko

B2712 – Density functional theory, molecular dynamics, and differential scanning calorimetry study of the RbF–CsF phase diagram

A multiscale modeling approach is developed to compute the phase diagram of the RbF–CsF binary system. The mixing enthalpies of the Rb,Cs F solid and liquid solutions are evaluated using density functional theory and classical molecular dynamics calculations, respectively. For the solid solution, 18 different configurations are studied with density functional theory and the surrounded atom model is applied in order to compute the configurational partition function. We also measure the solidus and liquidus equilibria using differential scanning calorimetry. Finally the RbF–CsF phase diagram is constructed using the calculated excess free enthalpies of the solid and liquid solutions and a very good agreement with our experimental data is found
O. Beneš, Ph. Zeller, M. Salanne, R. J. M. Konings, J. Chem. Phys. 130, 134716 (2009)

B2703 – Measurement and modelling of high temperature thermodynamic properties of URh3 alloy

The high temperature phase stability of arc-melted cubic URh3 intermetallic compound has been investigated using high temperature inverse drop calorimetry in the temperature range of 300–1273 K. URh3 exists as a line compound with negligible solubility range. Room temperature XRD profile and elemental X-ray mapping experiments on 1273 K/3 h homogenized samples have confirmed the homogeneity and L12 (cF8; pm3m) crystal structure of URh3. The drop measurements yielded accurate values for the enthalpy increment ?H0 T as a function of temperature, from which the specific heat CP has been estimated. The enthalpy data obtained in this study have been compared and combined with the reported data on low temperature CP and also with the ?H0 T in the temperature range, 0–840 K, for a comprehensive theoretical analysis using quasiharmonic Debye–Grüneisen formalism. It is found that this model with due allowance for thermal expansion effects can successfully account for the experimentally measured thermal property data in the entire temperature region spanning 0–1273 K. Invoking a combination of measurement and modelling, a comprehensive set of thermodynamic quantities have been obtained for URh3.
Arun Kumar Rai, Haraprasanna Tripathy, B. Jeya Ganesh, S. Raju, Journal of Nuclear Materials 427 (2012) 378–383

B2702 – Thermodynamic investigations of ThO2–UO2 solid solutions

Heat capacities and enthalpy increments of solid solutions Th1 yUyO2(s) (y = 0.0196, 0.0392, 0.0588, 0.098, 0.1964) and Simfuel (y = 0.0196) were measured by using a differential scanning calorimeter and a high temperature drop calorimeter. The heat capacities were measured in two temperature ranges: 127–305 K and 305–845 K and enthalpy increments were determined in the temperature range 891–1698 K. A heat capacity expression as a function of uranium content y and temperature and a set of self-consistent thermodynamic functions for Th1 yUyO2(s) were computed from present work and the literature data. The oxygen potentials of Th1 yUyO2+x(s) have been calculated and expressed as a polynomial functions of uranium content y, excess oxygen x and temperature T. The phase diagram, oxygen potential diagram of thorium–uranium–oxygen system and major vapour species over urania thoria mixed oxide have been computed using FactSage code
Smruti Dash, S.C. Parida, Ziley Singh, B.K. Sen, V. Venugopal, Journal of Nuclear Materials 393 (2009) 267–281

B2683 – Calorimetric measurements on (U,Th)O2 solid solutions

Enthalpy increments of urania – thoria solid solutions, (U0.10Th0.90)O2, (U0.50Th0.50)O2 and (U0.90Th0.10)O2 were measured by drop calorimetry in the temperature range 479 – 1805 K. Heat capacity, entropy and Gibbs energy function were computed. The heat capacity measurements were carried out also with differential scanning calorimetry in the temperature range 298 – 800 K. The heat capacity values of (U0.10Th0.90)O2, (U0.50Th0.50)O2 and (U0.90Th0.10)O2 at 298 K are 59.62, 61.02, 63.56 J K 1 mol 1, respectively. The results were compared with the data available in the literature. From the study, the heat capacity of (U,Th)O2 solid solutions was shown to obey the Neumann – Kopp’s rule.
R. Kandan, R. Babu, P. Manikandan, R. Venkata Krishnan, K. Nagarajan, Journal of Nuclear Materials 384 (2009) 231–235

B2682 – Heat capacity measurements and XPS studies on uranium–lanthanum mixed oxides

Heat capacity measurements were carried out on (U1?yLay)O2±x (y = 0.2, 0.4, 0.6, and 0.8) using differential scanning calorimeter (DSC) in the temperature range 298–800 K. Enthalpy increment measurements were carried out on the above solid solutions using high temperature drop calorimetry in the temperature range 800–1800 K. Chemical states of U and La in the solid solutions of mixed oxides were determined using X-ray photoelectron spectroscopy (XPS). Oxygen to metal ratios of (U1?yLay)O2±x were estimated from the ratios of different chemical states of U present in the sample. Anomalous increase in the heat capacity is observed for (U1?yLay)O2±x (y = 0.4, 0.6 and 0.8) with onset temperatures in the range of 1000–1200 K. The anomalous increase in the heat capacity is attributed to certain thermal excitation process, namely, Frenkel pair defect of oxygen. The heat capacity value of (U1?yLay)O2±x (y = 0.2, 0.4, 0.6, and 0.8) at 298K are 65.3, 64.1, 57.7, 51.9 JK?1 mol?1, respectively. From the XPS investigations, it was observed that the O/M ratios at the surface are higher than that in the bulk. In uranium rich mixed oxide samples, the surface O/M ratios are greater than 2 whereas that in La rich mixed oxides, they are less than 2, though the bulk O/M in all the samples are less than 2.
R. Venkata Krishnan, V.K. Mittal, R. Babu, Abhiram Senapati, Santanu Bera, K. Nagarajan, Journal of Alloys and Compounds 509 (2011) 3229–3237

B2681 – Specific heats of thoria–urania solid solutions

Thoria–urania solid solutions having compositions ThO2–4%UO2, ThO2–10%UO2, ThO2–20%UO2, ThO2–50% UO2 and ThO2–80%UO2 (all compositions are in wt%) were prepared by coated agglomerate pelletization (CAP) process and powder metallurgy (P/M) routes, characterized by ICP-AES, density, grain size, oxygen to metal (O/M) ratio, X-ray diffraction (XRD), lattice parameter and scanning electron microscope (SEM). Specific heats of pure ThO2, UO2 and these solid solutions were measured by differential scanning calorimeter in the temperature range from 300 to 1650 K. The results obtained in this study were compared with those available in the literature and on this basis it was found out that the molar specific heats of these substances obtained in the present study are within an accuracy limit of ±4%.
Joydipta Banerjee, S.C. Parida, T.R.G. Kutty, Arun Kumar, Srikumar Banerjee, Journal of Nuclear Materials 427 (2012) 69–78

B2680 – Thermophysical properties of Dy6UO12

Dysprosium uranate (Dy6UO12) was prepared by citrate gel combustion and characterized by X-ray diffraction (XRD). A single phase with a rhombohedral structure was observed (R3). The room temperature values of the lattice parameters ‘‘a’’ and ‘‘c’’ were found to be 9.9747 and 9.4370 A ° , respectively. Heat capacity and enthalpy increment measurements were carried out by using a differential scanning calorimeter (DSC) and inverse drop calorimetry in the temperature range 298–800 K and 600–1700 K respectively. The thermal expansion characteristics were measured by using High Temperature XRD (HTXRD) in the temperature range 298–1973 K. The heat capacity data of Dy6UO12 is being reported for the first time and its value at 298 K is 462 J K-1 mol-1. The coefficient of lattice thermal expansion in the temperature range 298–1973 K along ‘‘a’’ and ‘‘c’’ axes are 13.28x10-6 and 9.35x10-6 K-1 respectively. These data are being reported for the first time as well.
R. Venkata Krishnan, R. Babu, G. Panneerselvam, K. Ananthasivan, M.P. Antony, K. Nagarajan, Ceramics International 38 (2012) 5277–5280

B2679 – Calorimetric investigations on cubic BaTiO3 and Ba0.9Nd0.1TiO3 systems

Drop calorimetric studies were carried out on cubic BaTiO3 and Ba0.9Nd0.1TiO3. Enthalpy increments were measured by inverse drop calorimetric method in the temperature range 573–1523K using a multidetector high temperature calorimeter. Thermodynamic functions such as heat capacity, entropy, Gibbs energy functions in the temperature range 298–1600K were computed from the measured enthalpy increments.
R. Babu, R. Kandan, Hrudananda Jena, K.V. Govindan Kutty, K. Nagarajan, Journal of Alloys and Compounds 506 (2010) 565–568

B2640 – Specific heats of ternary oxides in the Li–U(VI)–O System

Seven ternary oxides; Li4UO5, Li2UO4, Li22U18O65, Li2U1.75O6.25, Li2U2O7, Li2U3O10 and Li2U6O19 in the system Li–U(VI)–O were prepared by solid-state reaction route and characterized by X-ray diffraction method. Specific heats of these compounds were measured by differential scanning calorimetry in the temperature range from 300 to 860 K. The specific heats show a decreasing trend with increase in UO3(s) content in these lithium uranates. However, the specific heat per gram atom shows an increasing trend with decrease in number of oxygen atoms in the formula unit.
S.K. Rakshit, Ram Avtar Jat, Y.P. Naik, S.C. Parida, Ziley Singh, B.K. Sen, Thermochimica Acta, 490 (2009) 60–63

B2629 – Calorimetric measurements on plutonium rich (U,Pu)O2 solid solutions

Enthalpy increments of U(1?y)PuyO2 solid solutions with y = 0.45, 0.55 and 0.65 were measured using a high-temperature differential calorimeter by employing the method of inverse drop calorimetry in the temperature range 956–1803 K. From the fit equations for the enthalpy increments, other thermodynamic functions such as heat capacity, entropy and Gibbs energy function have been computed in the temperature range 298–1800 K. The results are presented and compared with the data available in the literature. The results indicate that the enthalpies of U(1?y)PuyO2 solid solutions with y = 0.45, 0.55 and 0.65 obey the Neumann–Kopp's molar additivity rule.
R. Kandan, R. Babu, K. Nagarajan, P.R. Vasudeva Rao, Thermochimica Acta, 472 (2008) 46-49

B2628 – Thermodynamic functions of Ba10(PO4)6Cl2, Sr10(PO4)6Cl2 and Ca10(PO4)6Cl2

Alkaline earth chloroapatites, M10(PO4)6Cl2 (M = Ba, Sr, Ca) were prepared by solid state reaction route and characterized by XRD. The lattice constants of Ba10(PO4)6Cl2, Sr10(PO4)6Cl2, Ca10(PO4)6Cl2 were determined to be a = 10.27(1) Å, c = 7.66(1) Å; a = 9.87(2) Å, c = 7.18(1) Å and a = 9.52(1) Å, c = 6.85(1) Å respectively. The enthalpy increments were measured by inverse drop calorimetric method in the temperature range 523–1423 K using a high temperature calorimeter. Thermodynamic functions such as heat capacity, entropy and Gibbs energy functions in the temperature range 298–1500 K, were computed from the measured enthalpy increments.
R. Babu, Hrudananda Jena, K.V. Govindan Kutty, K. Nagarajan, Thermochimica Acta, 526 (2011) 78–82

B2443 – Thermodynamic properties of uranium plutonium mixed carbides

High plutonium containing hyperstoichiometric mixed carbide fuels are used as driver fuel in the Fast Breeder Test Reactor at Kalpakkam, India. The enthalpy increment of the mixed carbide fuels were measured in the temperature range 1000–1800 K using a high temperature drop calorimeter. The results are presented and compared with the data available in the literature.
R. Kandan, R. Babu, K. Nagarajan, J Therm Anal Calorim (2010) 99:713–717

B2374 – Test of a Large Volume Calorimeter in KEPRI Tritium Laboratory

A tritium assaying and dispensing system has been designed and is being constructed in KEPTL (Korea Electric Power Research Institute, Tritium Laboratory). This facility will be used to supply tritium from WTRF (Wolsong Tritium Removal Facility) to customers for industrial uses, R&D application, ITER operation and other purposes. In order to account the tritium amount loaded in and out this facility, KEPTL purchased a large volume calorimeter from Setaram. This calorimeter is integrated in a dedicated room of the KEPTL in 2008 and the site acceptance tests are completed. The KEPTL calorimeter is a twin cell type and its measuring range is from 0.01 gram of tritium to 60 gram of tritium. The tritium storage beds loaded into the measuring cell of the calorimeter reach a maximum size of 200 mm in diameter and 570 mm in height. The performance test has been successfully completed with a Joule effect cell simulating tritium decay heat. It is demonstrated that the precision of the calorimeter is measured to be less than 0.28% at 50 mW and 0.13% at 10W. Relative error on real tritium test is measured to be 0.06% on 37 TBq of tritium.
K. M. Song, B. W. Ko, K. W. Lee, S. H. Sohn

B2320 – Standard enthalpy of formation and heat capacity of compounds in the pseudo-binary Bi2O3–Fe2O3 system

Bismuth ferrites, Bi2Fe4O9, BiFeO3, and Bi25FeO39 have been prepared and characterized. The thermodynamic data such as standard molar enthalpies of formation ?fH°m (298.15 K) and standard molar heat capacity (Cp°) were obtained from the solution calorimetry and calvet calorimetric measurements.?fH°m (298.15 K) for Bi2Fe4O9(s), BiFeO3(s), and Bi25FeO39(s) were found to be (-2476.0 ± 4.3), (-768.4 ± 2.9) and (-7699.8 ± 17.3) kJ molÿ1, respectively. The dependence of molar heat capacity Cp with temperature can be given as C°pBi2Fe4O9, s) (J K-1 mol-1) = 353.81 + 0.01774T - 1.63606 x 105/T2 (313 < T/K < 911) CpðBiFeO3, s) (J K-1 mol-1) = 116.72 + 0.00968T - 1.9273 x 105/T2 (313 < T/K < 911) and CpðBi25FeO39, s) (J K-1 mol-1) = 1339.21 + 0.16550T - 28.3546 x 105/T2 (324 < T/K < 834) The standard Gibbs energy of formation of the above compounds have been derived using the values of ?fH°m (298.15 K), Cp and the estimated ?fS°m (298.15 K).
S. Phapale, R. Mishra, D. Das, Journal of Nuclear Materials 373 (2008) 137–141

B2294 – Large Volume Calorimetry : A Technique to Quantify Tritium

Characterization of nuclear samples can be achieved in terms of impurities by classical chemical analytical techniques, such as emission, absorption or mass spectrometries, but the determination of nuclear element inventory implies the use of specific techniques. Determination of plutonium can be carried out by coulometry, isotopic dilution mass spectrometry, UVvisible, alpha or gamma spectrometry. Techniques that are available to quantify tritium are for instance liquid or solid scintillation, mass spectrometry, helium-3 method or calorimetry. All these techniques differentiate by accuracy, limit of detection, analysis time, dependence on matrix but also by the fact that some are destructive when others are non destructive. Among non destructive and non intrusive methods, calorimetry is considered to be the most accurate, even for large volume samples. On account of these characteristics, determination has been achieved for more than 35 years by calorimetry at CEA Valduc. However, need to quantify lower amounts of tritium leaded CEA to invest in new apparatus. CEA entered into collaboration with Setaram Instrumentation in order to develop and manufacture high sensitive and accurate large volume calorimeter.
Jean-Charles Hubinois, Alain Godot, Sonia de Boyer, Guénaëlle Coindet, Armelle Collardey poster at 8th International Conference on Tritium Science and Technology September 16-21, 2007 Rochester, New York

B2283 – Calorimetric measurements on uranium-plutonium mixed oxides

Enthalpy increments of U(1-y)PuyO2 solid solutions with y = 0.21, 0.28 and 0.40 were measured using a high-temperature differential calorimeter by employing the method of inverse drop calorimetry in the temperature range 1000- 1780 K. From the fit equations for the enthalpy increments, other thermodynamic functions such as heat capacity, entropy and Gibbs energy function have been computed in the temperature range 298-1800 K. The results indicate that the enthalpies of (U,Pu)O2 solid solutions obey the Neumann-Kopp molar additivity rule.
R. Kandan, R. Babu, K. Nagarajan, P.R. Vasudeva Rao, Journal of Nuclear Materials 324 (2004) 215-219

B1690 – A Chinese view on summary of condensed matter nuclear science

Investigation on tritium was one of the recommendations in ERAB report of DOE in November, 1989. 15 year evolution of the related research proved that it was an important recommendation. A selective resonant tunneling model is attempted to explain this discovery. Deuterium flux might be a key issue to solve the problem of the reproducibility. A further investigation is suggested based on this model.
X.Z. Li, B. Liu, Q.M. Wei, S.X. Zheng, D.X. Cao, J.Fusion Energy 23 (2004) 217-221

B0572 – Thermodynamic properties of compounds of alkaline earth elements with other fission products

The alkaline earth fission products barium and strontium can combine with other major fission product elements such as Mo, Ce and Zr in a mixed oxide fuel pin of a fast breeder reactor to form compounds such as molybdates. cerates and zirconates. In order to understand the condition of their formation in the fuel pin basic thermodynamic data on these compounds applicable to the relevant temperature range are required. In this work enthalpy increments of BaMoO4, BaCeO3 and (Ba, Sr)ZrO3 were determined relative to room temperature using a high temperature differential calorimeter. The experimentally measured enthalpy data covered the temperature range of 985-1750 K. From these enthalpy values other thermodynamic functions such as heat capacities, entropies and free energy functions were generated using Cp2980 and S2980 values of the compounds available in the literature. The free energies of formation of these compounds and the equilibrium barium partial pressures for the formation reactions have been computed in the temperature range of 500-2000 K and in the oxygen potential range of -376.56 kJ mol-1 to -502.08 kJ mol-1.
Rita Saha, R. Babu, K. Nagarajan and C.K. Mathews, Thermochimica Acta 120 (1987) 29-39

B0571 – Thermodynamic functions of barium and strontium zirconates from calorimetric measurements

Enthalpy increment measurements have been carried out on BaZrO3 and SrZrO3 in the temperature range 1030-1700 K using a high-temperature differential calorimeter. From the measured enthalpy increment values, other thermodynamic functions, such as heat capacity, entropy and free energy functions, of these compounds have been calculated. The free energies of formation of BaZrO3 and the equilibrium barium partial pressures for the formation reactions have been computed and compared with literature data. The equilibrium strontium pressures for the formation reactions of SrZrO3 have also been computed.
K. Nagarajan, Rita Saha, R. Babu and C.K. Mathews, Thermochimica Acta 90 (1985) 297-304

A2280 – Candidate waste forms for immobilisation of waste chloride salt from pyroprocessing of spent nuclear fuel

Sodalite/glass bodies prepared by hot isostatic pressing (HIPing) at ?850 °C/100 MPa are candidates for immobilising fission product-bearing waste KCl–LiCl pyroprocessing salts. To study the capacity of sodalite to structurally incorporate such pyroprocessing salts, K, Li, Cs, Sr, Ba and La were individually targeted for substitution in a Na site in sodalite (Na vacancies targeted as charge compensators for alkaline and rare earths) and studied by X-ray diffraction and scanning electron microscopy after sintering in the range of 800–1000 °C. K and Li appeared to enter the sodalite, but Cs, Sr and Ba formed aluminosilicate phases and La formed an oxyapatite phase. However these non-sodalite phases have reasonable resistance to water leaching. Pure chlorapatite gives superior leach resistance to sodalite, and alkalis, alkaline and rare earth ions are generally known to enter chlorapatite, but attempts to incorporate simulated waste salt formulations into HIPed chlorapatite-based preparations or to substitute Cs alone into the structure of Ca-based chlorapatite were not successful on the basis of scanning electron microscopy. The materials exhibited severe water leachability, mainly in regard to Cs release. Attempts to substitute Cs into Ba- and Sr-based chlorapatites also did not look encouraging. Consequently the use of apatite alone to retain fission product-bearing waste pyroprocessing salts from electrolytic nuclear fuel reprocessing is problematical, but chlorapatite glass–ceramics may be feasible, albeit with reduced waste loadings. Spodiosite, Ca2(PO4)Cl, does not appear to be suitable for incorporation of Cl-bearing waste containing fission products.
E.R. Vance, J. Davis, K. Olufson, I. Chironi, I. Karatchevtseva, I. Farnan, Journal of Nuclear Materials 420 (2012) 396–404

A2060 – Development of Impregnated Agglomerate Pelletization (IAP) process for fabrication of (Th,U)O2 mixed oxide pellets

Impregnated Agglomerate Pelletization (IAP) technique has been developed at Advanced Fuel Fabrication Facility (AFFF), BARC, Tarapur, for manufacturing (Th,233U)O2 mixed oxide fuel pellets, which are remotely fabricated in hot cell or shielded glove box facilities to reduce man-rem problem associated with 232U daughter radionuclides. This technique is being investigated to fabricate the fuel for Indian Advanced Heavy Water Reactor (AHWR). In the IAP process, ThO2 is converted to free flowing spheroids by powder extrusion route in an unshielded facility which are then coated with uranyl nitrate solution in a shielded facility. The dried coated agglomerate is finally compacted and then sintered in oxidizing/reducing atmosphere to obtain high density (Th,U)O2 pellets. In this study, fabrication of (Th,U)O2 mixed oxide pellets containing 3–5 wt.% UO2 was carried out by IAP process. The pellets obtained were characterized using optical microscopy, XRD and alpha autoradiography. The results obtained were compared with the results for the pellets fabricated by other routes such as Coated Agglomerate Pelletization (CAP) and Powder Oxide Pelletization (POP) route.
P.M. Khot, Y.G. Nehete, A.K. Fulzele, Chetan Baghra, A.K. Mishra, Mohd. Afzal, J.P. Panakkal, H.S. Kamath, Journal of Nuclear Materials 420 (2012) 1–8

A1860 – Gibbs energy of formation of solid Ni3TeO6 from transpiration studies

The thermodynamic stability of nickel tellurate was established by studying the heterogeneous equilibrium Ni3TeO6(s)= 3NiO(s) + TeO2(g) + 1/2O2(g). Using the transpiration technique, the equilibrium constant of the reaction was obtained from the vapor pressure measurement of TeO2(g) over the biphasic mixture of Ni3TeO6(s) and NiO(s) under 1 bar oxygen pressure. From the equilibrium data the second-law value of ?fH° (Ni3TeO6(s), 298.15 K) was found to be -1216(±14) kJ molÿ1. The energy change of the heterogeneous reaction derived from the equilibrium constant was used to calculate the Gibbs energy of formation of Ni3TeO6 as could be expressed by ?fG° (Ni3TeO6(s)) (±13 kJmol) = -1307 + 0:64.T, (1122 K
M. Ali (Basu), R. Mishra, A.S. Kerkar, S.R. Bharadwaj, D. Das, Journal of Nuclear Materials 301 (2002) 183–186

A1811 – Melting and vaporization of salts in a U-LiCl-Li2O system

The electrochemical reduction of uranium oxide in the treatment of spent nuclear fuel requires a characterization of the LiCl–Li2O salt used as a reaction medium. Physical properties, melting and vaporization are important for the application of the salt and thus they have been investigated by differential scanning calorimetry (DSC) and thermogravimetry (TG), respectively. Experimental data suggest LiCl and Li2O compound formations, leading to a melting point depression of the LiCl and a co-vaporization of the LiCl–Li2O salt.
J. M. Hur, S. B. Park, C. S. Seo, K. J. Jung, S. W. Park, Journal of Radioanalytical and Nuclear Chemistry, Vol. 270, No.3 (2006) 489–493

A1810 – Characteristics of lagoon sludge waste generated from an uranium conversion plant

The Korea Atomic Energy Research Institute (KAERI) has launched a decommissioning program of the uranium conversion plant. The sludge waste, which was generated during the operation of the plant and stored in the lagoon, was characterized for the development of the treatment process. The physical properties were measured and chemical compositions and radiological properties analyzed. The main compounds of the sludge were ammonium nitrate, sodium nitrate, calcium nitrate, and calcium carbonate. All heavy radioactive elements such as uranium, thorium and 226Ra were precipitated and deposited at the bottom, and were not dissolved in the concentrated nitrate solution. A possible flow-scheme for processing is presented.
D. S. Hwang, K. I. Lee, Y. D. Choi, S. T. Hwang, J. H. Park, Journal of Radioanalytical and Nuclear Chemistry, Vol. 260, No. 2 (2004) 327–333

A1809 – Separation of 99Mo from a simulated fission product solution by precipitation with ?-benzoinoxime

The formation property of Mo precipitate was investigated and the existing process was improved using H2O2 that acts as an interfering compound in a subsequent alumina adsorption process. The property of the Mo precipitate was investigated by using SEM, FTIR, TG-DTA, and XRD. The simulated solution consisted of 1M nitric acid containing seven elements (Mo, I, Ru, Zr, Ce, Nd, Sr) and their radioactive tracers. As a result, the precipitate was composed of the Mo precipitate and re-precipitated ?-benzoinoxime which was added excessively for increasing the precipitation efficiency. It was confirmed that the Mo precipitate was formed by the reaction of two ?-benzoinoxime molecules and one MoO2 2+. Molybdenum precipitate was dissolved in 0.4M NaOH solution within 5 minutes without H2O2. Hydrogen peroxide induced only the rapid dissolution of the ?-benzoinoxime re-precipitate. Also, the dissolution method without H2O2 was favorable in the purification aspect because Zr and Ru were contained as a small fraction of 1.3% and 7.7%, respectively, in the dissolving solution.
D. S. Hwang, W. M. Choung, Y. K. Kim, J. H. Park, S. J. Park, Journal of Radioanalytical and Nuclear Chemistry, Vol. 254, No. 2 (2002) 255–262

A1776 – Preparation of thorium oxalate–silica sorbent and its application for the sorption of americium from aqueous solutions

A new sorbent, thorium oxalate incorporated in silica gel matrix was prepared. This material was characterized by X-ray, Thermo-gravimetric Analysis, surface area and porosity analysis. The material was obtained in the form of granular particles in the mesh size range of 80–150 American Standard of Testing Materials, yielding good liquid flow, when packed in ion exchange column. This sorbent was investigated for the sorption of americium from various aqueous media such as nitric acid, oxalic acid and sulphuric acid by distribution coefficient studies. Column experiments were carried out to study the practical application of this sorbent for removal of americium from oxalic acid-nitric acid solutions. Elution studies were also carried out for the recovery of americium.
D. M. Noronha, I. C. Pius, S. K. Mukerjee, J Radioanal Nucl Chem (2011) 289, 75–81

A1771 – Gibbs energy of formation of the Rh–Te intermetallic compounds Rh3Te2 and RhTe0:9

The vaporization behavior of the intermetallic compounds Rh3Te2 and RhTe0:9 was studied in the temperature range 1151–1234 and 1026–1092 K, respectively, by Knudsen effusion mass loss technique. The phase analysis of partially evaporated samples of Rh3Te2 (s) and RhTe0:9 (s) together with the available information on Te bearing vapor species revealed that the compounds incongruently volatilize as Rh3Te2 (s) = 3Rh (s)+ 2/nTen (g) and 3RhTe0:9 (s) = Rh3Te2 (s)+ 0.7/nTen (g), (n = 1; 2), respectively. The equilibrium vapor pressures of Te2 (g) and Te (g) were derived from the total pressure pðTenÞ measured over the mixtures Rh3Te2 (s) and Rh (s), and RhTe0:9 (s) and Rh3Te2 (s) in the respective cases. The standard Gibbs energy of formation of Rh3Te2 and RhTe0.9 derived using the above vapor pressure data and other auxiliary data could be expressed by the equations ?fG°(Rh3Te2, s) (kJ mol-1) = - 176.9 +0.039T ± 7.0 and ?fG°(RhTe0.9, s) (kJ mol-1) = -74.7+ 0.015T ± 3.0 kJ mol-1, respectively.
R. Mishra, M. Ali, S.R. Bharadwaj, D. Das, Journal of Nuclear Materials 321 (2003) 318–323

A1703 – Thermogravimetry-evolved gas analysis–mass spectrometry system for materials research

Thermal analysis is a widely used analytical technique for materials research. However, thermal analysis with simultaneous evolved gas analysis describes the thermal event more precisely and completely. Among various gas analytical techniques, mass spectrometry has many advantages. Hence, an ultra high vacuum (UHV) compatible mass spectrometry based evolved gas analysis (EGA–MS) system has been developed. This system consists of a measurement chamber housing a mass spectrometer, spinning rotor gauge and vacuum gauges coupled to a high vacuum, high temperature reaction chamber. A commercial thermogravimetric analyser (TGA: TG + DTA) is interfaced to it. Additional mass flow based gas/vapour delivery system and calibration gas inlets have been added to make it a versatile TGA–EGA–MS facility. This system which gives complete information on weight change, heat change, nature and content of evolved gases is being used for (i) temperature programmed decomposition (TPD), (ii) synthesis of nanocrystalline materials, (iii) gas–solid interactions and (iv) analysis of gas mixtures. The TPD of various inorganic oxyanion solids are studied and reaction intermediates/products are analysed off-line. The dynamic operating conditions are found to yield nanocrystalline products in many cases. This paper essentially describes design features involved in coupling the existing EGA–MS system to TGA, associated fluid handling systems, the system calibration procedures and results on temperature programmed decomposition. In addition, synthesis of a few nanocrystalline oxides by vacuum thermal decomposition, gas analysis and potential use of this facility as controlled atmosphere exposure facility for studying gas–solid interactions are also described
M Kamruddin, P Kajikumar, S Dash, A K Tyagi, Baldev Raj, Bull. Mater. Sci., Vol. 26, No. 4, June 2003, pp. 449–460

A1702 – Gibbs energy of formation of barium thorate (BaThO3) by reactive carrier gas technique

The Gibbs energy of formation of BaThO3 was determined employing the heterogeneous reaction between the compound and water vapour involving the formation of gaseous barium hydroxide species according to the reaction BaThO3(s) + H2O(g)ˆThO2(s) + Ba(OH)2(g). The vapour pressure of barium bearing species over the univariant mixture containing barium thorate and thorium dioxide as the condensed phases in equilibrium with a controlled pressure of water vapour was measured in the temperature range 1548±1683 K employing the automatic recording transpiration apparatus. The vaporization of BaThO3 was studied in the presence of ¯owing argon saturated with water vapour.
S.R. Bharadwaj, R. Mishra, M. Ali(Basu), D. Das, A.S. Kerkar, S.R. Dharwadkar, Journal of Nuclear Materials 275 (1999) 201-205

A1616 – On the origin of the sigmoid shape in the UO2 oxidation weight gain curves

Cracking and spalling are known to occur during the oxidation of UO2. However, these phenomena are not considered by the existing kinetic models of the oxidation of UO2 into U3O8. In this study the oxidation of UO2 samples of various sizes from the single crystal to nanopowders, has been followed by isothermal and isobaric thermogravimetry, environmental scanning electron microscopy and in situ X-ray diffraction at temperatures ranging from 250 to 370 ?C in air. It has been shown that cracking occurs once a critical layer thickness of intermediate oxide has been reached, which corresponds to the beginning of the sigmoid kinetic curve. Cracking contribution to the sigmoid kinetic curve is then discussed as a function of temperature, and on the basis of nucleation and growth processes.
L. Quémard, L. Desgranges, V. Bouineau, M. Pijolat, G. Baldinozzi, N. Millot, J.C. Nièpce, A. Poulesquen, Journal of the European Ceramic Society 29 (2009) 2791–2798

A1536 – Kinetic study of the effect of a sudden change in temperature during the reduction of U3O8 into UO2 by hydrogen

In the non-isothermal approach to the determination of kinetic constants, the mathematical analysis of the ?-time curve is based on the assumption of a rate equation, which in some cases, is not consistent with the isothermal curves. For example, during the reduction of U3O8 by hydrogen, we have shown from isothermal and isobaric experiments that a model of nucleation and anisotropic growth was able to fit the experimental rate curves. This makes the analysis of non-isothermal experiments very complex. A first attempt to solve this problem is presented here, in which the effect of a sudden change in the reaction temperature during an experiment is analysed. Taking into account the differences due to the sudden change in the calculation of the rate, a good agreement was found between the calculated and experimental d?/dt vs. ? curves
S. Perrin, M. Pijolat, F. Valdivieso, M. Soustelle, Solid State Ionics 141–142 (2001) 109–115

A1535 – Reduction of uranium oxide U3O8 into uranium dioxide UO2 by ammonia

The reduction of uranium oxide U3O8 into uranium dioxide UO2 has been studied by temperature-programmed thermogravimetry, up to 700°C. Experiments have been carried out either in ammonia (pNH3 ranging from 35 to 125 hPa). or hydrogen (PH2 ranging from 42 to 243 hPa.. The gases evolved and consumed during the reduction were followed H2 simultaneously by mass spectrometry. The reduction of U3O8 by ammonia into UO2 begins at 550°C, and is completed at about 650°C. It has been noticed that ammonia decomposition occurs at 700°C; moreover, it is catalysed by UO2 produced by the reduction of U3O8 , since no decomposition is observed in the absence of UO2 . Besides, some isothermal experiments carried out at 510°C have confirmed that ammonia reacts directly with U3O8 since the shape of the curves obtained either in ammonia or in hydrogen are different, particularly, the reaction is faster with ammonia compared to hydrogen, for the same partial pressure of the reducing gas.
Françoise Valdivieso, M. Pijolat, M. Soustelle, J. Jourde, Solid State Ionics 141–142 Ž2001. 117–122

A1520 – High temperature chromium volatilization from Cr2O3 powder and Cr2O3-doped UO2 pellets in reducing atmospheres

Chromium volatilization from Cr2O3 powder and Cr2O3-doped UO2 pellets during sintering in reducing atmospheres has been studied by thermogravimetry (TG) coupled with differential thermal analysis (DTA) up to 1700 °C. The sintering of Cr2O3-doped UO2 pellets was also followed by dilatometry. Oxygen partial pressures in the range 10ÿ20–10ÿ11 atm (10ÿ15–10ÿ6 Pa) have been fixed in all the experiments thanks to mixtures of hydrogen and carbon dioxide. A linear heating rate of 20 °C minÿ1 was applied to all the experiments. The dopant amount was in the range 0.18–0.9 mol% Cr in UO2. For all the oxygen potentials, the mass loss of Cr2O3 powder was found to start at temperatures as high as 1470 °C due to Cr2O3 dissociation, the lower the oxygen potential, the lower the starting temperature and the higher the volatilized amount. For intermediate oxygen potentials, an exothermic DTA peak observed during cooling, from 1700 °C to room temperature, attested for the crystallization of a liquid phase which was attributed to CrO(l) according to thermodynamic predictions. Then, the dilatometry and TG studies allowed following the behavior of Cr2O3-doped UO2 pellets. The mass loss at temperatures higher than 1470 °C was attributed to chromium volatilization for all the doped samples. During the sintering of doped UO2 pellets, the liquid phase CrO(l) seemed to appear at a lower oxygen potential than in Cr2O3 powder, which probably contributed to enhance the densification rate. For the highest dopant amount, 0.9 mol% Cr, the volatilization process was found to be rather similar to that of Cr2O3 powder, due to the part of chromia not solubilized in the UO2 crystal. Moreover, as the initial pellets were not dense, as long as the pellet porosity remained open, the chromia particles were able to dissociate as in the Cr2O3 powder. Thus the volatilization of chromium from doped UO2 pellets under sintering in reducing atmospheres must be understood as the result of several phenomena whose contribution depends on temperature, oxygen potential and heating rate: before the porosity closure, both dissociation of chromia into Cr(g) and oxygen from excess Cr2O3 particles, and Cr volatilization from doped UO2 particles; then chromium volatilization from the doped UO2 ceramic during further densification process.
V. Peres, L. Favergeon, M. Andrieu, J.C. Palussièreb, J. Balland, C. Delafoy, M. Pijolat, Journal of Nuclear Materials 423 (2012) 93–101

A1501 – Thermochemistry of decomposition of RE2O2CO3 (RE = Sm, Eu)

Thermochemistry in the decomposition of samarium di-oxycarbonate, Sm2O2CO3(s) and europium di-oxycarbonate, Eu2O2CO3(s) was studied over the temperature regions of 755–987 K and 773–989 K, respectively. The equilibrium properties of the decomposition reactions were obtained by tensimetric measurement of the CO2(g) pressure over the biphasic mixture of RE2O2CO3(s) and RE2O3(s) at different temperatures (RE = Sm, Eu) and also by thermogravimetric analysis of the decomposition temperature at 1 atmosphere of CO2 pressure. The median enthalpy and entropy of the decomposition of the oxycarbonates were calculated by the second law analysis and their thermodynamic stabilities were derived. The results are discussed in the light of available thermochemical data of the compounds
A.N. Shirsat, K.N.G. Kaimal, S.R. Bharadwaj, D. Das, Thermochimica Acta 477 (2008) 38-41

A1292 – Study of lanthanum orthophosphates polymorphism, in view of actinide conditioning

In order to perform researches on the substitution lanthanide-actinide in a view of actinide conditioning, a preliminary study of the polymorphism of lanthanum orthophosphates has been carried out by different techniques. LaPO4 formed by reaction of lanthanum nitrate with phosphoric acid contains 0.5 mol of water in open channel of the hexagonal structure (rhabdophane-type). The combination of thermogravimetric analysis, differential scanning calorimetry, X-ray diffraction and 31P solid-state nuclear magnetic resonance clearly shows the different steps of the thermal treatment. The zeolitic water evaporates between 180 and 280°C. After heating up to 700°C, a monoclinic structure (monazite-type) is formed by compacting the chains of PO4 tetrahedron alternating with LaO9 polyhedron.
B. Glorieux, M. Matecki, F. Fayon, J.P. Coutures,S. Palau, A. Douy, G. Peraudeau, Journal of Nuclear Materials 326 (2004) 156-162

A1285 – Structural study of (U0.90Ce0.10)4O(9-d), an anion-excess fluorite superstructure of U4O(9-d) type

In the U-Ce-O system, a solid solution (U,Ce)O(2+x) of fluorite type containing anionic excess is known in a wide composition range.For high values of x; it transforms to a (U(1-y)Cey)4O(9-d) phase deriving from the beta-U4O(9-d) type [ordered anion-excess fluorite superstructure phase; I-43d space group; a = 21.7484(1)Å for y = 0.10].The crystal structure of (U0.9Ce0.1)4O(9-d) has been refined by the Rietveld method on a powder sample measured on D2B at ILL Grenoble.The structural model, proposed by Bevan et al.for beta-U4O(9-d) and not fully confirmed till now, has been verified. The structure is based on an ordered distribution of cuboctahedral clusters U6O37 inside a fluorite matrix. A preferential ordering of Ce4+ (and U4+) on the so-called "centaur polyhedra" with 10 coordination is proposed, on the basis of bond valence calculations.The structure so determined has the composition M64O143 (MO2.234) and no traces of excess anions, completing the supposed composition up to M4O9, could be detected.
C.Rocanière, J.P. Laval, Ph.Dehaudt, B.Gaudreau, A.Chotard, and E.Suard, Journal of Solid State Chemistry 177 (2004) 1758-1767

A1269 – Immobilization of tetravalent actinides in phosphate ceramics

Three phosphate-based ceramics were studied for the immobilization of tri- and tetravalent actinides: britholite Ca9Nd(1-x)AnIV(x)(PO4)(5-x)(SiO4)(1+x)F2, monazite/brabantite LnIII(1-2x)CaxAnIV(x)PO4 and thorium phosphate diphosphate beta - Th(4-x)AnIV(x)(PO4)4P2O7 (beta-TPD). For each material, the incorporation of thorium and uranium (IV) was examined through dry chemistry routes, using mechanical grinding of the initial mixtures then heating at high temperature (1373- 1673 K). The quantitative incorporation of thorium in the britholite structure was obtained up to 20 wt% through the coupled substitution (Nd3+; (PO4)3- <=>(Th4+; (SiO4)4-). On the contrary, the incorporation of uranium was limited to 5-8 wt% and always led to a two-phase system composed by U-britholite and CaU2O(5+y). The incorporation of Th and U(IV) was also examined in both matrices, beta-TPD and monazite/brabantite solid solutions. Homogeneous and single phase samples of beta-TUPD and (Th,U)-monazite/brabantite solid solutions were obtained using successive cycles of mechanical grinding/calcination. The three matrices were prepared in the pellet form then leached in 10^(-1) M or 10^(-4) M HNO3 at 363 K. The very low normalized dissolution rates confirmed the good resistance of the materials to aqueous alteration. Moreover, in over-saturation conditions, the formation of neoformed phases onto the surface of the pellets was evidenced for several sintered samples.
O. Terra, N. Dacheux, F. Audubert, R. Podor, Journal of Nuclear Materials 352 (2006) 224-232

A1187 – Solid state reactions of CeO2, PuO2, (U,Ce)O2 and (U,Pu)O2 with K2S2O8

Solid state reactions of CeO2, PuO2 and mixed oxides (U,Ce)O2 and (U,Pu)O2 containing different mol.% of Ce and Pu, were carried out with K2S2O8 at different temperatures to identify the formation of various products and to investigate their dissolution behaviour. X-ray, chemical and thermal analysis methods were used to characterise the products formed at various temperatures. The products obtained by heating two moles of K2S2O8 with one mole each of CeO2, PuO2, (U,Ce)O2 and (U,Pu)O2 at 400°C were identified as K4Ce(SO4)4, K4Pu(SO4)4, K4(U,Ce)(SO4)4 and K4(U,Pu)(SO4)4, respectively. K4Ce(SO4)4 further decomposed to form K4Ce(SO4)3.5 at 600°C and mixture of K2SO4 and CeO2 at 950°C. Thus the products formed during the reaction of 2K2S2O8 + CeO2 show that cerium undergoes changes in oxidation state from +4 to +3 and again to +4. XRD data of K4Ce(SO4)4 and K4Ce(SO4)3.5 were indexed on triclinic and monoclinic system, respectively. PuO2 + 2K2S2O8 reacts at 400°C to form K4Pu(SO4)4 which was stable upto 750°C and further decomposes to form K2SO4 + PuO2 at 1000°C. The products formed at 400°C during the reactions of the oxides and mixed oxides were found to be readily soluble in 1-2 M HNO3.
Meera Keskar, U.M. Kasar, K.D. Singh Mudher, V. Venugopal, Journal of Nuclear Materials 334 (2004) 207-213

A1124 – Synthesis and characterization of low-temperature precursors of thorium-uranium (IV) phosphate-diphosphate solid solutions

Several compositions of new precursor of thorium-uranium (IV) phosphate-diphosphate solid solutions (Th(4-x)Ux (PO4)4P2O7, called beta-TUPD) were synthesized in closed PTFE containers either in autoclave (160°C) or on sand bath (90-160°C). All the samples appeared to be single phase. From XRD data and TEM observations, the diffraction lines matched well with that of pure thorium phosphate-hydrogenphosphate hydrate (TPHPH), Th2(PO4)2(HPO4).H2O, which confirmed the preparation of a complete solid solution between pure thorium and uranium (IV) compounds. TGA/DTA experiments showed that samples of thorium-uranium (IV) phosphate-hydrogenphosphate hydrate (TUPHPH) prepared at 150-160°C were monohydrated leading to the proposed formula Th((2-x)/2)U(x/2)(PO4)2(HPO4).H2O. The variation of the XRD diagrams versus the heating temperature showed that TUPHPH remained crystallized and single phase from room temperature to 200°C. After heating between 200°C and 800°C, the presence of diphosphate groups in the solid was evidenced. In this range of temperature, the solid was transformed into the low-temperature monoclinic form of thorium-uranium (IV) phosphate-diphosphate (alpha-TUPD). This latter compound finally turned into well-crystallized, homogeneous and single-phase beta-TUPD (orthorhombic form) above 930-950°C for x values lower than 2.80. For higher x values, a mixture of beta-TUPD, alpha-Th(1-z)UzP2O7 and U(2-w)ThwO(PO4)2 was obtained. By this new chemical route of preparation of beta-TUPD solid solutions, the homogeneity of the samples is significantly improved, especially considering the distribution of thorium and uranium.
N. Clavier, N. Dacheux, P. Martinez, V. Brandel, R. Podor, P. Le Coustumer, Journal of Nuclear Materials 335 (2004) 397-409

A1053 – Recycling of chemicals from alkaline waste generated during preparation of UO3 microspheres by sol-gel process

Internal gelation process, one of the sol-gel processes for nuclear fuel fabrication, offers many advantages over conventional powder pellet route. However, one of the limitation of the process is generation of large volume of alkaline liquid waste containing hexamethylenetetramine, urea, ammonium nitrate, ammonium hydroxide etc. Presence of ammonium nitrate with hexamethylenetetramine and urea presents a fire hazard which prevents direct disposal of the waste as well as its recycle by evaporation. The paper describes the studies carried out to suitably process the waste. Nitrate was removed from the waste by passing through Dowex 1 x 4 anion exchange resin in OH- form. 1.0 M NaOH was used to regenerate the resin. The nitrate-free waste was further treated to recover and recycle hexamethylenetetramine, urea and ammonium hydroxide for preparation of UO3 microspheres. The quality of the microspheres obtained was satisfactory. An optimized flow sheet for processing of the waste solution has been suggested.
A. Kumar, T.V. Vittal Rao, S.K. Mukerjee, V.N. Vaidya, Journal of Nuclear Materials 350 (2006) 254-263

A1014 – Immobilization of tetravalent actinides in phosphate ceramics

Three phosphate-based ceramics were studied for the immobilization of tri- and tetravalent actinides: britholite Ca9Nd(1-x)AnIV(x)(PO4)(5-x)(SiO4)(1+x)F2, monazite/brabantite LnIII(1-2x)CaxAnIV(x)PO4 and thorium phosphate diphosphate beta - Th(4-x)AnIV(x)(PO4)4P2O7 (beta-TPD). For each material, the incorporation of thorium and uranium (IV) was examined through dry chemistry routes, using mechanical grinding of the initial mixtures then heating at high temperature (1373- 1673 K). The quantitative incorporation of thorium in the britholite structure was obtained up to 20 wt% through the coupled substitution (Nd3+; (PO4)3- <=>(Th4+; (SiO4)4-). On the contrary, the incorporation of uranium was limited to 5-8 wt% and always led to a two-phase system composed by U-britholite and CaU2O(5+y). The incorporation of Th and U(IV) was also examined in both matrices, beta-TPD and monazite/brabantite solid solutions. Homogeneous and single phase samples of beta-TUPD and (Th,U)-monazite/brabantite solid solutions were obtained using successive cycles of mechanical grinding/calcination. The three matrices were prepared in the pellet form then leached in 10^(-1) M or 10^(-4) M HNO3 at 363 K. The very low normalized dissolution rates confirmed the good resistance of the materials to aqueous alteration. Moreover, in over-saturation conditions, the formation of neoformed phases onto the surface of the pellets was evidenced for several sintered samples.
O. Terra, N. Dacheux, F. Audubert, R. Podor, Journal of Nuclear Materials 352 (2006) 224-232

A1013 – Improvement of the preparation of sintered pellets of thorium phosphate-diphosphate and associated solid solutions from crystallized precursors

Several compositions of thorium-uranium (IV) phosphate-hydrogenphosphate hydrate (U(x/2)Th(2-x/2)(PO4)2- (HPO4).H2O, TUPHPH) were prepared starting from actinides chloride solutions and concentrated phosphoric acid. The experimental synthesis parameters were optimized in order to get the quantitative precipitation of the cations and a good crystallization state. The extensive characterization of the solids demonstrated the existence of a complete solid solution between Th and U end-members and evidenced the good homogeneity of the powders. Their behaviors during heating treatment were then checked and confirm the formation of anhydrous thorium-uranium (IV) phosphate-hydrogenphosphate (TUPHP) and alpha-UxTh(4-x)(PO4)4(P2O7) (alpha-TUPD) acting as intermediates. Finally, the low-temperature crystallized precursors were used in original sintering processes in order to improve the efficiency of the former cold-pressing sintering procedure.
N. Clavier, N. Dacheux, G. Wallez, M. Quarton, Journal of Nuclear Materials 352 (2006) 209-216

A0992 – Oxidation behaviour of unirradiated sintered UO2 pellets and powder at different oxygen partial pressures, above 350°C

The oxidation of sintered UO2 pellets and powder into U3O8 has been studied by thermogravimetry at 370°C, under controlled oxygen partial pressures (PO2 ranging from 2 to 40 kPa). Sigmoidal curves of oxidation weight gain were measured for both pellet and powder test samples. The rate of oxidation increased as the oxygen partial pressure increased. It has been shown, by simultaneous TG-DSC, that the reaction proceeds in a pseudo steady state. An experimental methodology based on temperature or PO2 jumps has shown that the assumption of a rate-limiting step is validated, and a mean value of activation energy for the formation of U3O8 of 103 kJ mol-1 was estimated.
F. Valdivieso, V. Francon, F. Byasson, M. Pijolat, A. Feugier, V. Peres, Journal of Nuclear Materials 354 (2006) 85-93

A0991 – Oxidation behaviour of unirradiated sintered UO2 pellets and powder at different oxygen partial pressures, above 350°C

The oxidation of sintered UO2 pellets and powder into U3O8 has been studied by thermogravimetry at 370°C, under controlled oxygen partial pressures (PO2 ranging from 2 to 40 kPa). Sigmoidal curves of oxidation weight gain were measured for both pellet and powder test samples. The rate of oxidation increased as the oxygen partial pressure increased. It has been shown, by simultaneous TG-DSC, that the reaction proceeds in a pseudo steady state. An experimental methodology based on temperature or PO2 jumps has shown that the assumption of a rate-limiting step is validated, and a mean value of activation energy for the formation of U3O8 of 103 kJ mol-1 was estimated.
F. Valdivieso, V. Francon, F. Byasson, M. Pijolat, A. Feugier, V. Peres, Journal of Nuclear Materials 354 (2006) 85-93

A0967 – A detailed study of UO2 to U3O8 oxidation phases and the associated rate-limiting steps

The kinetic and crystalline evolutions of UO2 during its oxidation into U3O8 at 250°C in air were studied by isothermal thermogravimetry and calorimetry, coupled with an in situ synchrotron X-ray diffraction on the D2AM-CRG beamline at ESRF. This study was aimed at determining experimentally the validity of the kinetic assumptions made in existing literature to account for the oxidation of UO2 into U3O8 and also to determine precisely the structural evolution, in relation to the kinetic behaviour. Our results provide evidence of four distinct kinetic time domains, and the assumption of a single rate-limiting step is verified only for two of them. The crystalline phases associated with these domains are also identified. In fact, the first kinetic domain corresponds to the reaction of UO2 into U4O9; the second one is linked to the two simultaneous reactions, UO2 into U4O9 and U4O9 into U3O7. Finally, the transition from U3O7 into U3O8 corresponds to the third and fourth kinetic domains. These results show that the oxidation of UO2 into U3O8 cannot satisfactorily be described with modelling approaches used in the literature. A new general outline is proposed to study the oxidation of uranium oxides. This outline will improve both the understanding and predictions of oxidation processes at the relatively low temperatures that are expected during interim storage of spent nuclear fuel.
G. Rousseau, L. Desgranges, F. Charlot, N. Millot, J.C. Nièpce, M. Pijolat, F. Valdivieso, G. Baldinozzi, J.F. Bérar, Journal of Nuclear Materials 355 (2006) 10-20

A0966 – A detailed study of UO2 to U3O8 oxidation phases and the associated rate-limiting steps

The kinetic and crystalline evolutions of UO2 during its oxidation into U3O8 at 250°C in air were studied by isothermal thermogravimetry and calorimetry, coupled with an in situ synchrotron X-ray diffraction on the D2AM-CRG beamline at ESRF. This study was aimed at determining experimentally the validity of the kinetic assumptions made in existing literature to account for the oxidation of UO2 into U3O8 and also to determine precisely the structural evolution, in relation to the kinetic behaviour. Our results provide evidence of four distinct kinetic time domains, and the assumption of a single rate-limiting step is verified only for two of them. The crystalline phases associated with these domains are also identified. In fact, the first kinetic domain corresponds to the reaction of UO2 into U4O9; the second one is linked to the two simultaneous reactions, UO2 into U4O9 and U4O9 into U3O7. Finally, the transition from U3O7 into U3O8 corresponds to the third and fourth kinetic domains. These results show that the oxidation of UO2 into U3O8 cannot satisfactorily be described with modelling approaches used in the literature. A new general outline is proposed to study the oxidation of uranium oxides. This outline will improve both the understanding and predictions of oxidation processes at the relatively low temperatures that are expected during interim storage of spent nuclear fuel.
G. Rousseau, L. Desgranges, F. Charlot, N. Millot, J.C. Nièpce, M. Pijolat, F. Valdivieso, G. Baldinozzi, J.F. Bérar, Journal of Nuclear Materials 355 (2006) 10-20

A0871 – Tungsten bronze-based nuclear waste form ceramics. Part 3: The system Cs0.3MxW(1-x)O3 for the immobilization of radio cesium

Previous studies in this series have indicated that Cs- and Sr-loaded Mo-doped hexagonal tungsten bronze (MoW-HTB) oxides, either in the form of fine grained powders, or as composite granules, can be converted to leach resistant ceramics at modest temperatures in the range 600-1200°C. In the present study it has been shown that such waste form ceramics can also be readily prepared through very simple conventional routes involving the blending of cesium nitrate with tungstic acid and other oxide components followed by heating in air. The phase chemistry resulting from the blending of these oxides has been explored. In the Cs0.3MxW(1-x)O3 compositional system where x = Ti, Zr, Nb and Ta the solid solution limit has been found to be where x = 0.2. For all values of x between 0 and 0.2 mixed phase materials of HTB andWO3 were obtained and Cs was found associated with HTB phases that are both rich and depleted in M element. At temperatures above about 1000°C, phase pure HTB compounds in the space group P63/mcm were obtained. Even when x greatly exceeds 0.2, the additional oxide content did not interfere with the formation of the HTB phase. Durability of the Cs0.3MxW(1-x)O3 compositions as gauged by the fractional Cs loss in de-mineralized water was lowest when M = Ti and Nb, and greatest when M = Zr. From these results the durability appears intimately linked with the unit cell a-dimension which in turn varies with M cation radius.
V. Luca, E. Drabarek, H. Chronis, T. McLeod, Journal of Nuclear Materials 358 (2006) 164-175

A0870 – Tungsten bronze-based nuclear waste form ceramics. Part 1. Conversion of microporous tungstates to leach resistant ceramics

The effective immobilization of Cs+ and/or Sr2+ sorbed on hexagonal tungsten oxide bronze (HTB) adsorbent materials has been achieved by heating in air at temperatures in the range 500-1000°C. Crystalline powdered HTB materials formed by heating at 800°C displayed leach characteristics comparable to Cs-containing hot-pressed hollandites in the pH range from 0 to 12. If the Cs-loaded HTB sorbents were pressed into pellets prior to calcination, ceramic monoliths could be prepared with negligible Cs volatilization losses. Heating to temperatures in excess of 1250°C under dynamic air flow resulted in the melting of the sorbent to form phase assemblages consisting of millimetre-sized crystals of bronzoid phases. Up to 5 wt% mass loss was observed for small scale samples of melted materials under dynamic air flow. Both the calcined and melted bronzoid waste forms are multiphase ceramics in which Cs+ remains bound within, and appears to stabilize, the hexagonal bronze phase, even after complete melting at 1300°C. The leachability of Sr from the phases prepared by heating appears to be somewhat worse than that of Cs. Saturation of the HTB adsorbents with lanthanide elements (Nd, La, Ce) gave rise to cubic bronze phases in which we propose that the lanthanides substitute at the tungsten or molybdenum sites rather than the tunnel positions. The lanthanides were rather easily leached from the calcined phases in 0.1 M HNO3 at 150°C.
V. Luca, C.S. Griffith, E. Drabarek, H. Chronis, Journal of Nuclear Materials 358 (2006) 139-150

A0587 – A thermodynamic approach for advanced fuels of gas-cooled reactors

For both high temperature reactor (HTR) and gas cooled fast reactor (GFR) systems, the high operating temperature in normal and accidental conditions necessitates the assessment of the thermodynamic data and associated phase diagrams for the complex system constituted of the fuel kernel, the inert materials and the fission products. A classical CALPHAD approach, coupling experiments and thermodynamic calculations, is proposed. Some examples of studies are presented leading with the CO and CO2 gas formation during the chemical interaction of [UO2±x/C] in the HTR particle, and the chemical compatibility of the couples [UN/SiC], [(U, Pu)N/SiC], [(U,Pu)N/TiN] for the GFR system. A project of constitution of a thermodynamic database for advanced fuels of gas-cooled reactors is proposed.
C. Guéneau, S. Chatain, S. Gossé, C. Rado, O. Rapaud, J. Lechelle, J.C. Dumas, C. Chatillon, Journal of Nuclear Materials 344 (2005) 191-197

A0399 – Application of differential thermal analysis for uranium oxide powders characterisation

Y. Balaji Rao, H.R. Ravindra, R.B. Yadav, B. Gopalan, S. Syamsundar, Thermans (2000) 158-160