Impregnation Process

The impregnation process being developed in BARC, is particularly suitable for the fabrication of (ThU²³³) O2 fuel pellets which requires a heavily shielded facility as U²³³always contains some amount of U²³² which has high energy gamma emitting daughter products [7].

The major advantage of this process is that a large part of the fabrication processes including the handling of fine thoria powders for low density pellet fabrication and that of thorium nitrate solution for the preparation of gel microspheres could be carried out in an unshielded facility. The remaining process steps including the impregnation of pellets or microspheres with U²³³ bearing nitrate solution could be carried out in the shielded facility.

In pellet impregnation process, partially sintered thoria pellets fabricated by conventional powder metallurgical process were used for impregnation with uranyl nitrate solution under vacuum. The impregnated pellets were dried and finally sintered to high density at 1923K in N2 – H2 atmosphere. The amount of uranium impregnated into partially sintered pellets depended mainly on the initial porosities and the molarity of the solution. However under similar conditions, the uranium content increased with the use of annular pellets.

In microsphere impregnation technique, dried thoria gel microspheres prepared by ammonia internal gelation process at Fuel Chemistry Division, BARC [8] were impregnated with uranyl nitrate solution. Impregnated microspheres were dried, calcined and reduced at 973K to obtain free flowable porous microspheres . Irregular shaped pores obtained within the microspheres due to dissociation of uranyl nitrate made the microspheres easily crushable.

Table II: Characteristics of undoped and doped pellets

These microspheres were cold – pelletised at 300 MPa to green pellets of density ~ 60% T.D.

Pellets were sintered at 1923K in reductive atmosphere for four hours. The process flow sheet is shown in Fig.2. The characteristics of calcined ThO2 – UO2 microspheres used as feed materials and the pellets made from these microspheres are shown in Table III. Higher density could have been achieved with MgO doped ThO2 microspheres.

Lattice parameters (Table IV) calculated from the high angle X-ray diffraction patterns of sintered pellets confirmed the formation of solid solution between ThO2 and UO2. The compositions of solid solutions were also found out from Vegard^’ s law (Table IV).

Depending on the molarity of uranyl nitrate solution, a wide range of uranium loading is possible in this technique. Electron probe micro-analysis results show the non-uniform distribution of uranium and thorium in the pellets made from micro spheres impregnated with 3.0 M uranyl nitrate solution. This may be due to poor penetration of highly viscous higher molarity uranyl nitrate solution to the fine pores of thoria micro spheres. However, U and Th distribution were substantially improved in case of lower molarity solutions as shown.

Table III: Characteristics of thoria – urania microspheres and pellets Phase analysis of sintered pellet

(by XRD) Single phase Single phase Single phase

Table IV: lattice parameters and solid solution compositions of (ThU)O2 pellets.

Sl.No. Material Lattice Parameter (nm)

Solid solution.

Composition

1

Pellets made from microspheres impregnated with 0.5 M uranyl

nitrate solution 0.5585 (Th0.881U0.119)O2

2

Pellets made from microspheres impregnated with 1.0 M uranyl

nitrate solution 0.5583 (Th0.865U0.135)O2

3

Pellets made from microspheres impregnated with 3.0 M uranyl

nitrate solution 0.5568 (Th0.744U0.256)O2

Thoria Prepared by

Microspheres Ammonia Internal (size : ~600 micron) Gelation Process

Impregnation with Uranyl nitrate soln.

Calcination

& Temp. : 973 K Reduction Atm. : N2 – H2

Porous Aparent density Microspheres Tap density Video Microscopy

Cold Pressure : 300 Mpa Compaction Green density : ~60%T.D

Sintering Temp. : 1923 K Atm. ; N2-H2

Sintered Pellets

Density XRD , EPMA

Characterization Microstructure Th & U analysis

FIG. 1. Flowsheet for the fabrication of thoria – urania fuel pellets by microsphere impregnation technique

5. CONCLUSIONS

1. The feasibility of SGMP-LTS route for the fabrication of UO2 and (UPuO2) pellets has been demonstrated. Two fuel bundles containing UO2 pellets prepared by SGMP – LTS route have already been irradiated in pressurized heavy water reactor.

2. TiO2 dopant and silicate additives were found to increase the grain size of oxide fuel pellets sintered in reducing atmosphere. Large grain pellets have also found to be higher thermal conductivity by 10–15%.

3. Fuel pins containing low temperature sintered (UPu) O2 pellets and pellets with modified microstructure (large grain & controlled porosity) have been test irradiated in research reactor at BARC.

4. Impregnation technique appears to be more suitable for the fabrication of highly gamma active U²³³ bearing mixed oxide fuel pellets. The most attractive feature of this technique is that the major part of fabrication processes can be carried out in an unshielded facility.

Handling of fine powders are not involved in the shielded facility.

REFERENCES

[1] RADFORD, K.C., POPE, J.M., UO2 pellet microstructure modification through impurity addition, J. Nucl. Mat. 116 (1983) 305.

[2] GANGULY, C., BASAK, U., Fabrication of high density UO2 fuel pellets involving sol-gel microsphere pelletisation and low temperature sintering, J.Nucl.Mat.178 (1991)179.

[3] GANGULY, C., BASAK, U., SOOD, D. D., VAIDYA, V.N., ROY, P. R., “Sol-gel microsphere pelletisation of UO2 and UO2 - PuO2 pellets of PHWR fuel specification using internal gelation process”, CANDU Fuel (Proc. 2^nd Int. Conf. 1989), CNS, Toronto (1990) 108.

[4] ALTEKAR, R.M.,BASAK, U.,PANDEY, V.D.,SHEIKH,I.H.,RAMCHANDRAN,R., MAJUMDAR,S.,Microstructural modifications of UO2 fuel pellets through silicate additives,Trans.PMAI.27 (2001) 92.

[5] GANGULY, C., BASAK, U., SOOD, D.D., VAIDYA, V.N., BALARAMAMOORTHY, K., “SGMP – LTS process for fabrication of high density UO2 and (UPu)O2 fuel pellets”, CANDU Fuel (Proc. 3^rd Int. Conf. 1992), CNS, Toronto (1993).

[6] VAIDYA, V.N., MUKHERJEE, S.K., JOSHI, J.K., KAMATH, R.V., SOOD,D.D., A study of chemical parameters of the internal gelation based sol-gel process for uranium dioxide, J. Nucl. Mat. 148 (1987) 324.

[7] BASAK, U., MISHRA, S., NAIR,M.R., RAMACHANDRAN, R., MAJUMDAR, S., KAMATH, H.S., “Process development for the fabrication of thoria and thoria based nuclear fuel, Characterization and Quality Control in Nuclear Fuel Fabrication (Proc. Int.

Conf. Hyderabad, 2002), NFC, Hyderabad (in press).

[8] KUMAR, N., SHARMA, R.K., GANATRA, V.R., MUKHERJEE, S.K., VAIDYA, V.N., SOOD, D .D., Studies on the preparation of thoria and thoria –urania microspheres using and internal gelation process, Nucl. Techn. 96 (1991) 169.

INVESTIGATION OF THERMAL-PHYSICAL AND MECHANICAL PROPERTIES OF URANIUM-GADOLINIUM OXIDE FUEL

YU.K. BIBILASHVILI, A.V. KULESHOV

O.V. MILOVANOV, E.N. MIKHEEV, V.V. NOVIKOV A.A. Bochvar ARSRIIM

S.G. POPOV, V.N. PROSELKOV

Russian Research Centre “Kurchatov Institute”, Institute of Nuclear Reactors

YU.V. PIMENOV JSC “TVEL”

YU. G. GODIN MEPhI

Moscow, Russia Federation Abstract

The work is devoted to investigation of thermal-physical (coefficient of linear thermal expansion, thermal conductivity coefficient) and mechanical (brittle-ductile transition temperature, Young’s modulus, thermal creep) properties of uranium-gadolinium oxide fuel.

1. INTRODUCTION

For compensation for the initial excess reactivity, power flattening over the core volume and maintaining the temperature coefficient of reactivity at the specified level in the VVER, PWR and BWR reactors, the burnable absorbers (BA) are used. In the PWR reactors the BA made of ¹⁰B in the form of tetraboron carbide (B₄C) is conventionally used. As B₄C is chemically incompatible with uranium dioxide it was found necessary to provide some fuel rods in the fuel assembly (FA) with the burnable absorber (BAR). The use of BAR as an independent structural element has the following disadvantages: the BAR takes some useful space in the FA thus increasing the linear heat generation rate of fuel rods; neutron-absorbing structural material of BAR claddings is introduced into the FA; necessity of storing and transportation of irradiated BAR. This made the FA designers to search for other ways of BA introducing into the FA. One of these solutions was to deposit a thin layer of zirconium diboride on the fuel pellet surface. However the use of gadolinium distributed in the form of oxide in the uranium dioxide fuel matrix, has found a wider application. Gadolinium possesses rare properties due to a high neutron absorption cross section (essentially higher than that of ¹⁰B) and the rate of burnup close, in the case of optimal composition, to the rate fuel burnup of ²³⁵U. In addition, in irradiation of gadolinium gives no daughter products with a high neutron capture cross section, and gadolinium oxide interacting with uranium dioxide forms solid solutions with concentration up to ≈ 30 wt % Gd₂O₃ at the temperature 1873K [3]. The technology of uranium-gadolinium fuel does not, in principle, differ from production of fuel pellets from uranium dioxide [11,12]. Although the uranium-gadolinium oxide fuel has been used for a long time (in the BWR – since the 70s) its physical-chemical properties has not been sufficiently studied. For example, the data on its thermal-physical properties (thermal expansion, thermal conductivity) are available, though not completely, in the open literature,

while there are almost no data on the mechanical properties of uranium-gadolinium oxide fuel. This work is devoted, in part, to elimination of this deficit.