R EPROCESSING OF T HORIUM -B ASE I RRADIATED F UELS AND W ASTE M ANAGEMENT

The true thorium cycle requires breeding and recycling of U-233. The reprocessing technology is proven on a commercial scale. Advanced dry processes are not proven for thorium, and for the time being, reprocessing options are limited to wet chemical operations in which the fuel

substance is dissolved in nitric acid and separated by solvent extraction. Thorium is considerably less amenable than uranium to such processing.

Reprocessing difficulties are especially severe with HTR-type fuels of whatever composition, since the particle coatings and graphite matrix are chemically resistant and troublesome to break down mechanically.

6.3.1 The “Head-End” Operations

The aim of these operations is to obtain, after preliminary operations leading to dissolution, a solution of U-Th nitrates together with a small quantity of minor actinides and fission products.

Extraction of U-233 from irradiated thorium rods was conducted extensively in the US from the 1950s until the 1970s. Almost 700 tonnes thorium were irradiated, delivering more than 1.5 tonnes of U-233. The separation operations were piloted at ORNL in the THOREX plant, and in the Knolls Atomic Power Laboratory. The process was later adapted to the Great “canyon” plants at Hanford and Savannah River (SRP), and the commercial direct maintenance plant of Nuclear Fuel Services at West Valley near Buffalo, now being dismantled, which reprocessed the first core of Indian Point 1 (95% ThO2, 5% U-235O2), but without recovering thorium (which was left in the waste).

The processes are explained in an interesting paper [146], which is one of the early documents giving details of the processes and of the adaptation of these large PUREX-type plants to the THOREX processes (Figure 6.3).

Figure 6.3: Outline of the THOREX Process (Thorium Extraction).

Head-end operations for fertile fuels to extract U-233: Due to its great stability, dissolution of thorium or ThO2 is not as straightforward as that of U, and especially of UO2. Strong nitric acid with HF is required, and the process takes a rather long time, especially for high temperature sintered compact ThO2. Dissolution can take up to 35 hours. To prevent corrosion of the equipment by HF, aluminium nitrate was added as a buffer.

Similar processes have been used in India on a small scale for thorium and ThO2 rods and in other countries (UK at Dounray, Canada at Whiteshell, Germany at Jülich and Karlsruhe). In

India, the engineering pilot scale laboratory has been replaced by a 50 tonnes/year reprocessing plant at the Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, India.

In summary, “traditional” reprocessing plants can process irradiated thoria fuels, even though the dissolution is more difficult and precautions are required against HF corrosion.

Head-end operations for oxide LWR fuels: The pilot tests made at Oak Ridge were confirmed by

“commercial runs” of the PWR Indian Point 1 (U,Th)O2 core in the West Valley plant, for sizeable irradiation rates (16 000 MWd/tHM). The PWR fuel bundles clad in Zircaloy are sheared and the fuel is dissolved. It was found that, although irradiated (U,Th)O2 dissolves quicker than unirradiated material, the dissolution is much slower than for UO2 alone, as the thorium oxide is more compact and sintered. Concentrated nitric acid is needed with the resulting corrosion problems. The dissolver offgases will contain somewhat more Kr-85, depending on the proportion of U-233, as the yield of Kr-85 by fission is about double that of U-235. Iodine yield is about the same as for U-235.

Head-end operations for oxide or carbide HTR fuels are quite different and pose problems which have not been totally solved on the industrial scale, although many engineering tests have been performed, mainly at Oak Ridge, at General Dynamics in the USA and in Jülich (Germany), for their HTR programs. Limited tests have been performed in other laboratories, especially at the National Committee for Nuclear Energy (CNEN), Italy. Coated oxide and carbide fuel is very difficult to process. This is mainly due to its PyC and impervious SiC coatings, and in part to the unavoidable carbon residue. The same is true for the TRISO-coated particles, for which different tests have been made in Jülich on the famous “pebbles” of the AVR and THTR reactors. Most of the irradiated pebbles have not been further processed. Direct dissolution of the coated particles has been tested at CNEN/ENEA and at IGCAR (India), using electrolytic dissolution. The laboratory tests are promising, although the dissolution times are longer and the equipment more complex than usual. It would be important to test such devices on a large scale to assess whether they work satisfactorily.

The complexity of these pre-treatments, and the technological problems associated with them, are one of the reasons why the High Temperature Reactor family has not been further developed.

Broadly speaking, the coated kernel fuel is so resistant that it cannot be reprocessed by conventional reprocessing methods. The coated kernels, today, are seemingly the typical “through” fuel. They could be reserved for high temperature “burners” working on long once-through cycles (lasting many years), delivering high temperature gas. On the other hand, the

“breeders”, thermal or fast, should have fuel elements directly amenable to a simple reprocessing process, i.e. of the traditional oxide-in-clad type, water or liquid metal cooled.

6.3.2 Separation and Purification Operations: the THOREX Process

The reprocessing of thorium containing fuels may be carried out using the THOREX (thorium extraction) process. This was originally developed in the US at the Knolls Atomic Power Laboratory (KAPL) and the Oak Ridge National Laboratory (ORNL) in order to reprocess large amounts of irradiated thorium from light water reactors. Development continued in Germany for the reprocessing of High Temperature Reactor (HTR) fuel, which is described in the report from Kugeler [48].

The heavy metal and matrix graphite are first separated in the head-end of the reprocessing plant by burning of the graphite. For this step, the fuel elements can be crushed to pieces smaller than 5 mm. The burning of the coatings, including SiC-coatings, can be carried out in fluidised

bed ovens. The heavy metal ash is dissolved in the THOREX solution in the next step.

Afterwards, separation into uranium, thorium and fission products is carried out in a solvent–

extraction process in the chemical process part of the reprocessing plant. Additional equipment for the conditioning of high active waste, and for filtering gaseous fission products and other pollutants are necessary for the reprocessing of HTR fuel. The experiences gained with the reprocessing of LWR fuels may be used to a high degree in the THOREX process.

Extended research and development of the reprocessing of thorium containing fuel has been carried out in the last decades; and the feasibility of the process has been proven. The JUPITER facility, which includes the THOREX process and head end facilities, has been built and operated in Germany.

The PUREX (plutonium uranium extraction) process has become the reprocessing procedure generally used for all fuel types containing natural, slightly or highly enriched uranium together with lower or higher contents of plutonium. The THOREX process, on the other hand, has been developed for reprocessing uranium and thorium from thorium-based fuel. When the fuel contains appreciable amounts of U-238, plutonium will be produced and a combination of the THOREX and PUREX processes must be applied.

The THOREX process is technically less advanced and has the drawback that thorium nitrate exhibits a much lower distribution coefficient than uranium and plutonium. To drive thorium into the organic tributyl phosphate (TBP) phase, a strong salting agent is required. Aluminium nitrate, which has been recommended previously, has now been replaced by nitric acid in order to reduce the amount of radioactive waste. However, high acid concentrations are counter-effective in achieving high fission product decontamination. Therefore, several flow-sheet variants, with acid and acid deficient feed solutions, have been investigated in the past. In order to achieve high decontamination factors, a dual cycle THOREX process was developed. This process uses an acid feed solution in the first cycle and an acid deficient one in the second cycle.

According to recent investigations, a single cycle process with acid feed solution should provide the necessary decontamination factors. An immediate separation of thorium and uranium appears advisable in view of both fuel cycle strategy and process feasibility. Pulse columns should be preferentially used as extraction apparatus, at least for the extraction step.

Reprocessing of HTR-LEU fuel in existing PUREX plants has associated problems of criticality prevention due to the residual total fissile isotope content in the feed of greater than 2 wt%.

Special precautions must therefore be taken in several processing units. The addition of soluble poison to the feed proved to be an unfavourable measure. A more suitable method to overcome criticality difficulties may be the application of a flow-sheet with lower tributyl phosphate (TBP) concentrations, less than 10 %.

6.3.3 Waste Treatment

The waste treatment from a THOREX reprocessing plant will not differ much from that of a modern PUREX reprocessing plant. However, two remarks seem appropriate:

• The fact that the thorium fuel cycle waste will contain less minor actinides and hence will be less radiotoxic for the first 10 000 years or so.

• Dissolution of thoria-based fuels requires a small proportion of HF, buffered by aluminium nitrate. The effect of the fluoride on the vitrification of the waste must be ascertained.

6.3.4 The Dry Processes

Dry processes have been envisaged for either the processing of carbide kernels or for the in-line processing of the molten salts of the molten salt breeder reactors.

For the kernels, a heat treatment to up to 2800 – 3000ºC followed by chloridation has been envisaged. For the molten salt fluorides, the fluoride volatilization process has been extensively investigated in a few countries (USA, Argone National Laboratory, France, UK, etc.), generally with success. Advantages of the process include its relative compactness and absence of aqeous effluents. The drawbacks are the problems of working with fluorine and the corrosion of the equipment.

In the case of the ADS (Energy Amplifier Project), pyroprocessing is regarded as a key technology in many aspects. In comparison with acqueous reprocessing, it promises:

• Compactness and simplicity;

• Less secondary wastes;

• Proliferation resistance (no separation of the TRUs);

• Fuel fabrication and reprocessing at the reactor site.

The spent fuel rods are chopped into small sections, which allow separation of the spent fuel from the cladding. Fission product gases released at this stage are collected through the ventilation systems on filters and sent for storage. The spent oxide fuel is then converted to metal. In this process calcium reacts with the oxide fuel to produce calcium oxide and heavy metals (U, Np, Pu, Am, Cm). The reaction takes place in a high temperature molten calcium chloride salt bath.

Again, fission product gases released are collected and sent for storage. Metals such as Cs, Sr, and Ba are partitioned from the molten salt, which is periodically removed for storage. The resulting heavy metal is then sent for electro-refining. This is an electro-chemical process in which the thorium is separated from the actinide and fission product mixture. A NaCl-KCl molten salt at 1000 K is the transport medium. The thorium is collected at the cathode and removed periodically for further processing. Noble metal fission products (Zr, Mo, Ru, etc.) remain at the anode heel in the cell. The actinides and rare earth fission products remain in the molten salt.

This salt is then sent for further treatment in the winning process. This is also an electro-chemical process and is used to deposit the actinides (present in the form of chlorides) from the NaCl-KCl molten salt at the cathode of the cell.

However, development of the technology has not yet reached the stage where it can be demonstrated on an industrial scale.