Depletion with Online-Processing - Analysis and Evaluation of the Dual Fluid Reactor Concept

10.1 Description

10.1.1 Reprocessing Techniques

Just as the origin of the nuclear science can be traced to military applications, the reprocessing techniques of spent nuclear fuel (SNF) were also used for the extrac-tion of plutonium for nuclear weapons. Some early methods, such as the Bismuth Phosphate Process [BNG⁺01, p.1241-p.1256], were widely used for this purpose.

Nonetheless, these methods could not recover the uranium in the SNF and increased the volume of the waste by adding signicant, unnecessary amount of material to the nal radioactive waste. Such a characteristic precluded the use of those technologies for civilian applications.

The development of reprocessing techniques for civilian applications accompanied the commercialization of the nuclear power with the following goals:

• Recovery of the valuable ssile nuclides for further reuse as nuclear fuel;

• Reduction of the total amount/volume of the radioactive wastes;

• Extraction of specially valuable or needed nuclides.

The extracted plutonium and uranium were recovered and put back into the fuel cycle for the thermal reactors as Mixed Oxide Fuel (MOX), so that the plutonium would not be diverted for the development of nuclear weapons, thus contributing to the non-proliferation goals set by the International Atomic Energy Agency (IAEA).

The PUREX (Plutorium and Uuranium Recovery by EXtraction) reprocessing tech-nology was rst developed in 1949 at ORNL and is still in practice today. As an aqueous organic reprocessing technique, PUREX is similar to other processes in the class of hydro-metallurgy. It is, however, not suitable for online reprocessing, because it must let the fuel cool for not less than 5 years and then it converts the fuel into a highly acidic solution for further processing (see Fig. 10.1).

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Figure 10.1: Standard PUREX owsheet [NEA67]

Compared to the PUREX method, which operates at low temperature, pyroprocess-ing technology [Iiz01] [WW06] is a method suitable for high temperature situations, such as on-line reprocessing. The method was developed at the Argonne National Laboratory in the Integral Fast Reactor (IFR) [IFR] project. It dissolves the spent nuclear fuel in molten salts and molten metals, and then uses electro-rening or distillation to separate the ssion products. The most obvious dierence between hydro-metallurgy and pyro-metallurgy is that the former one uses a water-based sol-vent, while the latter uses a molten salt or metal as solvent. The use of a molten salt solvent confers this method with a higher compatibility than aqueous methods, which avoids the transportation of spent fuel and its related security issues. The spent fuel can just be stored on-site instead. Yet, this possesses also the disadvan-tage that the processed salt is less suitable for conversion into glass form if needed by the waste management strategy. In Russia, for instance, the pyro- and hydro-processing techniques are combined to prot from both advantages and to avoid their disadvantages [SVD⁺13].

In the IFR project a reprocessing unit using the pyro-metallurgical method was established on-site, which was already the embryonic form of the online processing envisioned for the DFR. Its succeeding project, the S-PRISM, has also used this processing technique, but abandoned the on-site design, by relocated the reprocessing facility to an o-site place [FLC⁺89].

With on-line reprocessing in the molten salt reactor it is possible to realize its unique advantage, that the reactor can keep normal operation while undergoing on-site, continuous removal of the ssion products generated from the liquid fuel by chemical and physical methods. Nevertheless, pyro-processing is just one possible option.

Since on-line reprocessing improves the performance of the reactor and the neutron economy, as well as, the breeding ratio by reducing the inventory of absorbers, it has been investigated from the very beginning as an integral part of the development of the molten salt reactor concept. Already in the course of the ARE project in

10.1. DESCRIPTION 101

1959 [CC59], a coupled liquid HF dissolution procedure with uorination for the recovery of the uranium and the solvent salt was developed.

Another example is the fuel processing facility integrated in the MSRE reactor build-ing in order to remove any accumulated oxides in the fuel or ush salt by chemical reactions with H2-HF and to recover the original ssionable nuclides for further reac-tor operation [Lin69, p.1]. Based on the successful experience of the MSRE project, the MSBR [McW70] project was designed with an on-line processing unit from the start, which was an improved processing method to separate thorium from the rare earth ssion products. It required a single molten salt mixture LiF-BeF2-ThF4 -UF4 [Nee71]. Unfortunately, this idea was not to be realized because of the eventual cancellation of the MSBR project [CN72].

In the optimization of the MSFR project, a combined fuel salt management mech-anism combining a salt control unit, an on-line gaseous extraction system, and an o-line lanthanide extraction component by pyro-chemistry was introduced. This system extracts gaseous elements with a removal period of 30sand a fraction of the fuel is then withdrawn to extract the lanthanides in an o-line procedure, the rate of which depends on the desired breeding performance. With a rate of 40l/day, the whole core can be reprocessed in 450 days. The ssion products are slowly removed at a rate of 0.4l/day, which results in a total clean up of the fertile salt volume (7.3m³) in 52.7 years [BMLR⁺13, p.12-p.13].

10.1.2 Fuel Reprocessing of the DFR

The DFR is designed to have an on-line processing unit for the purication of the fuel, for the removal of the ssion products, and for the processing of the breeder salt. Considering the composition of the fuel salt and the ssion products, there are generally four dierent categories of ssion products that need to be removed:

• Gaseous elements or compounds

• Non-soluble elements or compounds

• Lanthanide and other elements or compounds

• Actinide elements or compounds

Based on the composition described in the previous sections, the fuel salt is supposed to remove the rst three categories: gaseous, non-soluble, as well as, lanthanide ele-ments or compounds; all eleele-ments whose Z<90. The gases have a very low solubility in the molten salts (approximately 10⁻⁸moles/cm³) and they diuse towards the core boundaries [ABC⁺08, p.14]. These elements are, thus, 100% removed, which means that at the beginning of each cycle their amount is set to 0. The elements whose Z is larger than 90, the actinides, are handled individually.

The capture-to-ssion ratios of the actinides are plotted in Fig. 10.2. Depending on this value, it is clear to decide which nuclide should remain in the salt and which will be removed. From the point of view of the neutron population inside the reactor, if

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