GB2536857A

GB2536857A - Simple fuel cycle for molten salt reactors

Info

Publication number: GB2536857A
Application number: GB1418031.9A
Authority: GB
Inventors: Richard Scott Ian
Original assignee: Individual
Current assignee: Individual
Priority date: 2014-10-12
Filing date: 2014-10-12
Publication date: 2016-10-05
Also published as: GB201418031D0

Abstract

Fuel for a molten salt fuelled nuclear reactor is prepared from spent nuclear fuel by removal of excess uranium, isolation of actinides by conversion to halides, reduction of the halides to metal and conversion of the metal to halide in such a way that noble metal fission products and group 1 and 2 elements are largely removed and lanthanide fission products are partially removed. This is performed by reducing the uranium content via volitilisation as a tetrahalide, under conditions where the plutonium, curium and americium remain as trihalides, which are then reduced to metals, and finally conversion of the resulting metals to halide salts for use in a molten salt reactor. Also provided is a means of recycling spent nuclear fuel by the addition of concentrated fissile isotopes coupled with reduction in concentration of uranium or other neutron absorbing but non fissile isotope in the reactor and removal of noble metal fission products.

Description

SIMPLE FUEL CYCLE SYSTEM FOR MOLTEN SALT REACTORS

BACKGROUND

A novel design for a molten salt based nuclear reactor was disclosed in UK patent application number 1402908.6 entitled "A practical molten salt fission reactor". The basis for the design was to place the molten salt fissile material in static tubes from which heat was transferred to a coolant liquid by a combination of conduction and convection.

The fuel cycle used for conventional nuclear reactors is not suitable either for producing fuel for molten salt reactors nor for reprocessing such fuel after use. Fuel processing systems based on electrolysis of molten salts containing the fissile isotopes are known but are designed to produce fuel in metallic form for metal fuelled reactors, see for example Soucek et al, Energy Procedia, 7 (2011)396-404. A primary focus of such systems is the recovery of uranium or other actinides in the relatively pure form required for fabrication of solid nuclear fuels.

Certain nuclear reactors using molten salts as fuel are able to use much less pure fissile materials than are usable for solid fuel fabrication. The reactor described in GB1402908.6 is particularly able to use impure fuels because of its resistance to blockage due to insoluble deposits. There is therefore a need for a fuel cycle that takes advantage of this ability to use impure fuel to simplify and make cheaper and safer the process of making and reprocessing the fuel.

DESCRIPTION OF THE INVENTION

Fuel for the molten salt reactor can be prepared from spent fuel from conventional light water or gas cooled reactors. Such fuel contains about 95% uranium which has been largely depleted in fissile 235U and 1% plutonium and small amounts of higher actinides. It also contains about 4% of fission products. To convert this to usable fuel for a molten salt reactor, most of the uranium and some of the fission products must be removed.

Electrolytic processes described in the literature convert the spent fuel into halide salts (usually chloride) and then selectively deposit uranium on a solid iron cathode followed by depositing other actinides on metal cathodes (molten bismuth, molten cadmium or aluminium) that provide some selectivity between deposition of actinides and lanthanides. The -100 fold higher concentration of uranium than higher actinides in such spent fuel makes the apparatus bulky and the separation of higher actinides from the large excess of uranium relatively difficult.

It has been found that an initial treatment of the spent fuel can almost quantitatively remove uranium from it while leaving plutonium and higher actinides behind. The spent fuel, in the form of oxides or other compounds, is first chlorinated by any of the known methods, which include chlorine gas plus carbon, CC14 or other known chlorinating agents. Chlorination is continued until the uranium and neptunium in the spent fuel are in the form of tetrachlorides, which volatilise at the temperature of the reaction. The chlorination is stopped before PuCI3 is chlorinated to PuC14. The different reactivities of uranium and neptunium from the higher actinides permit this as shown below.

Optionally, the zircalloy cladding of the spent fuel pins can he left in place during this process and chlorinated to ZrC14 which is volatilised at a lower temperature than needed for the actinide tetrachlorides. Other volatile fission products including iodine, cadmium and tellurium are also volatilised at the lower temperature and can hence optionally be separated from the uranium and neptunium chlorides.

With chlorine as the sole chlorinating agent, it is relatively difficult to achieve complete chlorination to UC14 without producing a little PuC14. This can be avoided by using a chlorinating agent capable of converting UC13 to UCI4 but not PuC13 to PuC14.

A particularly useful chlorinating agent is MoC13 which can be used either as the sole chlorinating agent or to complete the last stage of the chlorination. Its reaction with UC13 and PuCk is shown below and it allows complete conversion of UCI3 to UCI4 with negligible production of PuC14.

In all cases, superior specificity of chlorination can he achieved by addition of molten salts such as chlorides of the group 1 and 2 metals so that the chlorination reaction takes place in the liquid medium rather as a gas/solid reaction as was described for example in US2875021.

Following this chlorination step, the mixed chloride salts are decanted away from the residual solids which are principally the noble metal fission products. UCI4 and NpC14 are then removed by volatilisation leaving a mixture highly enriched in fission products and higher actinide trichlorides. These can optionally be diluted by addition of NaCI and/or MgCl2 or other salts to lower their melting point.

The mixed fission product/actinide chlorides arc then used as feedstock for a conventional molten salt electrolytic separation. However, such separations usually fail to provide effective Pluu)nium and higher actinide corncaonds nuan and Neptu ompouncis * in titan by tvint i at SOO'kt la %) separation of actinides from lanthanides, requiring the use of complex electrode systems such as aluminium or cadmium electrodes to achieve sufficient separation. In preparing fuel for the molten salt reactor a far larger degree of contamination of the actinides with lanthanides is acceptable making us of simple electrodes such as bismuth practical. Recovery of the actinides as trichlorides is then straightforwardly done by partial chlorination of the electrode material resulting in the formation of a separate phase of actinide chlorides.

An alternative and rather simpler process is to reduce the molten salt mixture by slow addition of a reactive metal such as sodium, potassium or lithium The graph below shows the reduction of the actinides and lanthanides by sodium at 600°C in an NaCl based molten salt. Americium is the most similar of the actinides to the lanthanides, as is well known, and is the last actinide to be reduced, but the graph shows that complete reduction to metal of all actinides including americium can be achieved with only a small part of the lanthanides (mostly Pm, Y and a fraction of the Nd) being reduced. Notably, Sm which includes the most potent neutron poisoning isotope is barely reduced at all.

Rat.attior, of Actiftwo and Lanthani4e chlorides. to metil by sathu The metallic actinides, together with a small amount of metallic lanthanides are separated by decantation/filtration from the residual molten salt and rechlorinated to the trichloride form as fuel ready salts. Minor contamination of the metallic materials with the residual molten salt mixture is acceptable for use a fuel, thereby making a washing stage with fresh molten salt an optional and in most cases unnecessary step.

Fuel salt which has been used in the reactor described in GB 1402908.6 still contains substantial amounts of residual fissile material. Conventionally it is normal to reprocess such fuel using chemical or electrochemical methods. However, it has surprisingly been discovered that the fuel can he recycled into the reactor by simply adding additional fissile dded material and removing insoluble noble metal fission products. For this "refreshing" process to be practical several requirements must all be met.

The reactor must be an epithermal or fast neutron reactor.

The fuel must contain 238U or another neutron absorbing, preferably fertile, but non fissile isotope in addition to its fissile isotopes The noble metal fission products must be readily removable from the spent fuel. In the reactor described in GB1402908.6 that is simply done be replacing the fuel tubes on which the noble metals will have plated out.

While not essential, it is desirable that zirconium fission products should have been volatilised from the spent fuel salt during operation of the reactor.

The basis of the success of the process is shown in the figures below.

On the left is shown the neutron absorption (poisoning) effect of different components of the spent fuel. Removal of the noble metals (atomic numbers 41-47), the zirconium and noble gasses reduces the neutron poisoning by about 60% depending on the neutron energy. What remains are primarily the lanthanides and the elements of group 1 and 2 of the periodic table, especially cesium which is the strongest neutron absorber produced.

On the right is shown the equivalent amount by which 238U would have to be reduced to offset 100% of the fission product neutron poisons.

It is clear that a small reduction in232U concentration, such as would be achieved by a combination of consumption of 235U by fission or breeding to -239 Pu during reactor operation and by dilution by addition of plutonium with low levels of uranium contamination during "refreshment" will be sufficient to offset the increased neutron absorption by accumulation of the fission products This refreshing process can be repeated several times until the concentration of 238U in the fuel falls to a low level. r,o:c

NetitrOfx Eno, Vj 1..EE,03 Neutron energy;eV) 23813 reduction to compensate for fission products from 1 mol 239PU

Claims

CLAIMS1) A method for producing fissile fuel for a molten salt nuclear reactor from spent nuclear fuel comprising reducing the uranium content by volatilisation of uranium at the tetrahalide under conditions where plutonium, curium and americium remain as trihalides, reduction of the resulting plutonium, curium and americium trihalides to metals under conditions where only a fraction of the lanthanide halides are reduced to metal and conversion of the resulting metals to halide salts for use as fuel in the molten salt reactor 2) Recycling of spent fuel from an epithermal or fast neutron molten salt fuelled nuclear reactor by addition of concentrated fissile isotopes coupled with reduction in concentration of uranium or other neutron absorbing but non fissile isotope in the reactor and removal of noble metal fission products.