GB2556944A

GB2556944A - Use of decay heat from spent nuclear fuel processed by electro-reduction

Info

Publication number: GB2556944A
Application number: GB1620110.5A
Authority: GB
Inventors: Richard Scott Ian
Original assignee: Individual
Current assignee: Individual
Priority date: 2016-11-28
Filing date: 2016-11-28
Publication date: 2018-06-13
Also published as: GB201620110D0

Abstract

A method of reprocessing spent oxide nuclear fuel is described comprising: removing the spent oxide fuel from its cladding; electrolytically reducing the spent fuel to an alloy of uranium and other metallic elements in a molten salt at a temperature above the melting point of uranium; continuing the electrolysis without adding further oxide fuel until the majority of plutonium and other actinides have been reduced to metal and have mixed into the molten alloy; and then transferring the molten alloy to a molten metal/molten salt contacting apparatus where the plutonium and higher actinides are extracted into a further molten salt containing uranium trichloride. The process may be repeated multiple times without changing the molten salt electrolyte, the resulting highly radioactive heat generating salt being used to heat a working fluid which is then used to generate electricity. The further molten salt may also be used as fuel for a molten salt nuclear reactor.

Description

(71) Applicant(s):

Ian Richard Scott

Lambourne House, Lower Binton,

STRATFORD UPON AVON, Warwickshire, CV37 9TQ, United Kingdom (72) Inventor(s):

Ian Richard Scott (74) Agent and/or Address for Service:

Ian Richard Scott

Lambourne House, Lower Binton,

STRATFORD UPON AVON, Warwickshire, CV37 9TQ, United Kingdom (51) INT CL:

G21C 19/44 (2006.01) (56) Documents Cited:

US 3298935 A US 3052611 A

Progress in Natural Science: Material International Vol. 25, December 2015, E-Y Choi et al, Electrochemical processing of spent nuclear fuels: An overview of oxide reduction in pyroprocessing technology, pages 572-582 (58) Field of Search:

INT CL G21C, G21F

Other: WPI; EPODOC; Patent fulltext (54) Title of the Invention: Use of decay heat from spent nuclear fuel processed by electro-reduction Abstract Title: Reprocessing spent oxide type nuclear fuel by electro-reduction (57) A method of reprocessing spent oxide nuclear fuel is described comprising: removing the spent oxide fuel from its cladding; electrolytically reducing the spent fuel to an alloy of uranium and other metallic elements in a molten salt at a temperature above the melting point of uranium; continuing the electrolysis without adding further oxide fuel until the majority of plutonium and other actinides have been reduced to metal and have mixed into the molten alloy; and then transferring the molten alloy to a molten metal/molten salt contacting apparatus where the plutonium and higher actinides are extracted into a further molten salt containing uranium trichloride. The process may be repeated multiple times without changing the molten salt electrolyte, the resulting highly radioactive heat generating salt being used to heat a working fluid which is then used to generate electricity. The further molten salt may also be used as fuel for a molten salt nuclear reactor.

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Use of decay heat from spent nuclear fuel processed by electro-reduction

Background

Spent fuel from nuclear reactors continues to produce heat for many years after the nuclear fission process has ended. There have been many suggestions as to how to use this decay heat for productive purposes but all have failed to have any commercial relevance due to a combination of factors.

• Spent fuel in the form it emerges from the reactor rapidly decays to a point after a few years where the heat output is of the order of lkW per tonne of fuel making any meaningful power generator impractically large.

• Spent fuel typically is in a form where it cannot be exposed to high temperatures without risk of leakage, limiting the temperature of whatever medium is used to transfer to heat.

• Spent fuel emits high levels of radiation so any installation must be extensively screened to protect people and the environment • Spent fuel contains plutonium so any installation using it becomes subject to IAEA nuclear safeguards standards which impose an impractically high cost on the relatively small amounts of electricity generated.

• Chemical processing of spent fuel to separate the small quantity of heat generating fission products from the large volume of non heat producing uranium is expensive and typically involves using a process that can produce weapons usable plutonium.

The combination of these factors have prevented any attempt to use the decay heat from spent nuclear fuel on a commercial scale.

Description of the invention

A novel method of reprocessing spent uranium oxide fuel has been devised where the oxide pellets are electrochemically reduced in a suitable electrolyte operating above the melting point of uranium. The high temperature is the key innovation as it prevents formation of layers of metallic uranium in the outer regions of the oxide pellet which otherwise prevent the oxide at the centre of the pellet from being reduced to the metal. Previous attempts to use electrolytic reduction on spent nuclear fuel have been broadly unsuccessful due to this problem of incomplete oxidation of solid oxides at lower temperatures as reviewed in Progress in Natural Science. Materials International,vol25, 572-582 (2015) see end of section

2.1

The electrolyte can be any molten salt that allows significant levels of metal oxides to be dissolved in the salt. Calcium chloride, containing small amounts of calcium oxide is suitable, but salts or mixtures of salts such as barium chloride, strontium chloride, sodium chloride, potassium chloride, caesium chloride, rubidium chloride are also suitable and may be preferred in certain circumstance. Strontium and cesium salts for example are already present in spent nuclear fuel as fission products and therefore do not introduce new elements to the system. Fluoride salts may be used instead of chloride salts.

This reprocessing method is radically simpler and therefore cheaper than aqueous based reprocessing such as the PUREX process or other molten salt based processes which require electrochemical transfer of uranium from one electrode to another as a key part of the process (eg Argonne National Lab reprocessing of EBR2 metallic reactor fuel).

An exemplary apparatus is shown in figure 1. Spent nuclear fuel pellets are fed into the apparatus continually and reduced to uranium metal by passage of electric current from anode to cathode. Initially the molten metal alloy produced contains uranium, neptunium and the noble metals but when sufficient molten alloy has accumulated, addition of fuel pellets is stopped and electrolysis continued. A large portion of the plutonium and other actinides and a fraction of the lanthanides are reduced to metal at this time and mix into the molten alloy.

This alloy is drained from the electro-reduction cell through a counter flow column where it is contacted with a molten salt containing uranium trichloride which extracts the plutonium and higher actinides, and a fraction of the lanthanides from the molten alloy which is then stored for later use. The molten salt from this process is suitable for use in a molten salt fuelled nuclear reactor such as that described in GB2508537.

The process is then repeated many times with fresh batches of spent oxide fuel (though metallic fuel may be added or used instead of oxide fuel) until the heat production or chemical change from the accumulation of fission products in the electrolyte reaches a level where replacement of the electrolyte is required. At this point addition of fuel pellets is stopped and a more exhaustive electrolysis carried out so that essentially all the remaining actinides and the majority of the lanthanides are reduced to metal. This alloy is passed down a counter flow apparatus contacting a molten salt which may be uranium trichloride or another suitable salt so as to extract much of the lanthanides but substantially none of the actinides into the salt and to transfer substantially all the uranium from the salt into the alloy. The salt is either stored for recovery of the lanthanides or returned to the electrolysis apparatus to mix with the spent electrolyte. The alloy containing the actinides, including the uranium transferred from the uranium trichloride is stored to be added back to the apparatus with a fresh batch of electrolyte. The spent electrolyte is drained from the electrolysis cell for disposal, or as set out in an embodiment of the current invention to be used as a useful heat source.

This process produces three outputs • a gas stream of the noble gas fission products plus oxygen or carbon oxides (depending on whether an inert of graphite electrode is used), • a metal alloy comprising uranium, other actinides, some lanthanides and all the noble and semi-noble metal fission products • a molten salt stream containing the majority of the fission products but substantially no actinides

The process has a number of advantages over other spent fuel reprocessing methods.

• It is very simple and capable of high throughput and hence low cost. This is driven by a number of factors but of especial importance is the tolerance of the electro-reduction cell to high current densities which are an order of magnitude higher than are possible in alternative electro-refining cells where uranium is transferred between electrodes.

• It is relatively poor at separating plutonium from other species, especially lanthanides. As such it cannot be reconfigured to produce plutonium of a purity that can be used in a nuclear weapon. It is therefore substantially more proliferation resistant than alternative reprocessing methods.

• It produces an actinide free waste stream containing the majority of the fission products except for the noble metals including technetium-99. This is important for two reasons. The absence of actinides means it is not subject to IAEA nuclear safeguard regulations but only to the safety and security requirements for storage of nuclear waste. The absence of Tc-99 and actinides means that its radioactivity falls rapidly when the heat producing caesium and strontium have decayed making its long term disposal much simpler.

The spent electrolyte has a high heat production capacity. Depending on the amount of spent nuclear fuel processed through it, this may be as high as 100-500 times the heat production per tonne of the input spent fuel. Depending on the time since the spent fuel was removed from the reactor this could be of the order of lOOkW per tonne.

The spent electrolyte can optionally be rendered substantially non corrosive to standard steels by addition of small amounts of metallic zirconium or other reactive metal and packaged in steel (preferably stainless steel) containers which can be welded closed with a gas space inside to allow for expansion of the salt.

This salt has a heat generation capacity high enough to be of practical value but has a high emission of radiation. For optimum power production, its temperature must be around 600C or more so that it can power a standard high efficiency superheated steam cycle, though other options for conversion of the heat to electric power such as Stirling engines, thermoelectric apparatus, Brayton cycle engines etc may be preferred for certain purposes. It must also be effectively screened against its emissions of gamma radiation.

A suitable apparatus to utilise this heat source is shown in figure 2. It is a tank containing a heat storage medium in which the heat generating spent electrolyte salt containers are immersed. Particularly useful media are “solar salt” which is a mixture of sodium and potassium nitrates or alternative salts such as Hitec salt. The molten solar salt can be bulked out with cheap solid materials such as crushed rock. Other media such as high temperature oils are also suitable though radiolysis by the gamma radiation from the heat source would limit their practical life.

Heat exchangers around the periphery of the tank allow the heat in the solar salt to be transferred to a working fluid for the turbo-generator set. Flow of solar salt in the tank is ideally via natural convection resulting in no moving parts internal to the tank though pumped flow is an option. The large volume of the solar salt provides an effective radiation screen but also gives the apparatus a high thermal inertia so that if the heat source emits heat at, say, lOOkW continually, the turbogenerator can withdraw heat at double that rate for half of the day. This power output variation is especially important when an apparatus such as this is used in remote communities where other methods of production/demand management are difficult to achieve.

Waste heat from the turbo-generator can be discarded or used for heating if a distribution system exists.

The coolant in the tank can also be used directly as a source of process heat.

Claims

1) A process for separating spent nuclear oxide type fuel into a new nuclear fuel and waste products comprising • removing spent oxide fuel from its metal cladding, • reducing the spent fuel to an alloy of uranium with other metallic elements by electrolysis in a molten salt in the presence of the oxide fuel at a temperature above the melting point of uranium, • continuing the electrolysis process without adding further oxide fuel until the majority of the plutonium has been reduced to metal, • transferring the molten uranium alloy to a molten metal/molten salt contacting apparatus where the plutonium is transferred into the molten salt

2) The process of claim 1 where the molten salt in the contacting apparatus includes uranium trichloride

3) The process of claim 1 where the steps are repeated multiple times without changing the molten salt electrolyte and a final electrolysis in the absence of added oxide fuel is used to reduce essentially all the actinides in the electrolyte to a molten alloy after which the molten salt electrolyte is removed for storage or further use

4) The process of claim 1 where the steps are repeated multiple times without changing the molten salt electrolyte and a final electrolysis in the absence of added oxide fuel is used to reduce essentially all the actinides in the electrolyte to a molten alloy after which the molten salt electrolyte is removed for storage or further use and fresh molten salt electrolyte added together with more oxide spent fuel, this process resulting in accumulation of the lanthanide fission products in the electrolyte

5) The process of claim 1 where the steps are repeated multiple times without changing the molten salt electrolyte and a further electrolysis in the absence of added oxide fuel is used to reduce essentially all the actinides in the electrolyte to a molten alloy after which the molten salt electrolyte is removed for storage or further use and the molten alloy passed through a molten metal/molten salt contactor to transfer some non actinide metals from the molten metal, the remaining metal alloy being returned to the electrolysis cell together with fresh molten salt electrolyte and spent oxide fuel for continued use thereby preventing accumulation of lanthanides in the electrolysis cell while avoiding actinides entering the waste streams from the process.

6) The process of claim 1 where the molten salt is calcium chloride, strontium chloride, barium chloride, sodium chloride, potassium chloride, caesium chloride, rubidium chloride or a mixture thereof

7) The process of claim 1 where the electrolysis apparatus is lined with or constructed from tantalum or tungsten metal optionally coated with yttrium oxide

8) The process of claim 1 where the anode is graphite and carbon oxides are generated consuming the anode

9) The process of claim 1 where the anode is iridium and oxygen is generated without consuming the anode

10) The process of claim 1 where the steps are repeated multiple times without changing the molten salt electrolyte and the resulting highly radioactive heat generating salt is used in an apparatus transferring the heat to a working fluid which is then used to generate electricity or provide process heat

11) The process of claim 1 where the steps are repeated multiple times without changing the molten salt electrolyte and the resulting highly radioactive heat generating salt is used in an apparatus transferring the heat to a working fluid which is then used to generate electricity or provide process heat where the molten salt electrolyte is enclosed in leak proof thermally conductive containers of stainless steel or other material resistant to molten salt attack and immersed in a tank of coolant which is capable of storing the heat output of the heat generating salt for a period of hours allowing temporally flexible use of the heat for electricity generation or process heat.

12) The process of claim 1 where the steps are repeated multiple times without changing the molten salt electrolyte and the resulting highly radioactive heat generating salt is used in an apparatus transferring the heat to a working fluid which is then used to generate electricity or provide process heat where the molten salt electrolyte is enclosed in leak proof thermally conductive containers of stainless steel or other material resistant to molten salt attack and immersed in a tank of coolant where the coolant is a molten salt of low melting point such as a mixture of nitrates and optionally nitrites

13) The process of claim 1 where the steps are repeated multiple times without changing the molten salt electrolyte and the resulting highly radioactive heat generating salt is used in an apparatus transferring the heat to a working fluid which is then used to generate electricity or provide process heat where the molten salt electrolyte is enclosed in leak proof thermally conductive containers of stainless steel or other material resistant to molten salt attack and immersed in a tank of coolant where the thickness of the layer of coolant surrounding the heat generating salt containers is sufficient to effectively screen the radiation from said containers.

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