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INTRODUCTION

In a PWR or BWR type nuclear reactor, the fuel elements are removed once the fuel has finished its burn cycle. In this case, every 18-24 months, depending on the regulator technical specifications, the fuel elements are transferred to temporary storage pools.

These structures are made of stainless steel and covered with concrete to keep everything inside: spent fuel elements, LPRM string or instrumentation tubes, neutron sources used in the start-up of the plant (e.g. Sb-Be sources), boron steel poisons, stellite bearings and shafts, water channels, irradiated guide tubes and baskets with damaged elements.

There are two general processes for the separation and reprocessing of fuel elements present in irradiated nuclear fuels. These processes are hydrometallurgical separation processes (among which, the PUREX process is the most used) and pyrometallurgical processes.

The PUREX process is a liquid-liquid extraction process with tributyl phosphate used as an organic solvent; it treats irradiated fuel to reuse the uranium and plutonium.

Once the life cycle of the nuclear fuel is finished, there are different ways of treating the elements that contain the burned fuel. The fuel elements in its structure, normally zirconium alloys, contain the fission fragments generated in nuclear reactions, transuranium elements and especially uranium and plutonium, in a sealed form.

This process takes place in the closed cycle case. When discharged from the reactor core, the spent fuel has a composition of 96% uranium, 1% plutonium and 3% minority actinides and fission products (by mass).

These structures, called fuel elements, contain a determined number of rods (17 x 17-21) in a PWR reactor and cells with 8 x 8 or 9x 9 rods in a BWR type reactor; these are deposited in temporary storage pools (SFP).

Their function is to keep the burned fuel in a state of dormancy so that its radioactivity decreases, reducing the heat activity of the element so that in the dismantling stage it can be handled with operational shielding to manage the residue with the minimum risk applied.

Another primary function is to ensure the subcriticality of the distribution of burned fuel elements at all times.

Nuclear fuel reprocessing purex

Image credits: https://www.nrc.gov/

The other alternative to reprocessing is to dispose of the waste from temporary storage pools before a cooling period, where waste reduces activity and reduces its calorific value, in deep geological repositories via canisters approved by the regulator.

These chemical species, originating in the fission process, elevate the radioactive activity of the fuel element to values that require the adoption of serious, regulated radiation protection measures.

RADIONUCLIDES OF LONG-TERM INTEREST IN IRRADIATED ELEMENTS
ACTIVATED PRODUCTS FISSION PRODUCTS TRANSURANIUM ELEMENTS
14C 5730a 3H 12.33a 235U 7.030.108a
60Co 5.27a 85Kr 10.7a 238U 4.468.109a
55Fe 2.7a 99Tc 2.14.105a 238Pu 87.74a
59Ni 7.5 104a 129I 1.6.107a 239Pu 2.41.104a
134Cs 2.06a 240Pu 6.57.103a
137Cs 30.17a 241Am 433a
90Sr-90Y 28.8a(64.1h) 243Am 7.37.103a
243Cm 28.5a
237Np 2.14.106a
244Cm 18.11a

There is a different method for treating fuel cells: Reprocessing.

Specifically, the PUREX process, developed in France, recovers fissile material with non-recoverable material undergoing subsequent vitrification.

The following table shows the content of a fuel element before loading and after unloading in a reactor, when the element is considered spent.

The high formation of fission and transuranium fragments that did not exist at the beginning can be seen.

The table shows the elements obtained in the discharge and evolution of the activity at 10 and 1000 years per half a ton of uranium loaded in the fuel element of a PWR reactor.

Load (g) Discharge (g) Amount at 10 years (g) Amount at 1000 years (g)
235U 1.43.104 3.24.103 3.24.103 3.24.103
238U 4.27.105 4.17.105 4.17.105 4.17.105
238Pu 0 6.04.101 5.91.101 2.92.10-1
239Pu 0 1.93.103 1.93.103 1.88.103
240Pu 0 9.24.102 9.27.102 8.43.102
241Am 0 8.31 1.54.102 8.32.101
243Am 0 3.58.101 3.58.101 3.27.101
243Cm 0 3.22.10-2 2.59.10-2 1.27.10-11
244Cm 0 1.03.101 7.02 2.46.10-16
237Np 0 1.94.102 2.102 5.09.102

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PROCESS OF RECEPTION AND MECHANICAL TREATMENT

First, the fuel elements are received in the temporary storage pools, each in their coordinate positions.

Using a shearing process, the fuel elements are cut and subjected to an oxidizing acid medium such as nitric acid. The fuel element rods are mechanically cut into small pieces with the active material inside so that the acid can penetrate and attack them.

The nitric acid dissolves the oxides present in the fuel elements, extracting a highly active solution. Nitric acid is used because its attack produces little solid residue and keeps most active ions in solution in high oxidation states due to its oxidative capacity.

In particular, it keeps the uranium in the oxidation state (VI) and the plutonium in an optimal oxidation state (IV) to start the extraction process.

This allows the structural elements to be separated from the solutions containing uranium and plutonium fission fragments.

The zirconium alloy structural elements are stored and treated as half-life waste. These structures activated by radiation from the fission fragments that were previously located inside are stored in specific silos.

ACID DIGESTION PROCESS

The solubilized oxides and gaseous effluents are present in the formation of the nitric solution. Its treatment follows two entirely different routes.

For gases generated inside the zirconium alloy sheaths, the gaseous effluent is first evacuated to the iodine and krypton-xenon isotope treatment line. Iodine isotopes can be retained by silver nitrate filters, forming silver iodide, a solid residue that is treated as a high activity residue.

The Krypton-Xenon isotopes group is retained by activated carbon filters and removed to gas retention tanks. The remaining zero or low activity gases are treated conventionally.

The solution formed by the nitric acid attack can produce precipitates that are separated and treated as insoluble solid waste; for example, zirconium oxides and aluminum oxides.

Subsequently, the solution is clarified via a coagulation-flocculation process, forming an optimal solution to be treated by extraction processes.

The solution containing the uranium and plutonium nitrates is treated with sodium nitrite to maintain the plutonium in the IV state.

The oxidation state of plutonium under strongly oxidizing conditions can reach up to VI, hindering extraction with the solvent.

TRIBUTYL PHOSPHATE EXTRACTION PROCESS

The extraction process is carried out with tributyl phosphate in different cycles, producing two phases. Apart from heavy metals, the non-organic phase contains americium, curium and fission fragments. This solution is evaporated paying special attention to the steam due to its possible significant tritium content.

The organic phase contains uranium and plutonium. In the form of uranium (VI) and plutonium (IV) nitrate, these salts are easily and effectively extracted from the aqueous solution by the solvent tributyl phosphate, with the fission fragments being extracted by the organic phase to a lesser extent.

Tributyl phosphate is stable as an extractant in highly concentrated nitric acid solutions. It is used due to its low volatility and high flash point, which are important for the safety of the industrial process in question.

The nitric acid solution enters the extraction tower in such a way that the extract solution of tributyl phosphate and kerosene circulates counter currently, to reduce the density of the organic mixture.

The nitric acid solutions of plutonium (IV) and uranium (VI) go into the organic phase. In the upper part of the column, the organic solution is washed with nitric acid acting as a saline agent.

This step is important, as the increase in salinity favors the fission products going into aqueous solution, while preventing the uranium (VI) and plutonium (IV) nitrates from doing so.

The aqueous effluent from the extraction phase contains practically all fission fragments and small amounts of uranium (VI) and plutonium (IV) nitrate.

REDUCTION PROCESS

The organic phase, containing uranium and plutonium, is treated in a second stage by a reducing solution circulating in countercurrent. This solution consists of ferrous sulfamate and a certain concentration of saline agent with the aim of preparing the separation of uranium and plutonium.

The ferrous ion acts as a reducer, especially by converting plutonium (IV) to plutonium (III), which is transferred to the aqueous phase and separated from uranium. The sulfamate acts on the nitrite so as not to impede the reduction of plutonium.

This reducing solution descends through the column, while fresh tributyl phosphate ascends to ensure that no uranium fraction goes into the aqueous solution. At this point, there is a stream containing the plutonium (III) in the aqueous phase and the uranium separated in the organic phase with tributyl phosphate.

The organic phase carrying the uranium is transferred to a column where it again interacts with a stream of dilute nitric acid to convert it into uranium (VI) nitrate.

The organic tributyl phosphate solvent is recovered by distillation and reused in the process as regenerated tributyl phosphate. Once the uranium and plutonium streams have separated, the two chemical species are purified.

The uranium solution is again extracted with tributyl phosphate and then washed with reducing solution to extract the purified uranium in the aqueous phase.

The plutonium (III) in the aqueous phase is converted to plutonium (IV) with a solution of sodium nitrate and nitric acid. After this step has been carried out, it is extracted with tributyl phosphate.

Once the extraction has taken place in the organic phase, and therefore the degree of purification is increased, it is re-extracted with hydroxylamine sulfate as a reducing agent to convert plutonium (IV) back into plutonium (III). This gives a highly purified solution of plutonium nitrate.

Another version of the PUREX process extracts the plutonium with an amine. The plutonium is purified from the oxidation state (III) by oxidizing to plutonium (IV) by treatment with sodium nitrate and nitric acid.

Subsequently, it is extracted by trilaurylamine (tertiary amine diluted with diethylbenzene). This organic phase is washed with dilute nitric acid and separated with acetic acid containing a certain concentration of nitric acid.

In the above two cases, plutonium with a high degree of chemical purity is obtained. In this phase, it is mandatory to indicate the constant control carried out throughout the plutonium reprocessing to ensure the subcriticality of the system.

OBTAINING URANIUM

The results of the Purex process are concentrated solutions of uranyl nitrate, plutonium nitrate and nitrates of fission products. Uranyl nitrate is converted to uranium trioxide by precipitation with sodium hydroxide and calcination to obtain UO3.

Uranium trioxide can be treated with fluorine to produce uranium hexafluoride, which is recycled in enrichment plants. Subsequent treatment with HF gas produces UF6.

From the following reactions:

UF6 + Fe -> UF4 + FeF2

It is also possible to obtain UF4. By fluorination, both types of fluorides can be produced. Normally, UF6 is used for enrichment processes.

UO2 can be obtained by reducing UO3 and can be reused as fuel after being properly sintered and with the appropriate proportion of 235U.

OBTAINING PLUTONIUM

Plutonium nitrate is converted into the ceramic plutonium dioxide for recycling in thermal or fast breeder reactors. Plutonium (VI) nitrate is reduced and precipitated with oxalic acid to obtain PuO2 . This plutonium dioxide can be used as fuel in breeder reactors.

TREATMENT OF FISSION PRODUCTS

The amount of fission fragments produced inside the sheaths depends strongly on the degree of enrichment of the fuel, the type of fuel and the type of reactor, among other parameters.

In a PWR type reactor with an enrichment rate of 3.25% UO2 in 235U for each ton of fuel burned, the following is obtained:

Quantity of main fission products (g/Ton) Fission products Amount
Krypton 370
Strontium 880
Yttrium 470
Zirconium 3650
Niobium 13
Molybdenum 3450
Ruthenium 2250
Rhodium 390
Palladium 1300
Tellurium 560
Iodine 270
Xenon 5400
Cesium 2700
Barium 1400
Lanthanum 1250
Cerium 2850
Praseodymium 1200
Neodymium 3900
Promethium 110
Others 2587
Total 35000
Quantity of actinides (g/Ton) Actinides Amount
Neptunium 760
Plutonium 9100
Americium 150
Curium 35
Uranium 955000
Radioactive activity [ β ; ɣ ] 4500 Ci/Kg

The solutions containing fission fragments are evaporated and subsequently calcined to obtain the corresponding oxides of the actinide elements and fission fragments.

Once the oxides are obtained, they are combined with silicon dioxide at high temperatures to form materials which are mixture of silicates from the fission fragments, similar to glass.

Silicates are chemical species with a high melting point, which resist high temperatures well and are not appreciably soluble. They are optimal candidates for a deep geological repository.

Vitrification is considered the most reliable way to dispose of highly active radioactive waste in a deep geological repository. There are fundamental differences in terms of the needs for reprocessing the different fuel cycles, due to the different fuel characteristics.

The irradiated fuel from fast breeder reactors has a higher content of plutonium and fission products than fuel from thermal neutron reactors. Higher concentrations of plutonium cause criticality complications that require a different design.

CONCLUSIONS

The PUREX process was designed to reprocess spent fuel and recycle uranium oxide and plutonium oxide. The knowledge developed and nuclear chemistry applied will have to be used in the future for different dismantling processes.

Aspects such as unitary operations of singular evaporation with concentrated solutions of highly active fission fragments with prior calcination, vitrification technology of highly active waste and, in particular, the specific radiological control of these processes, which arise in one way or another in dismantling hot zones, make this process an element of knowledge to keep in mind for future steps in the nuclear industry.

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Sergio Tuset is the CEO of Condorchem Envitech, with over 20 years’ experience in management of industrial companies.

Specially focused on environmental projects for customers, recognized specialist in conceptual engineering applied in wastewater, liquid &solid wastes treatment and air pollution treatment.

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