WO2001033575A2

WO2001033575A2 - Process for recycling irradiated fuel

Info

Publication number: WO2001033575A2
Application number: PCT/CA2000/001263
Authority: WO
Inventors: James D. Sullivan; Holly B. Hamilton
Original assignee: Atomic Energy of Canada Ltd AECL
Current assignee: Atomic Energy of Canada Ltd AECL
Priority date: 1999-10-29
Filing date: 2000-10-26
Publication date: 2001-05-10
Anticipated expiration: 2002-04-29
Also published as: AU1121501A; WO2001033575A3

Abstract

A simple proliferation resistant process for recycling irradiated fuel. The entire irradiated fuel matrix is dissolved in a strong acid reagent and the fissile and fertile elements of interest along with a portion of the highly radioactive fission products are precipitated out of solution by raising the pH. By proper selection of the pH, the neutron absorbing rare earth elements do not precipitate with the fissile/fertile elements and the highly radioactive fission products and instead remain in solution. The presence of a significant portion of the fission products in the fissile/fertile mixture renders it highly radioactive and inaccessible without the use of shielded facilities or specialized handling equipment. The irradiated fuel from which the proliferation resistant fuel of the present invention can be recycled is preferably derived from a LWR or a reactor using thorium as a fertile fuel element. The proliferation resistant recycled fuel in accordance with the present invention is preferably recovered for use in a CANDU® type reactor.

Description

PROCESS FOR RECYCLING IRRADIATED FUEL

The present invention relates to a recycling process for irradiated fuel, including uranium and thorium based fuels and a process for producing proliferation-resistant recycled fuel.

BACKGROUND OF THE INVENTION

During irradiation in nuclear fission reactors, various nuclides accumulate in the fuel as by-products of the fission reaction. Some of these nuclides are strong absorbers of thermal neutrons and the buildup of these nuclides is a factor that limits the useful life of the fuel. When this occurs, the fuel must be removed from the reactor and placed in a waste storage facility. Alternatively, the fissile/fertile material can be extracted from the neutron-absorbing nuclides in order that new fuel can be fabricated and further energy produced from the unused fissile/fertile material.

There are a number of known methods for reprocessing uranium based fuels. The most widely used is the PUREX process which is a solvent extraction process using tributyl phosphate in kerosene as the organic phase extracting agent, and a nitric acid aqueous phase. There are also a number of known methods of reprocessing thorium based fuels. The most widely used is the THOREX process which is based on solvent extraction of aqueous solutions of thorium nitrate. Thoria is dissolved in concentrated nitric acid catalysed by fluoride ion followed by solvent extraction. The PUREX and THOREX processes, while effective, involve many extraction and purification stages. Recycling facilities employing these processes are complex and costly to set up and operate.

In conventional chemical reprocessing including the PUREX and THOREX processes, the fissile elements, the fertile elements and the fission products are separated from one another. In the PUREX process, not only is the plutonium extracted as an essentially pure separation product from the uranium, the plutonium is also separated from the radioactive fission products. Similarly, in the THOREX process, the fissile U²³³ is separated from the fertile thorium matrix and from the radioactive fission products.

The result of conventional reprocessing which separates the fissile and fertile species from one another as well as from highly radioactive fission product waste streams is that the fissile material constitutes a proliferation threat. In particular, fissile material that has been separated from the highly radioactive fission products is readily accessible without the use of shielded facilities or specialized handling equipment.

A process for recycling irradiated fuel that use a simple acid dissolution process is disclosed in Canadian Patent No. 1,269,251 which issued on May 22, 1990 to Atomic Energy of Canada Limited. The process uses weak acid which is incapable of dissolving the fuel oxide matrix but is effective to dissolve a portion of the neutron absorbers. The process leaves much of the radioactive fission products in the fuel matrix and accordingly would be proliferation resistant. However, the process is described as being effective to remove only about 50% of many of the heavy neutron absorbing fission products and reduce the number of parasitic neutron absorptions occurring in the fuel during subsequent irradiation by a factor of only about 2. The disclosure at page six indicates that the commercial viability of the process must be determined on an analysis of processing costs and the extended burnup achieved. The process disclosed in Canadian Patent No. 1,269,251 has not been commercially employed and one reason may be the inability of the process to remove substantially all of the heavy neutron absorbing fission products thereby seriously impairing its commercial viability.

One other known method of recycling spent LWR fuel proposes that it be recycled into fuel for use in CANDU® type reactors. This process is known as DUPIC (Direct Use of spent PWR fuel In Candu®). DUPIC is essentially a fuel reconfiguration process that involves disintegration of spent LWR pellets into powder and forming the powder into sintered pellets suitable for use as fuel in a CANDU® type reactor. DUPIC reconfigured fuel contains substantial quantities of non-volatile highly radioactive fission products and accordingly must be handled in hot cells, rendering it proliferation resistant. Separation of the rare earth elements is not required when recycling PWR enriched fuel into a CANDU® type reactor, as the U ³ⁱ concentration remains above natural even in the spent fuel and together with the plutonium, provides a sufficiently high fissile content to overcome the neutron burden of the rare earth fission products.

Accordingly, there is a need for a simple irradiated fuel recycling process which is effective to remove substantially all of the heavy neutron absorbing rare earth elements from t e fissile/fertile fuel elements. There is also a need for such a process that leaves substantial quantities of the highly radioactive fission products in the recycled fuel to render it proliferation resistant.

SUMMARY OF THE INVENTION

The present invention provides a simple process for recycling irradiated fuel, including uranium and thorium based fuels. The entire irradiated fuel matrix is dissolved in a strong acid reagent and the fissile and fertile elements of interest are precipitated out of solution by raising the pH. By proper selection of the pH, the neutron absorbing rare earth elements do not precipitate with the fissile/fertile elements and instead remain in solution. Thus a simple filtration of the dissolved fuel matrix is effective to separate the fissile/fertile elements of interest from the neutron absorbing rare earth elements and provide a source material for recycled fuel suitable for use in a CANDU® type reactor. This process is far less complex than the known solvent extraction processes in current commercial use.

In another aspect, the present invention provides a proliferation resistant process for recycling irradiated fuel. The entire irradiated fuel matrix is dissolved in a strong acid reagent and the fissile and fertile elements of interest along with a portion of the highly radioactive fission products are precipitated out of solution by raising the pH. By proper selection of the pH, the neutron absorbing rare earth elements do not precipitate with the fissile/fertile elements and the highly radioactive fission products and instead remain in solution (along with the balance of the fission products). The presence of a significant portion of the fission products in the fissile/fertile mixture renders it highly radioactive and inaccessible without the use of shielded facilities or specialized handling equipment.

Thus in accordance with the present invention there is provided a process for recycling irradiated fuel comprising fissile and fertile elements and fission products, the process comprising the steps of contacting the fuel with a strong acid reagent having a concentration sufficient to dissolve into solution the fissile and fertile elements and the fission products, raising the pH of the solution to a value effective to precipitate the fissile and fertile elements and a first portion of the fission products containing highly radioactive elements and leave in solution a second portion of the fission products containing heavy neutron absorbing rare earth elements, separating the precipitant from the solution, and recovering the precipitant as a proliferation resistant recycled fuel. In accordance with another aspect of the present invention, there is provided a proliferation resistant nuclear fuel source material recycled from irradiated nuclear fuel comprising fissile and fertile elements and radioactive fission products, said nuclear fuel source material comprising major amounts of fissile and fertile elements and sufficient quantities of other highly radioactive fission products to render the recycled material inaccessible without radioactive shielding facilities and being substantially devoid of neutron-absorbing rare earth fission products.

The irradiated fuel from which the proliferation resistant fuel of the present invention can be recycled is preferably derived from a LWR or a reactor using thorium as a fertile fuel element. The proliferation resistant recycled fuel in accordance with the present invention is preferably recovered for use in a CANDU® type reactor.

BRIEF DESCRIPTION OF THE DRAWINGS

These and other features of the invention will become apparent from the following description in which reference is made to the appended drawings wherein:

FIGURE 1 is a graph showing the amount of various elements in dissolved

SIMFUEL that remain in solution as a function of pH during the separation process of the present invention.

DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENT

The present invention relies upon a process to separate neutron absorbing rare earth elements from irradiated fuel containing fissile, fertile and radioactive fission product species while leaving sufficient radioactive fission products in the recycled material to render it inaccessible without the use of shielded facilities.

The process of the present invention has been applied to SIMFUEL, which is a chemical and physical simulate of irradiated fuel. SIMFUEL is fabricated using unirradiated uranium and thorium and inactive, naturally occurring isotopes of the various fission products typically found in irradiated nuclear fuel. SIMFUEL is radioactive, but the fields are the same as for natural uranium or thorium (very small) and studies can be conducted without shielding. The fission products present in irradiated fuel and their concentrations are a function of the fuel type, reactor type, burnup, cooling time since discharge and other factors. The make-up of SIMFUEL used to validate the process of the present invention is given below in Table 1.

Table 1

Compound Mass

ThO₂ 24.07975

UO₂ 0.28250

ZrO₂ 0.08400

MoO₃ 0.08900

PdO 0.03675

BaCO₃ 0.03750

Y₂ O₃ 0.01000

SrO 0.05575

Rh2O₃ 0.00700

RuO₂ 0.09000

La₂O₃ 0.02825

CeO₂ 0.07600

Nch_O₃ 0.12350

TOTAL 25.00000

The compounds in Table 1 in the quantities listed were milled (ground) with the thoria in a vibratory mill. The resulting powder was pressed on a single action hydraulic press to produce green pellets. The green pellets were sintered at 1700°C for 2 hours in a 40% hydrogen in nitrogen atmosphere. The high density sintered product was examined microscopically to confirm that the desired microstructure was obtained and in particular that the simulated fission product additives were present and evenly distributed. This SIMFUEL product was then subjected to the process of the present invention.

Sintered SIMFUEL pellets were crushed in a custom made device to produce a coarse powder. The coarse powder was further milled in a vibratory mill to produce a fine powder having an average particle size of 2.5 microns. Crushing and milling were done to enhance the rate of dissolution in the subsequent dissolution step, but are not required for the process of the present invention to work.

25 g of the SIMFUEL powder was dissolved in a dissolution reagent similar to that used in the conventional THOREX process. The reagent comprises 13M nitric acid catalysed with 0.04M HF acid and 0.1M aluminum nitrate. Dissolution was carried out in a 500 mL round bottom flask at a temperature of 100°C. Dissolution was complete (with the exception of insoluble components) in 80 minutes. The insoluble components, notably Mo and Ru were removed at this stage by filtration. 4 molar ammonium hydroxide was added to the acidic solution containing the dissolved SIMFUEL and the pH was constantly monitored. Samples were removed at various pH values as the various species precipitated and were centrifuged to separate the precipitant from the liquor. A small sample was pipetted from the clear liquor for analysis. The test results are given in FIGURE 1 which shows the content in solution (in arbitrary units) as a function of pH for the dissolved SIMFUEL. As is evident from FIGURE 1. thorium and uranium are the first to precipitate commencing at a pH of about 3.7 while the other species remain in solution and do not precipitate until much higher pH values. Thus a simple filtration or centrifugation at a predetermined pH value will separate the uranium and thorium from other species remaining in solution, including the rare earth elements (La. Ce, Y, Nd). In addition, some incidental fission products (Pd, Ba, Sr) will remain with the rare earth elements.

Of significance to the present invention is the co-precipitation of Zr and minor amounts of dissolved Rh, Mo and Ru fission products with the thorium and uranium. These elements are among the highly radioactive fission products in irradiated fuel and their presence renders the separated uranium and thorium proliferation resistant. Because these elements are not significant neutron absorbers, their presence will not detrimentally affect the recycled fuel material.

In addition to the rare earth elements included in the SIMFUEL in Table 1, it is recognized that other rare earth elements, including Sm. Eu. Gd provide a substantial portion of the total neutron burden in spent PWR fuel. A further test was conducted using a larger number of fission product elements including these particular rare earth elements of interest. The elements were dissolved directly in the same acidic reagent as used in the previously described test and samples were drawn in the pH range of interest. The further test confirmed that each of these heavy neutron absorbing rare earth elements behave much like Nd as shown in FIGURE 1 and remained in solution beyond the pH values at which uranium and thorium precipitate. Accordingly, the uranium and thorium can be easily separated from the heavy neutron absorbing rare earth elements by a subsequent filtering operation at a selected pH value.

As is evident from FIGURE 1, almost all of the fissile/fertile elements precipitate at pH values from about 3.7 to about 6.5. The precise value of pH at which separation is carried out will depend in large part on the amounts of fissile/fertile elements and rare earth elements desirable or tolerable in the final product. From the data shown in FIGURE 1, separation at a pH value of approximately 5.75 (~6) provides a product substantially devoid of rare earth elements.

The simple precipitation and filtration method used in the process of the present invention is far less complicated than conventional recycling methods including such known solvent extraction processes as PUREX and THOREX. The present invention also provides a process effective to selectively separate the heavy neutron absorbing fission products from the fissile and fertile elements and other highly radioactive fission products thereby providing a proliferation resistant recycled fuel source material. This material can be reformed into recycled fuel pellets suitable for use in CANDU® type nuclear reactors.

It will be understood that while the acid dissolution and precipitation method of the present invention inherently produces a proliferation resistant recycled fuel material by virtue of the co-precipitation of some highly radioactive fission products, the process could be used in combination with other means for separating the highly radioactive fission products to produce a "clean" recycled fuel material.

Claims

The embodiments of the invention in which an exclusive property or privilege is claimed are defined as follows:

1. A process for recycling irradiated fuel comprising fissile and fertile elements and fission products, the process comprising the steps of: a) contacting the fuel with a strong acid reagent having a concentration sufficient to dissolve into solution the fissile and fertile elements and the fission products; b) raising the pH of the solution to a value effective to precipitate the fissile and fertile elements and a first portion of the fission products and leave in solution a second portion of the fission products containing heavy neutron absorbing rare earth elements; c) separating the precipitant from the solution; d) recovering the precipitant as a recycled fuel material.

2. The process of claim 1 wherein the pH value is in the range of approximately 3.7 to 6.5.

3. The process of claim 2 wherein the pH value is approximately 6.

4. The process of claim 1 wherein the strong acid reagent comprises nitric acid catalysed with HF acid.

5. The process of claim 1 wherein the step of raising the pH comprises adding ammonium hydroxide to the solution.

6. The process of claim 1 wherein the step of separating the precipitant from the solution comprises filtration.

7. The process of claim 1 wherein the step of separating the precipitant from the solution comprises centrifuging.

8. The process of claim 1 wherein the fissile element is uranium and the fertile element is thorium.

9. The process of claim 1 wherein the fissile element is plutonium and the fertile element is uranium.

10. The process of claim 1 wherein the highly radioactive fission products include zirconium.

1 1. The process of claim 1 wherein the irradiated fuel is derived from a LWR reactor and the proliferation resistant recycled fuel is recovered for use in a CANDU® type reactor.

12. A recycled nuclear fuel source material produced by the process of any one of claims 1 to 11.

13. A proliferation resistant nuclear fuel source material recycled from irradiated nuclear fuel comprising fissile and fertile elements and radioactive fission products, said nuclear fuel source material comprising major amounts of fissile and fertile elements and sufficient quantities of other highly radioactive fission products to render the recycled material inaccessible without radioactive shielding facilities and being substantially devoid of neutron-absorbing rare earth fission products.

14. The proliferation resistant recycled nuclear fuel source material of claim 9 wherein the fissile element is uranium and the fertile element is thorium.

15. The proliferation resistant recycled nuclear fuel source material of claim 9 wherein the fissile element is plutonium and the fertile element is uranium.

16. The proliferation resistant recycled nuclear fuel source material of claim 9 wherein the highly radioactive fission products include zirconium.

17. A process for recycling irradiated fuel derived from a PWR reactor comprising uranium, thorium and fission products including zirconium and heavy neutron absorbing rare earth elements, the process comprising the steps of: a) contacting the fuel with a strong acid reagent comprising nitric acid catalysed by HF acid having a concentration sufficient to dissolve into solution the uranium, thorium and fission products; b) raising the pH of the solution to a value of approximately 6 to precipitate the uranium, thorium and zirconium and leave in solution the heavy neutron absorbing rare earth elements; c) separating the precipitant from the solution; d) recovering the precipitant as a proliferation resistant recycled fuel for use in a CANDU® type reactor.

18. A process for recycling irradiated fuel derived from a LWR reactor comprising plutonium and uranium and fission products including zirconium and heavy neutron absorbing rare earth elements, the process comprising the steps of: a) contacting the fuel with a strong acid reagent comprising nitric acid catalysed by HF acid having a concentration sufficient to dissolve into solution the plutonium, uranium and the fission products; b) raising the pH of the solution to a value of approximately 6 to precipitate the plutonium, uranium and zirconium and leave in solution the heavy neutron absorbing rare earth elements; c) separating the precipitant from the solution; d) recovering the precipitant as a proliferation resistant recycled fuel for use in a CANDU® type reactor.