**4. Results and discussion**

The spectra of twenty five surface soil samples surrounded the abandoned Uranium mine hole have been analyzed. The specific activity of 238U, 232Th, 40K and Radium equivalent activity (Raeq) are given in Table 3. The specific activity (Bq/kg) varied from 37.31 to 1112.47 (mean = 268.16), 0.28 to 18.57 (mean = 6.68) and 132.25 to 678.33 (mean = 277.49) for 238U, 232Th and 40K respectively.

The obtained results are comparable to the worldwide average recommended by UNSCEAR which are 30, 35 and 400 Bq/kg for 238U, 232Th and 40K respectively (UNSCEAR, 2000). It was found that all values of 238U specific activities are higher than the worldwide average whereas those of 232Th are less than it. For 40K, it is clear that the specific activities, with the exception of five samples, are found to be less than worldwide average.

Obviously demonstrate that the minimum and maximum specific activity values of 238U are least by factor of 4 and 37 higher than the corresponding values obtained worldwide average. The large variation between the specific activities obtained for 238U and other two radionuclides can be easily ascribed to the high content of uranium in the neglected waste of drilling and exploration operations on the surface soil surrounding the mine. The contour maps (radiological maps) of the activity distribution of 238 U, 232Th and 40K in the study area are shown in Figures 5, 6 and 7. From Figure 5, we can observe three regions with a highest specific activity values of 238U situated at northeast, east and south-west portions of the hole mine. In contrast, Figure 7 indicates that high concentrations of 40K occupies the same positions of 238U while for 232Th there are no placements have activities require attention as shown in Figure 6.

The calculated Raeq values for all samples were also presented in Table 3. It may be seen that Raeq oscillates between 52.727 and 1189.845 with an average of 299.09 Bq/kg. It is observed that the values of Raeq in twenty one samples were less than the acceptable safe limit of 370 Bq/kg (OECD, 1979; UNSCEAR, 1982; UNSCEAR, 1988). As shown in Table 3 there are four values greater than worldwide average. As a rule, the matter whose Raeq exceeds 370 Bq/kg is discouraged (Beretka and Mathew, 1985). Figure 8 demonstrates the distribution of Raeq and it appears three positions have highest values.

The calculated absorbed dose rate of samples was listed in Table 3. As shown in Table 3, the values ranged from 25.02 to 553.01 with an average value of 139.61 nG/h which is nine fold higher than the world average of 15 nG/h recommended by UNSCEAR (UNSCEAR, 2000). It can be seen that all values were much higher than the world average.

There is another hazard index called internal hazard index (Hin), which is given by

259 A

The values of the index must be less than the unity in order to keep the radiation hazard to be insignificant unity corresponds to the upper limit of radiation equivalent activity

The spectra of twenty five surface soil samples surrounded the abandoned Uranium mine hole have been analyzed. The specific activity of 238U, 232Th, 40K and Radium equivalent activity (Raeq) are given in Table 3. The specific activity (Bq/kg) varied from 37.31 to 1112.47 (mean = 268.16), 0.28 to 18.57 (mean = 6.68) and 132.25 to 678.33 (mean = 277.49) for 238U,

The obtained results are comparable to the worldwide average recommended by UNSCEAR which are 30, 35 and 400 Bq/kg for 238U, 232Th and 40K respectively (UNSCEAR, 2000). It was found that all values of 238U specific activities are higher than the worldwide average whereas those of 232Th are less than it. For 40K, it is clear that the specific activities, with the exception of five samples, are found to be less than worldwide

Obviously demonstrate that the minimum and maximum specific activity values of 238U are least by factor of 4 and 37 higher than the corresponding values obtained worldwide average. The large variation between the specific activities obtained for 238U and other two radionuclides can be easily ascribed to the high content of uranium in the neglected waste of drilling and exploration operations on the surface soil surrounding the mine. The contour maps (radiological maps) of the activity distribution of 238 U, 232Th and 40K in the study area are shown in Figures 5, 6 and 7. From Figure 5, we can observe three regions with a highest specific activity values of 238U situated at northeast, east and south-west portions of the hole mine. In contrast, Figure 7 indicates that high concentrations of 40K occupies the same positions of 238U while for 232Th there are no placements have activities require attention as

The calculated Raeq values for all samples were also presented in Table 3. It may be seen that Raeq oscillates between 52.727 and 1189.845 with an average of 299.09 Bq/kg. It is observed that the values of Raeq in twenty one samples were less than the acceptable safe limit of 370 Bq/kg (OECD, 1979; UNSCEAR, 1982; UNSCEAR, 1988). As shown in Table 3 there are four values greater than worldwide average. As a rule, the matter whose Raeq exceeds 370 Bq/kg is discouraged (Beretka and Mathew, 1985). Figure 8 demonstrates the distribution of Raeq

The calculated absorbed dose rate of samples was listed in Table 3. As shown in Table 3, the values ranged from 25.02 to 553.01 with an average value of 139.61 nG/h which is nine fold higher than the world average of 15 nG/h recommended by UNSCEAR (UNSCEAR, 2000). It can be seen that all values were much higher than the world

185 <sup>A</sup> <sup>H</sup> <sup>U</sup> Th <sup>K</sup> in

4810 A

equation.

(370 Bq/kg).

average.

shown in Figure 6.

average.

and it appears three positions have highest values.

**4. Results and discussion** 

232Th and 40K respectively.


Table 3. Specific activity, Radium equivalent activity and absorbed dose rate of soil samples.

Natural Occurring Radionuclide Materials 17

Fig. 7. Specific activity distribution of 40K.

Fig. 8. Distribution of Radium equivalent in surface soil around mine.

Fig. 5. Specific activity distribution of 238 U.

Fig. 6. Specific activity distribution of 232Th.

Fig. 5. Specific activity distribution of 238 U.

Fig. 6. Specific activity distribution of 232Th.

Fig. 7. Specific activity distribution of 40K.

Fig. 8. Distribution of Radium equivalent in surface soil around mine.

Natural Occurring Radionuclide Materials 19

In addition, the calculated values of hazard index for the soil samples were ranged from 0.142 to 3.216 with an average value of 0.808 and from 0.243 to 6.222 with an average value

Out of 25 positions, 4 for Hex and 13 for Hin, have values very higher than unity. Since these values are dispersed randomly within a limited area around the min hole, therefore, according to the report of European Commission in Radiation Protection, the area study is not safe and

The surface soil layer around the uranium mine hole has uranium activities greater than worldwide average; this can mainly due to the waste of drilling and exploration left on the

The thorium activities were within normal level in the studied area. Generally, potassium

The absorbed dose rates of studied area are higher than the criterion limit of gamma

Finally, from the radiation protection point of view the studied area is considered to be not safe inhabitants because the values of both internal and external hazard indexes associated with the samples are higher than unity. Thus, the human inside the area are supposed to

Abusini M. (2007). Determination of Uranium, Thorium and Potassium Activity Concentrations in Soil Cores in Araba valley, Jordan, *Radiation Protection Dosimetry*, 128: 213-216. Benenson W. (2002). *Hand book of physics*. Fourth Edition. Spriger –Velarg, New York, INC. Beretka J. and Mathew P. J. (1985). Natural Radioactivity of Australian Building Materials,

Cottingh W. N. and Greenwood D. A. (2001). *An introduction to Nuclear physics*, second

Dipak G., Argha D., Sukumar B., Rosalima S., Kanchan K. P. (2008). Measurement of

European Commission (1999). *Radiological Protection Principle concerning the Natural* 

Harb S. (2004). *On the Human Radiation Exposure as Derived From the Analysis of Natural and* 

Henery S. and John R. A. (1972). *Introduction to Atomic and Nuclear Physics*, Fifth edition,

IAEA (2004). *Soil Sampling for Environmental Contaminants*, International Atomic Energy

Ibrahim N. (1999). Natural Activities of 238U, 232Th and 40K in Building Materials, *J. Environ.* 

ICRP (1993). International Commission on Radiological Protection, publication 65,Annals of

*Man-Made Radionuclides in Soil*, Ph.D. Thesis, Hannover University.

Natural Radioactivity in Chemical Fertilizer and Agriculture Soil: Evidence of High

*Radioactivity of Building Materials*. Radiation Protection 112, European Commission

Industrial Wastes and by Products, *Health Physics*, 48: 87-95.

Alpha Activity, *Environmental Geochemistry and Health*, 30: 79-86.

edition, University of Cambridge, United kingdom.

Gerrado C. Maxino (1974). Radioactive Potassium, *The nucleous*, 12: 4-8.

Holt, Rineart and Winston INC.

*Radioact.*, 43: 255-258.

Agency, TECDOC-1415, Vienna, Austria.

the ICRP 23(2), Pergamon press, Oxford.

of 1.533 for external (Hex) and internal (Hin) respectively as mentioned in Table 2.

posing significant radiological threat to the population (European Commission, 1999).

radionuclide in soil samples was in the range of worldwide average.

**5. Conclusion** 

**6. References** 

surface layer of soil surrounding the mine.

acquire radiological complication.

,Brussels.

radiation dose rate with an average of nine times.

The annual effective dose values were calculated and listed in Table 4. They were found to be in the range 0.123 to 2.713 mSv/y with an average value 0.68 mSv/y and from 0.031 to 0.6780 with an average value of 0.17 mSv/y for indoor and outdoor annual effective dose respectively. In general and as shown in Table 4, for indoor annual effective dose, It is important here to notice that there are fourteen sample have values higher than the word average whereas, the values of the rest samples are close or slightly above of the world average value of soil. In other words, all values of outdoor annual effective dose were below the worldwide average.


Table 4. Annual effective dose and hazard indexes of soil samples.

The international commission on Radiological Protection (ICRP) has recommended the annual effective dose equivalent limit of 1 mSv/y for the individual members of the public and 20 mSv/y for the radiation workers (ICRP, 1993). The worldwide average annual effective dose is approximately 0.5 mSv and the results for individual countries being generally within the 0.3 to 0.6 mSv range (UNSCEAR, 2000).

In addition, the calculated values of hazard index for the soil samples were ranged from 0.142 to 3.216 with an average value of 0.808 and from 0.243 to 6.222 with an average value of 1.533 for external (Hex) and internal (Hin) respectively as mentioned in Table 2.

Out of 25 positions, 4 for Hex and 13 for Hin, have values very higher than unity. Since these values are dispersed randomly within a limited area around the min hole, therefore, according to the report of European Commission in Radiation Protection, the area study is not safe and posing significant radiological threat to the population (European Commission, 1999).

#### **5. Conclusion**

18 Radioisotopes – Applications in Physical Sciences

The annual effective dose values were calculated and listed in Table 4. They were found to be in the range 0.123 to 2.713 mSv/y with an average value 0.68 mSv/y and from 0.031 to 0.6780 with an average value of 0.17 mSv/y for indoor and outdoor annual effective dose respectively. In general and as shown in Table 4, for indoor annual effective dose, It is important here to notice that there are fourteen sample have values higher than the word average whereas, the values of the rest samples are close or slightly above of the world average value of soil. In other words, all values of outdoor annual effective dose were below

Sample code Annual dose (mSv) Hazard index

Table 4. Annual effective dose and hazard indexes of soil samples.

generally within the 0.3 to 0.6 mSv range (UNSCEAR, 2000).

The international commission on Radiological Protection (ICRP) has recommended the annual effective dose equivalent limit of 1 mSv/y for the individual members of the public and 20 mSv/y for the radiation workers (ICRP, 1993). The worldwide average annual effective dose is approximately 0.5 mSv and the results for individual countries being

S11 0.22 0.05 0.26 0.45 S12 0.17 0.04 0.20 0.36 S13 0.54 0.13 0.64 1.22 S14 0.12 0.03 0.14 0.25 S15 1.19 0.29 1.41 2.71 S21 0.35 0.08 0.42 0.79 S22 1.60 0.40 1.89 3.63 S23 0.81 0.20 0.95 1.80 S24 0.63 0.15 0.74 1.42 S25 0.12 0.03 0.14 0.24 S31 0.31 0.07 0.37 0.70 S32 0.73 0.18 0.86 1.64 S33 0.32 0.08 0.38 0.71 S34 0.68 0.17 0.81 1.55 S35 0.23 0.05 0.27 0.48 S41 0.38 0.09 0.44 0.83 S42 0.81 0.20 0.95 1.83 S43 2.13 0.53 2.51 4.84 S44 0.45 0.11 0.53 1.01 S45 2.71 0.67 3.21 6.22 S51 0.46 0.11 0.54 0.99 S52 0.69 0.17 0.82 1.56 S53 0.37 0.09 0.43 0.81 S54 0.26 0.06 0.30 0.53 S55 0.75 0.18 0.88 1.66 Min. 0.12 0.03 0.14 0.24 Max. 2.71 0.67 3.21 6.22 Mean 0.68 0.17 0.80 1.53

indoor outdoor Hex Hin

the worldwide average.

The surface soil layer around the uranium mine hole has uranium activities greater than worldwide average; this can mainly due to the waste of drilling and exploration left on the surface layer of soil surrounding the mine.

The thorium activities were within normal level in the studied area. Generally, potassium radionuclide in soil samples was in the range of worldwide average.

The absorbed dose rates of studied area are higher than the criterion limit of gamma radiation dose rate with an average of nine times.

Finally, from the radiation protection point of view the studied area is considered to be not safe inhabitants because the values of both internal and external hazard indexes associated with the samples are higher than unity. Thus, the human inside the area are supposed to acquire radiological complication.

#### **6. References**


**2** 

*Brazil* 

**Research Reactor Fuel Fabrication to** 

**Produce Radioisotopes** 

H. G. Riella2 and M. Durazzo1

A. M. Saliba-Silva1, E. F. Urano de Carvalho1,

*Brazilian Commission of Nuclear Energy, São Paulo,* 

*1Nuclear Fuel Center of Nuclear and Energy Research Institute* 

*2Chemical Engineering Department of University of Santa Catarina, Florianopólis,* 

This chapter describes the manufacturing technology of fuel used in research reactors that produce radioisotopes. Besides this production, the research reactors are also used for materials testing. The most common type of research reactors is called "MTR" - Materials Testing Reactor. The MTR fuel elements use fuel plates, which are quite common around the world. There was a historic development in that fuel type over the years to reach the current

The basic MTR fuel element is an assembled set of aluminum fuel plates. It consists of regularly spaced plates forming a fuel assembly. These spaces allow a stream flow of water that serves as coolant and also as moderator to nuclear reaction. The fuel plates have a meat containing the fissile material, which is entirely covered with aluminum. They are manufactured by adopting the traditional assembling technique of dispersion fuel briquette inserted in a frame covered by aluminum plates, which are welded with subsequent rolling. This technique is known internationally under the name "picture-frame technique". Powder metallurgy techniques are used in the manufacture of the fuel plate meats, making briquettes using ceramic or metallic composites. The briquette is made with powdered nuclear material

and pure aluminum powder, which is the structural material matrix of the briquette.

uranium enriched up to 20% of 235U isotope, which is the nuclear fissile material.

Using UF6 in the chemical plant, it is able to produce several intermediate compounds of uranium. One of these compounds is UF4, which is the main raw material to produce metallic uranium. It could be made by several routes. The production of metallic uranium uses the UF4 reduction through calcio- and magnesiothermic reaction. The metallic uranium is alloyed with Al, Si or Mo. Previously, stable uranium oxides were used as MTR fuels, but they had very small densities to accomplish a good operational performance of the reactors. The fuel material candidate mostly prone to be used in nuclear research reactors is based on alloys carrying more U-density toward the fuel meat. In present state, the U-Mo alloys are good candidates, but it would not be subject of the present chapter since it on its path to be certified to future use in research reactors. Currently, the most used material is U3Si2 LEU, which is low enriched

**1. Introduction** 

state-of-art in this technology.


A. M. Saliba-Silva1, E. F. Urano de Carvalho1, H. G. Riella2 and M. Durazzo1 *1Nuclear Fuel Center of Nuclear and Energy Research Institute Brazilian Commission of Nuclear Energy, São Paulo, 2Chemical Engineering Department of University of Santa Catarina, Florianopólis, Brazil* 

#### **1. Introduction**

20 Radioisotopes – Applications in Physical Sciences

José A. dos Santos, Jorge J. R., Cleomacio M. da Silva, Suêldo V. S. and Romilton dos Santos

Littlefield T.A. and Thorley N. (1974). *Atomic and Nuclear Physics*, Third edition. Van

Maher O. El-Ghossain and Raed M. Abu Saleh (2007). Radiation Measurements in Soil in the

Marcelo F. M. and Pedro M. J. (2007). Gamma Spectroscopy in the Determination of

NCRP (1987), National Council on Radiation Protection and Measurements, *Exposure of the Population in the United States and Canada from natural background radiation*. No.94, USA. Nour K. A. (2004), Natural Radioactivity of Ground and Drinking Water in Some Areas of

OECD (1979). *Exposure to Radiation From the Natural Radioactivity in Building Materials*,

Podgorsak E .B. (2005). *Radiation Physics for Medical Physicist*, Springer Berlin Heidelberg,

Rowland R. E. (1993). Low-Level Radium Retention by the Human Body; A Modification of the ICRP Publication 20 Retention Equation, *Health Phys*, 65: 507-513. Scholten L. C. and Timmermans. C. W. M. (1996). Natural Radioactivity in Phosphate

UNSCEAR (1982). *Sources effects and risks of ionizing radiation*, Report to the General

UNSCEAR (1988). *Sources effects and risks of ionizing radiation*, Report to the General

UNSCEAR (1993). *Sources effects and risks of ionizing radiation*, Report to the General

UNSCEAR (2000). *Sources effects and risks of ionizing radiation*, Report to the General

Valter A. B., Francisco J. F. F., William C. P. (2008). Concentration of Radioactive Elements

Vosniakos F., Zavalaris K., Papaligas T. (2003). Indoor Concentration of Natural

Vosniakos F., Zavalaris K., Papaligas T., Aladjadjiyan A. and Ivanova D. (2002).

WHO (1993). *Guideliens for Drinking –Water Quality "Recommendations*, World Health

Yasir M.S., Majid A. Ab., Yahaya R. (2007). Study of Natural Radionuclides and Its Radiation Hazard Index in Malaysia Building Material, *Radioanal. Nucl. Chem.*, 273: 539-541.

Greece, *Journal of Environmental protection and Ecology*, 3: 24-29.

*Archives of Biology and Technology*, 48: 221-228.

Nostrand Reinhold company, London, UK.

*Conference-INAC*, Santos, SP, Brazil.

Fertilizers, *Fertilizer Resrch*., 43: 103-107.

Atomic Radiation, New York, United Nations.

Atomic Radiation, New York, United Nations.

Atomic Radiation, New York, United Nations.

Atomic Radiation, New York, United Nations.

*Archives of Biology and Technology*, 51: 1255-1266.

*and Ecology*, 4: 733-737.

Organization, Geneva.

New York, USA.

*Journal (Series of Natural Studies and Engineering)*, 1 5: 23-37.

for Economic Cooperation and Development, Paris, France.

Upper Egypt, *Turkish J. Eng. Env. Sci.*, 28: 345-354.

A. (2005), Analysis of the 40K Levels in Soil using Gamma Spectrometry, *Brazilian* 

Middle of Gaza-Strip Using Different Type of Detectors, *The Islamic University* 

Radionuclides Comprised in Radioactive Series, *International Nuclear Atomic* 

Report by a Group of Experts of the OECD Nuclear Energy Agency, Organization

Assembly, With Annexes, United Nations Scientific Committee on the Effects of

Assembly, With Annexes, United Nations Scientific Committee on the Effects of

Assembly, With Annexes, United Nations Scientific Committee on the Effects of

Assembly, With Annexes, United Nations Scientific Committee on the Effects of

(U, Th and K) Derived From Phosphatic Fertilizers in Cultivated Soils, *Brazilian* 

Radioactivity and The Impact to Human Health, *Journal of Environmental Protection* 

Measurements of Natural Radioactivity Concentration of Building Material in

This chapter describes the manufacturing technology of fuel used in research reactors that produce radioisotopes. Besides this production, the research reactors are also used for materials testing. The most common type of research reactors is called "MTR" - Materials Testing Reactor. The MTR fuel elements use fuel plates, which are quite common around the world. There was a historic development in that fuel type over the years to reach the current state-of-art in this technology.

The basic MTR fuel element is an assembled set of aluminum fuel plates. It consists of regularly spaced plates forming a fuel assembly. These spaces allow a stream flow of water that serves as coolant and also as moderator to nuclear reaction. The fuel plates have a meat containing the fissile material, which is entirely covered with aluminum. They are manufactured by adopting the traditional assembling technique of dispersion fuel briquette inserted in a frame covered by aluminum plates, which are welded with subsequent rolling. This technique is known internationally under the name "picture-frame technique". Powder metallurgy techniques are used in the manufacture of the fuel plate meats, making briquettes using ceramic or metallic composites. The briquette is made with powdered nuclear material and pure aluminum powder, which is the structural material matrix of the briquette.

Using UF6 in the chemical plant, it is able to produce several intermediate compounds of uranium. One of these compounds is UF4, which is the main raw material to produce metallic uranium. It could be made by several routes. The production of metallic uranium uses the UF4 reduction through calcio- and magnesiothermic reaction. The metallic uranium is alloyed with Al, Si or Mo. Previously, stable uranium oxides were used as MTR fuels, but they had very small densities to accomplish a good operational performance of the reactors. The fuel material candidate mostly prone to be used in nuclear research reactors is based on alloys carrying more U-density toward the fuel meat. In present state, the U-Mo alloys are good candidates, but it would not be subject of the present chapter since it on its path to be certified to future use in research reactors. Currently, the most used material is U3Si2 LEU, which is low enriched uranium enriched up to 20% of 235U isotope, which is the nuclear fissile material.

made possible to measure the uptake of iodine by the thyroid and thereby assess the functioning of the gland. Another radioactive element widely used to study the various functions of the human organism is technetium-99m. This isotope can be chemically combined with various organic complexes, which evaluate liver disorders, bone and brain, among others. In bone scintigraphy, the radioactivity of technetium reveals the existence of tumors from six to eight months before they have reached sufficient size to be picked up by X-ray examinations. With this, it is possible to start treatment much earlier with greater cure

Nuclear reactors that produce radioisotopes are called research reactors. This type of reactor is also used to perform tests on materials and nuclear fuels in the development phase. The modern research reactors are designed with both purposes, radioisotope production and

Unlike power reactors, which are well known and are intended to generate heat for electricity generation, the research reactors or the modern multipurpose reactors aim to generate neutrons used for radioisotopes production or for testing materials in terms of verify their performance under irradiation. Unlike power reactors, research reactors operating with much higher power density, which is necessary to get high neutron fluxes. For this reason its fuel is usually in the form of a metal plate, usually covered by aluminum. They are very different from the fuel rod with ceramic pellets (UO2) as used in the fuel for

The research reactors moderated and cooled with light water and using plate-type fuel elements has been named MTR type reactors (Materials Testing Reactor). After the construction of the first MTR, a joint venture of ORNL (Oak Ridge National Laboratory) and ANL (Argonne National Laboratory) operated it since March 31, 1952. Many research reactors around the world uses MTR type fuel elements, which are formed by assembling fuel plates fabricated by a well-known and established technique of assembling a core, commonly named fuel meat, which incorporates the fissile material, a frame plate and two cladding plates, with subsequent deformation by hot and cold-rolling (picture frame

Initially, the fuel plates usually used as the core material an uranium-aluminum alloy (U-Al) containing 18 wt% of highly enriched uranium (93 wt% 235U) (1) (3). Even in the 50's, with the concern about nuclear weapons non-proliferation, the research reactors began to use fuels containing low-enriched uranium (20 wt% 235U) (4). With enrichment lowering, in order to maintain the reactivity and lifetime of the reactor cores, it became necessary to increase the amount of uranium in each fuel plate. In the U-Al alloy, the uranium concentration had to be

Fuel plates containing the meat based on the U-Al alloy with 18 wt% of highly enriched uranium were easily fabricated. However, difficulties arise in fabricating fuel plates with meats of U-Al alloy containing 45 wt% of low-enriched uranium, because of the fragility and propensity for segregation of this alloy (4) (5) (6). An alternative to overcome this problem was the use of cores manufactured by powder metallurgy, which used dispersions of uranium compounds in aluminum and could incorporate quantities of low-enriched uranium significantly greater. For instance, the Argonauta reactor (10 MW), in Rio de Janeiro, Brazil, started its operation in 1956 and was developed by the Argonne National Laboratory, USA. This pioneer Brazilian research reactor used fuel plates with the meat

based on an U3O8-Al dispersion containing 39 wt% of U3O8 with low enrichment (7).

increased to 45 wt% to compensate the decrease in the enrichment level.

testing of materials, and for this reason are called Multipurpose Reactors.

perspective.

power reactors.

technique) (1) (2).

The production procedures of U3Si2 fuel fabrication will be discussed in this chapter, starting from U3Si2 fabrication and powder manufacture. This powder is mixed with aluminum powder and pressed, resulting in a solid briquette with good mechanical strength. After quality inspection, the briquette becomes the fuel plate meat.

The fuel plate manufacturing procedures will be described according to the picture frame technique. This technique includes the assembling of the briquette inside the frame sandwiched with cover plates. The assembly is welded and hot and cold rolled to get the fuel plate, where the fuel meat is completely sealed inside aluminum. All the process and quality control during fabrication will be commented ahead.

Once the plates having been fabricated, the fuel assembly is finally made by fixing the fuel plates and the other mechanical components, such as nozzle, handle and screws. This finishing process to produce the element is also commented in this chapter.

The characteristics of the fuel plates must meet specifications of each particular research reactor characteristics. Inspections and qualifications are carried out in various stages of fuel plate manufacture.

As this chapter describes nuclear fuel manufacturing for research reactors, the sub-items of fabrication process can be divided into the following topics: evolution of nuclear fuel materials for MTR fuel; production of uranium hexafluoride (UF6); production of uranium tetrafluoride (UF4); production of metallic uranium; U3Si2 production; production of fuel cores from U3Si2 powder and aluminum; production of fuel plates with U3Si2-Al dispersion briquettes; assembling of fuel elements; recovery of uranium; effluent treatment; quality control.

In this chapter, the experience of IPEN/CNEN-SP (Energy and Nuclear Research Institute of Brazilian Commission of Nuclear Energy, São Paulo, Brazil) will be given as a productive route to produce MTR nuclear fuels for research reactors, since this is the main expertise of all the authors of this chapter.

## **2. Evolution of nuclear materials for research reactors fuel**

The use of radioisotopes in medicine is certainly one of the most important social uses of nuclear energy. Radiopharmaceuticals are radioactive substances that help doctors to make important decisions for treatments in oncology, cardiology, neurology, among other areas. For patients, the diagnoses represent safety and pain relief, as in the case of samarium-153 use, which is employed to relieve bone pain caused by metastatic tumors.

Nuclear medicine is a medical specialty that uses radioactive material for diagnostic tests and therapeutic purposes. Although it is often confused with radiotherapy, the last application has a lot of different procedures and applications. The main distinction between the two specialties is the way both use the radioactive material. While radiotherapy (radiation therapy) uses sealed sources (or closed), which emit radiation outside the patient, nuclear medicine uses open sources of radiation, administered in vivo (oral or intravenous). If, in radiotherapy, radiation is directed toward the point to be discussed, in nuclear medicine is the body own metabolism of the patient who is in charge of carrying radioactive material into the organ to be examined or treated.

The success of nuclear medicine in diagnosis is due to its ability to show the functioning of various body organs, avoiding the use of invasive techniques such as biopsy and catheterization. The use of ultrapure iodine-123 to examine thyroid function is one example. By scintigraphy, a diagnostic imaging technique, which has several medical applications,

The production procedures of U3Si2 fuel fabrication will be discussed in this chapter, starting from U3Si2 fabrication and powder manufacture. This powder is mixed with aluminum powder and pressed, resulting in a solid briquette with good mechanical

The fuel plate manufacturing procedures will be described according to the picture frame technique. This technique includes the assembling of the briquette inside the frame sandwiched with cover plates. The assembly is welded and hot and cold rolled to get the fuel plate, where the fuel meat is completely sealed inside aluminum. All the process and

Once the plates having been fabricated, the fuel assembly is finally made by fixing the fuel plates and the other mechanical components, such as nozzle, handle and screws. This

The characteristics of the fuel plates must meet specifications of each particular research reactor characteristics. Inspections and qualifications are carried out in various stages of fuel

As this chapter describes nuclear fuel manufacturing for research reactors, the sub-items of fabrication process can be divided into the following topics: evolution of nuclear fuel materials for MTR fuel; production of uranium hexafluoride (UF6); production of uranium tetrafluoride (UF4); production of metallic uranium; U3Si2 production; production of fuel cores from U3Si2 powder and aluminum; production of fuel plates with U3Si2-Al dispersion briquettes; assembling of fuel elements; recovery of uranium; effluent treatment; quality

In this chapter, the experience of IPEN/CNEN-SP (Energy and Nuclear Research Institute of Brazilian Commission of Nuclear Energy, São Paulo, Brazil) will be given as a productive route to produce MTR nuclear fuels for research reactors, since this is the main expertise of

The use of radioisotopes in medicine is certainly one of the most important social uses of nuclear energy. Radiopharmaceuticals are radioactive substances that help doctors to make important decisions for treatments in oncology, cardiology, neurology, among other areas. For patients, the diagnoses represent safety and pain relief, as in the case of samarium-153

Nuclear medicine is a medical specialty that uses radioactive material for diagnostic tests and therapeutic purposes. Although it is often confused with radiotherapy, the last application has a lot of different procedures and applications. The main distinction between the two specialties is the way both use the radioactive material. While radiotherapy (radiation therapy) uses sealed sources (or closed), which emit radiation outside the patient, nuclear medicine uses open sources of radiation, administered in vivo (oral or intravenous). If, in radiotherapy, radiation is directed toward the point to be discussed, in nuclear medicine is the body own metabolism of the patient who is in charge of carrying radioactive

The success of nuclear medicine in diagnosis is due to its ability to show the functioning of various body organs, avoiding the use of invasive techniques such as biopsy and catheterization. The use of ultrapure iodine-123 to examine thyroid function is one example. By scintigraphy, a diagnostic imaging technique, which has several medical applications,

strength. After quality inspection, the briquette becomes the fuel plate meat.

finishing process to produce the element is also commented in this chapter.

**2. Evolution of nuclear materials for research reactors fuel** 

use, which is employed to relieve bone pain caused by metastatic tumors.

material into the organ to be examined or treated.

quality control during fabrication will be commented ahead.

plate manufacture.

all the authors of this chapter.

control.

made possible to measure the uptake of iodine by the thyroid and thereby assess the functioning of the gland. Another radioactive element widely used to study the various functions of the human organism is technetium-99m. This isotope can be chemically combined with various organic complexes, which evaluate liver disorders, bone and brain, among others. In bone scintigraphy, the radioactivity of technetium reveals the existence of tumors from six to eight months before they have reached sufficient size to be picked up by X-ray examinations. With this, it is possible to start treatment much earlier with greater cure perspective.

Nuclear reactors that produce radioisotopes are called research reactors. This type of reactor is also used to perform tests on materials and nuclear fuels in the development phase. The modern research reactors are designed with both purposes, radioisotope production and testing of materials, and for this reason are called Multipurpose Reactors.

Unlike power reactors, which are well known and are intended to generate heat for electricity generation, the research reactors or the modern multipurpose reactors aim to generate neutrons used for radioisotopes production or for testing materials in terms of verify their performance under irradiation. Unlike power reactors, research reactors operating with much higher power density, which is necessary to get high neutron fluxes. For this reason its fuel is usually in the form of a metal plate, usually covered by aluminum. They are very different from the fuel rod with ceramic pellets (UO2) as used in the fuel for power reactors.

The research reactors moderated and cooled with light water and using plate-type fuel elements has been named MTR type reactors (Materials Testing Reactor). After the construction of the first MTR, a joint venture of ORNL (Oak Ridge National Laboratory) and ANL (Argonne National Laboratory) operated it since March 31, 1952. Many research reactors around the world uses MTR type fuel elements, which are formed by assembling fuel plates fabricated by a well-known and established technique of assembling a core, commonly named fuel meat, which incorporates the fissile material, a frame plate and two cladding plates, with subsequent deformation by hot and cold-rolling (picture frame technique) (1) (2).

Initially, the fuel plates usually used as the core material an uranium-aluminum alloy (U-Al) containing 18 wt% of highly enriched uranium (93 wt% 235U) (1) (3). Even in the 50's, with the concern about nuclear weapons non-proliferation, the research reactors began to use fuels containing low-enriched uranium (20 wt% 235U) (4). With enrichment lowering, in order to maintain the reactivity and lifetime of the reactor cores, it became necessary to increase the amount of uranium in each fuel plate. In the U-Al alloy, the uranium concentration had to be increased to 45 wt% to compensate the decrease in the enrichment level.

Fuel plates containing the meat based on the U-Al alloy with 18 wt% of highly enriched uranium were easily fabricated. However, difficulties arise in fabricating fuel plates with meats of U-Al alloy containing 45 wt% of low-enriched uranium, because of the fragility and propensity for segregation of this alloy (4) (5) (6). An alternative to overcome this problem was the use of cores manufactured by powder metallurgy, which used dispersions of uranium compounds in aluminum and could incorporate quantities of low-enriched uranium significantly greater. For instance, the Argonauta reactor (10 MW), in Rio de Janeiro, Brazil, started its operation in 1956 and was developed by the Argonne National Laboratory, USA. This pioneer Brazilian research reactor used fuel plates with the meat based on an U3O8-Al dispersion containing 39 wt% of U3O8 with low enrichment (7).

The problem encountered in using these intermetallics with high concentrations of uranium as fissile material in the form of dispersions in aluminum is related to its dimensional stability during operation, leading to swelling of the fuel plates and therefore the problems that compromise the thermohydraulic security of the reactor. In mid-1988, based on results from irradiation tests (12) (13), the U3Si2-Al based dispersion fuel was qualified by the U.S. Nuclear Regulatory Commission and released for sale with uranium densities up to 4.8

Fig. 1. Density of uranium in terms of concentration of dispersed phase for different fissile

Research continued aiming at the use of intermetallic with even higher concentrations of uranium, such as U3Si, U3SiAl and U6Fe as fissile material in the form of dispersions in aluminum. However, results of irradiation tests showed an unacceptable dimensional stability of these new fuels. Due to its high concentration of uranium (96 wt%) the U6Fe was mainly considered (15), and the research was virtually abandoned in 1986 due to high swelling observed in irradiation tests, coupled with the promising results obtained with the

Only through the use of U3Si2 as fissile material in the dispersion with aluminum was not possible to convert all research reactors. Many research reactors are awaiting a high-

U3Si2-Al dispersion, being considered a viable alternative (16).

uranium compounds.

gU/cm3, with a swelling consistent with the commonly used dispersions (14).

Efforts were made to increase the concentration of uranium in this type of dispersion fuel, getting 65 wt% of U3O8 in the fuel fabricated for the Puerto Rico Research Reactor of the Puerto Rico Nuclear Center to the end of the 70's (8).

Aiming at obtaining more and more high neutron fluxes, the development of research reactors with higher power required a continuous production of fuels, which used highly enriched uranium (93 wt% 235U), yielding higher specific reactivity and economics, since these fuels could stay longer in the reactor core (long life). The 100 MW HFIR (High Flux Isotope Reactor) used dispersion U3O8-Al with 40 wt% U3O8 (9) and the ATR (Advanced Test Reactor), with 250 MW, used the same type of dispersion with 34 wt% highly enriched U3O8 (10). In addition to the U3O8-Al dispersions, UAlx-Al dispersions were commonly used (x is approximately 3), all these fuels systems still using highly enriched uranium. At this time, in late 70's, the highest uranium density obtained inside the fuel was 1.7 gU/cm3, which was quite well qualified.

Since highly enriched uranium was easily obtainable in the 70's, the commercial reactors that were using the low-enriched uranium started gradually to convert their cores to highly enriched fuel. Thus, it reached a total of approximately 156 research reactors in 34 countries using highly enriched uranium, resulting in an annual circulation of approximately 5000 kg of this material (11). In 1977, arose again the concern about the proliferation risk associated with loss of fuel during manufacture, transport and storage, leading to restriction by the U.S. government's sale of uranium with high enrichment (above 90 wt% 235U) and producing an impact on the availability and use of the highly enriched fuel for research reactors.

From 1978 programs, it was established for the enrichment reduction, aimed at developing the technology base for replacement of highly enriched uranium by low-enriched uranium (less than 20 wt% 235U) in research reactors. The main program, still active today, is the *RERTR Program* (Reduced Enrichment for Research and Test Reactors), which aims to develop the technology necessary to convert the reactors that use highly enriched uranium (= or > 20% 235U) by low-enriched uranium (less than 20% 235U). During the existence of this program more than 40 research reactors have been converted. At this time, the decrease of enrichment has demanded an effort bigger that previously, because, in most high power research reactors, which are designed to operate in extremes, this substitution involved the development and qualification of new fuels with maximum possible concentration of uranium, which limits are imposed for manufacturability and performance under severe and prolonged irradiation.

In this context, the developments were based initially on increasing the concentration of uranium in the fuel currently used at the beginning of RERTR program, until the practical limit of 2.3 gU/cm3 in the case of UAlx-Al and 3.2 gU/cm3 in the case of U3O8-Al. Also, an effort was made in developing new fuels that would allow obtaining uranium densities of 6- 7 gU/cm3, well above the density that can be achieved with the UAlx-Al and U3O8-Al fuel. The development of new fuels would allow the conversion to low enriched from virtually all existing research reactors.

High density of uranium in the dispersion can only be achieved by using the dispersion of fissile compounds with high uranium content. Figure 1 shows the potential of various uranium compounds. The technological limit for the use of dispersions is 45% by volume of fissile material dispersed, since it must be kept a solid aluminum matrix as dispersant. The uranium silicides and U6Fe compounds were initially considered promising.

Efforts were made to increase the concentration of uranium in this type of dispersion fuel, getting 65 wt% of U3O8 in the fuel fabricated for the Puerto Rico Research Reactor of the

Aiming at obtaining more and more high neutron fluxes, the development of research reactors with higher power required a continuous production of fuels, which used highly enriched uranium (93 wt% 235U), yielding higher specific reactivity and economics, since these fuels could stay longer in the reactor core (long life). The 100 MW HFIR (High Flux Isotope Reactor) used dispersion U3O8-Al with 40 wt% U3O8 (9) and the ATR (Advanced Test Reactor), with 250 MW, used the same type of dispersion with 34 wt% highly enriched U3O8 (10). In addition to the U3O8-Al dispersions, UAlx-Al dispersions were commonly used (x is approximately 3), all these fuels systems still using highly enriched uranium. At this time, in late 70's, the highest uranium density obtained inside the fuel was 1.7 gU/cm3,

Since highly enriched uranium was easily obtainable in the 70's, the commercial reactors that were using the low-enriched uranium started gradually to convert their cores to highly enriched fuel. Thus, it reached a total of approximately 156 research reactors in 34 countries using highly enriched uranium, resulting in an annual circulation of approximately 5000 kg of this material (11). In 1977, arose again the concern about the proliferation risk associated with loss of fuel during manufacture, transport and storage, leading to restriction by the U.S. government's sale of uranium with high enrichment (above 90 wt% 235U) and producing an impact on the availability and use of the highly enriched fuel for research

From 1978 programs, it was established for the enrichment reduction, aimed at developing the technology base for replacement of highly enriched uranium by low-enriched uranium (less than 20 wt% 235U) in research reactors. The main program, still active today, is the *RERTR Program* (Reduced Enrichment for Research and Test Reactors), which aims to develop the technology necessary to convert the reactors that use highly enriched uranium (= or > 20% 235U) by low-enriched uranium (less than 20% 235U). During the existence of this program more than 40 research reactors have been converted. At this time, the decrease of enrichment has demanded an effort bigger that previously, because, in most high power research reactors, which are designed to operate in extremes, this substitution involved the development and qualification of new fuels with maximum possible concentration of uranium, which limits are imposed for manufacturability and performance under severe

In this context, the developments were based initially on increasing the concentration of uranium in the fuel currently used at the beginning of RERTR program, until the practical limit of 2.3 gU/cm3 in the case of UAlx-Al and 3.2 gU/cm3 in the case of U3O8-Al. Also, an effort was made in developing new fuels that would allow obtaining uranium densities of 6- 7 gU/cm3, well above the density that can be achieved with the UAlx-Al and U3O8-Al fuel. The development of new fuels would allow the conversion to low enriched from virtually

High density of uranium in the dispersion can only be achieved by using the dispersion of fissile compounds with high uranium content. Figure 1 shows the potential of various uranium compounds. The technological limit for the use of dispersions is 45% by volume of fissile material dispersed, since it must be kept a solid aluminum matrix as dispersant. The

uranium silicides and U6Fe compounds were initially considered promising.

Puerto Rico Nuclear Center to the end of the 70's (8).

which was quite well qualified.

and prolonged irradiation.

all existing research reactors.

reactors.

The problem encountered in using these intermetallics with high concentrations of uranium as fissile material in the form of dispersions in aluminum is related to its dimensional stability during operation, leading to swelling of the fuel plates and therefore the problems that compromise the thermohydraulic security of the reactor. In mid-1988, based on results from irradiation tests (12) (13), the U3Si2-Al based dispersion fuel was qualified by the U.S. Nuclear Regulatory Commission and released for sale with uranium densities up to 4.8 gU/cm3, with a swelling consistent with the commonly used dispersions (14).

Fig. 1. Density of uranium in terms of concentration of dispersed phase for different fissile uranium compounds.

Research continued aiming at the use of intermetallic with even higher concentrations of uranium, such as U3Si, U3SiAl and U6Fe as fissile material in the form of dispersions in aluminum. However, results of irradiation tests showed an unacceptable dimensional stability of these new fuels. Due to its high concentration of uranium (96 wt%) the U6Fe was mainly considered (15), and the research was virtually abandoned in 1986 due to high swelling observed in irradiation tests, coupled with the promising results obtained with the U3Si2-Al dispersion, being considered a viable alternative (16).

Only through the use of U3Si2 as fissile material in the dispersion with aluminum was not possible to convert all research reactors. Many research reactors are awaiting a high-

When the reduction process to produce metallic uranium is performed at higher pressures and lower temperatures, normal tolerances up to 4% of UO2 + UO2F2 should be reduced. It is recommended that the tapped density of loose UF4 should be greater than 1 g.cm-3.The good quality of the metallic uranium to be produced should have the UF4specification as

Elements Al B Cd C Co Cr Cu Fe Mn Ni Si % in mass 70 0,2 0,1 150 5 25 40 75 15 40 30

In the case the oxide content is high, there would be larger losses of metal with the slag. As the reaction develops high amount of heat, it should be avoided evolution of tetrafluoride volatile components such as water and ammonia. During smelting, the metal is slightly contaminated with impurities from reducing agent and crucible. For this reason, UF4 should be pure enough to allow slight contamination degree during this process. It must also be sufficiently dense. The load consists of blending of UF4 powder and chips of calcium or magnesium tetrafluoride. The higher the density of UF4, the greater the density of the load,

The production of uranium tetrafluoride can be made by several processes which are divided into two groups, namely dry (fluorination of uranium oxide or hexafluoride reduction) and aqueous (preparation of UF4 from U+6 salt) pathway (18) (19) (20) (21) (22). The first task of obtaining UF4 were carried through water (22) (23) by the end of the 19th century, and from an industrial standpoint that prevailed till the beginning of 20th century. The process essentially comprises the steps of reducing the uranium contained in uranyl fluoride solutions, uranyl chloride or uranyl sulfate up to its tetravalent state, followed by

With the development of dry processes, the aqueous processes were abandoned because they had difficulties in filtration, washing and drying, in spite of their simplicity and safety. Nowadays, the production via aqueous route is only used in plants to produce UF4 for small quantities, which is the present experience of IPEN in producing LEU UF4. Nevertheless,

The UF4 preparation methods through water have been developed mostly by the British and

Essentially, the process consists in reducing the uranium, contained in solutions of uranyl fluoride, uranyl chloride and uranyl sulfate to the tetravalent state and the precipitation of uranium tetrafluoride by adding hydrofluoric acid. Several compounds of uranium have been used as starting materials and various reducing agents have been used. An overview of the process can be obtained from the reaction of uranyl fluoride with stannous chloride and

UO2F2 + SnCl2 + 4HF UF4 + 2H2 O + SnCl2F2 (1)

↔

UO2 F2 + Na2S2O4 + UF4 + 2HF Na2SO3 + SO2 + H2O (2)

↔

and the greater the amount of heat involved per unit volume of the furnace (17).

Table 1. Specification of the limits for main impurities in UF4

UF4 precipitation by adding hydrofluoric acid.

IPEN also developed the Brazilian technology for dry route.

its modifications were based on work done by Bolton in 1866 (24; 25)

**3.1 Procedures for obtaining UF4 via wet process** 

**3.1.1 Preparation of UF4 from salts U+6** 

sodium hyposulphide.

displayed in Table 1.

performance technology solution finale, needing a uranium density of 6-9 gU/cm3. In an effort to convert these reactors, other high density fuels has been studied, including dispersions based on U-Mo, U3SiCu, U3Si1.5, U3Si1.6, U75Ga15Ge10, U75Ga10Si15 and uranium nitrides. Still, innovative manufacturing techniques has been investigated, which are based on hot isostatic compaction (HIP - Hot Isostatic Pressing) or increasing the volume fraction of U3Si2 beyond 50% (the limit currently accepted for this technology is 45%) or using wires of U3Si and/or U75Ga10Si15 and/or U75Ga15Ge10 metallurgically bonded with aluminum in a geometry such that result in plates with a density close to 9 gU/cm3 in the fuel core. The nearest alternative to be commercially deployed is the UMo alloy dispersion in aluminum, which enables to achieve the density near to 8 gU/cm3. The performance under irradiation of this type of fuel is being tested with promising results. However, it is not a commercial fuel yet.

Thus, currently the most advanced manufacturing technology commercially available for the MTR type fuel plates is based on the U3Si2-Al dispersion, with a concentration of U3Si2 resulting in a uranium density into the fuel meat of 4.8 gU/cm3. The next commercially available technology will probably use a dispersion of UMo alloy with 7-10 wt% Mo, resulting in a uranium density of between 6 and 8 gU/cm3.

Each type of MTR fuel element is produced in accordance with a manufacturing specification and a set of manufacturing drawings agreed between the fabricator and the reactor operator or his representative. The specification sets down the scope and general conditions, the requirements of manufacturing method, together with the inspection requirements and acceptance criteria. In addition to the specification, an inspection schedule is normally produced which includes all of the supporting documentation such as the inspection and record sheet and certification (17; 18; 19; 20; 21).

#### **3. Production of uranium tetrafluoride**

The UF4 has a specific role in nuclear fuel technology. It is an important intermediate product, being the basic substance to produce either uranium as metal (Uo) or uranium hexafluoride (UF6) (17).

Uranium tetrafluoride (UF4) is a green crystalline solid that melts at about 96°C and has an insignificant vapor pressure. It is slightly soluble in water. UF4 is less stable than uranium oxides and produces hydrofluoric acid in reaction with water; thus it is a less favorable form for long-term disposal. The bulk density of UF4 varies from about 2.0 g/cm3 to about 4.5 g/cm3 depending on the production process and the properties of the starting uranium compounds. Uranium tetrafluoride (UF4) reacts slowly with moisture at room temperature, forming UO2 and HF, which are very corrosive.

In principle, several other compounds may also be used for the production of metal and hexafluoride uranium, however, the use of UF4 is prescribed by technological and economic considerations. It is considerably easier to obtain metallic uranium from UF4, due to the reactivity of UF4 mixture with reducing agent (mainly Ca and Mg) with large thermal outcome, which makes easy the production of uranium ingot.

According to the production process, the UF4 must have certain specifications in regard to its purity. The content of uranium oxides and uranyl fluoride (UO2F2) may vary and also its density and its granulometric composition. The major technical requirement for tetrafluoride is observed during metallic uranium fabrication. It must contain at least 96% of tetrafluoride, virtually free of impurities. It should be anhydrous and having sufficiently high density.

performance technology solution finale, needing a uranium density of 6-9 gU/cm3. In an effort to convert these reactors, other high density fuels has been studied, including dispersions based on U-Mo, U3SiCu, U3Si1.5, U3Si1.6, U75Ga15Ge10, U75Ga10Si15 and uranium nitrides. Still, innovative manufacturing techniques has been investigated, which are based on hot isostatic compaction (HIP - Hot Isostatic Pressing) or increasing the volume fraction of U3Si2 beyond 50% (the limit currently accepted for this technology is 45%) or using wires of U3Si and/or U75Ga10Si15 and/or U75Ga15Ge10 metallurgically bonded with aluminum in a geometry such that result in plates with a density close to 9 gU/cm3 in the fuel core. The nearest alternative to be commercially deployed is the UMo alloy dispersion in aluminum, which enables to achieve the density near to 8 gU/cm3. The performance under irradiation of this type of fuel is being tested with promising results. However, it is not a commercial

Thus, currently the most advanced manufacturing technology commercially available for the MTR type fuel plates is based on the U3Si2-Al dispersion, with a concentration of U3Si2 resulting in a uranium density into the fuel meat of 4.8 gU/cm3. The next commercially available technology will probably use a dispersion of UMo alloy with 7-10 wt% Mo,

Each type of MTR fuel element is produced in accordance with a manufacturing specification and a set of manufacturing drawings agreed between the fabricator and the reactor operator or his representative. The specification sets down the scope and general conditions, the requirements of manufacturing method, together with the inspection requirements and acceptance criteria. In addition to the specification, an inspection schedule is normally produced which includes all of the supporting documentation such as the

The UF4 has a specific role in nuclear fuel technology. It is an important intermediate product, being the basic substance to produce either uranium as metal (Uo) or uranium

Uranium tetrafluoride (UF4) is a green crystalline solid that melts at about 96°C and has an insignificant vapor pressure. It is slightly soluble in water. UF4 is less stable than uranium oxides and produces hydrofluoric acid in reaction with water; thus it is a less favorable form for long-term disposal. The bulk density of UF4 varies from about 2.0 g/cm3 to about 4.5 g/cm3 depending on the production process and the properties of the starting uranium compounds. Uranium tetrafluoride (UF4) reacts slowly with moisture at room temperature,

In principle, several other compounds may also be used for the production of metal and hexafluoride uranium, however, the use of UF4 is prescribed by technological and economic considerations. It is considerably easier to obtain metallic uranium from UF4, due to the reactivity of UF4 mixture with reducing agent (mainly Ca and Mg) with large thermal

According to the production process, the UF4 must have certain specifications in regard to its purity. The content of uranium oxides and uranyl fluoride (UO2F2) may vary and also its density and its granulometric composition. The major technical requirement for tetrafluoride is observed during metallic uranium fabrication. It must contain at least 96% of tetrafluoride, virtually free of impurities. It should be anhydrous and having sufficiently high density.

resulting in a uranium density of between 6 and 8 gU/cm3.

inspection and record sheet and certification (17; 18; 19; 20; 21).

**3. Production of uranium tetrafluoride** 

forming UO2 and HF, which are very corrosive.

outcome, which makes easy the production of uranium ingot.

hexafluoride (UF6) (17).

fuel yet.

When the reduction process to produce metallic uranium is performed at higher pressures and lower temperatures, normal tolerances up to 4% of UO2 + UO2F2 should be reduced. It is recommended that the tapped density of loose UF4 should be greater than 1 g.cm-3.The good quality of the metallic uranium to be produced should have the UF4specification as displayed in Table 1.


Table 1. Specification of the limits for main impurities in UF4

In the case the oxide content is high, there would be larger losses of metal with the slag. As the reaction develops high amount of heat, it should be avoided evolution of tetrafluoride volatile components such as water and ammonia. During smelting, the metal is slightly contaminated with impurities from reducing agent and crucible. For this reason, UF4 should be pure enough to allow slight contamination degree during this process. It must also be sufficiently dense. The load consists of blending of UF4 powder and chips of calcium or magnesium tetrafluoride. The higher the density of UF4, the greater the density of the load, and the greater the amount of heat involved per unit volume of the furnace (17).

The production of uranium tetrafluoride can be made by several processes which are divided into two groups, namely dry (fluorination of uranium oxide or hexafluoride reduction) and aqueous (preparation of UF4 from U+6 salt) pathway (18) (19) (20) (21) (22).

The first task of obtaining UF4 were carried through water (22) (23) by the end of the 19th century, and from an industrial standpoint that prevailed till the beginning of 20th century. The process essentially comprises the steps of reducing the uranium contained in uranyl fluoride solutions, uranyl chloride or uranyl sulfate up to its tetravalent state, followed by UF4 precipitation by adding hydrofluoric acid.

With the development of dry processes, the aqueous processes were abandoned because they had difficulties in filtration, washing and drying, in spite of their simplicity and safety. Nowadays, the production via aqueous route is only used in plants to produce UF4 for small quantities, which is the present experience of IPEN in producing LEU UF4. Nevertheless, IPEN also developed the Brazilian technology for dry route.

#### **3.1 Procedures for obtaining UF4 via wet process 3.1.1 Preparation of UF4 from salts U+6**

The UF4 preparation methods through water have been developed mostly by the British and its modifications were based on work done by Bolton in 1866 (24; 25)

Essentially, the process consists in reducing the uranium, contained in solutions of uranyl fluoride, uranyl chloride and uranyl sulfate to the tetravalent state and the precipitation of uranium tetrafluoride by adding hydrofluoric acid. Several compounds of uranium have been used as starting materials and various reducing agents have been used. An overview of the process can be obtained from the reaction of uranyl fluoride with stannous chloride and sodium hyposulphide.

$$\text{UO}\_2\text{F}\_2 + \text{SnCl}\_2 + 4\text{HF} \overset{\longleftrightarrow}{\longleftrightarrow} \text{UF}\_4 + 2\text{H}\_2\text{O} + \text{SnCl}\_2\text{F}\_2\tag{1}$$

$$\text{UO}\_2\text{F}\_2 + \text{Na}\_2\text{S}\_2\text{O}\_4 + \text{UF}\_4 + 2\text{HF} \overset{\Theta}{\longleftrightarrow} \text{Na}\_2\text{SO}\_3 + \text{SO}\_2 + \text{H}\_2\text{O} \tag{2}$$

Table 2 shows the chemical characteristics of UO2F2 solution obtained from UF6

Uranium in its tetravalent state is very important in different technological processes. Essentially, the preparation process (aqueous way) from solutions containing uranyl ion (hexavalent) involves the reduction towards tetravalent state, and later precipitation as UF4 using HF solution. In aqueous solutions, these reductions can be carried out by chemical,

All the trials for the preparation of UF4 using chemical reduction have been carried out using UO2F2 solution inside a stainless steel reactor, coated with Teflon. The solution has been heated under continuous stirring to reach a temperature set, and the reducing agent has been added. Next, the precipitating agent solution is slowly added to UO2F2 in solution with hydrofluoric acid (HF). Tests have been carried out using some reducing agents, such

UO2 F2 + SnCl2 + 4HF → UF4 + SnClF2 + 2H2O (4)

UO2 F2 + 4HF + Fe → UF4 + FeF2 + 2H2O (5)

UO2 F2 + CuCl + 4HF → UF4 + CuClF2 + 2H2O (6)

 UO2 F2 + Na2S2O4 + 2HF → UF4 + Na2 SO3 + H2O (7) Upon UF4 precipitation the suspension is left in rest up to reaching room temperature. After over 12 hours, it was performed the solid/liquid separation by vacuum filtration, washing and drying in a muffle kiln. The salts obtained were all identified as being uranium tetrafluoride. According to the results shown in Figure 3, it is evident that, from all used reducing agents, only SnCl2 and FeCl2 have shown significant results in regards of getting UF4. Nevertheless,

The influence of the temperature upon UO2F2 and UO2 contents in obtained UF4 is shown in Figure 4. It was employed SnCl2 as the reducing agent in this study to precipitate UO2F2 solution. The residual moisture is dried at 130°C. The tin content in all obtained UF4 has

SnCl2 is more consistent reducing agent at higher temperature of process.

Cd B P Fe Cr Ni Mo Zn Si Al <0.1 0.2 <100 1500 100 40 <2 100 300 40

Mn Mg Pb Sn Bi V Cu Ba Co 10 15 <2 <2 <2 <3 3 1 <10

Table 2. Chemical characteristics of UO2F2 solution

**3.1.3 Chemical reduction of UF6 to UF4**

electrochemical or photochemical methods.

as SnCl2, CuCl, FeCl2, Na2S2O4.

shown to be in the range of 0.15 – 0.15%.

hydrolysis.

Uranium (g/L) 60 Fluoride (g/L) 17

Metallic impurities ( g/mL)

An alternative to this process is the replacement of the electrolytic reduction by reducing agents that prevents the possible contamination with the reducing agent. This process has been adopted in countries like USA, Spain, Australia, Japan, Canada, England, South Africa and India (26; 27; 28; 29; 30; 31; 32). For the production of UF4 with nuclear purity from UO2F2 acids solutions, some fundamental stages are required such as obtaining the solution, reduction to uranium valence and precipitation of the formed U+4. These stages are shown in Figure 2, as schematized operations.

#### **3.1.2 Obtaining UO2F2 solutions**

Uranium hexafluoride is a crystalline substance at normal pressure and temperature conditions. At the temperature of 900C under a pressure of 3kgf/cm2, UF6 becomes gas and when it is injected into water, it hydrolyzes immediately according to the following:

$$\rm UF\_6 + H\_2O \to UO\_2F\_2 + 4HF \tag{3}$$

An alternative to this process is the replacement of the electrolytic reduction by reducing agents that prevents the possible contamination with the reducing agent. This process has been adopted in countries like USA, Spain, Australia, Japan, Canada, England, South Africa and India (26; 27; 28; 29; 30; 31; 32). For the production of UF4 with nuclear purity from UO2F2 acids solutions, some fundamental stages are required such as obtaining the solution, reduction to uranium valence and precipitation of the formed U+4. These stages are shown

Uranium hexafluoride is a crystalline substance at normal pressure and temperature conditions. At the temperature of 900C under a pressure of 3kgf/cm2, UF6 becomes gas and

UF6 + H2O → UO2F2 + 4HF (3)

METALOTHERMIC REDUCTCTION

U **UF4** <sup>o</sup> **U3Si2**

**Metallurgical Process**

when it is injected into water, it hydrolyzes immediately according to the following:

PRECIPITATION

FILTRATION

WASHING

UO2F2

HYDROLYSIS

**UF6**

**Chemical Process**

in Figure 2, as schematized operations.

**3.1.2 Obtaining UO2F2 solutions** 

Fig. 2. Wet Process to produce UF4

atmosphere DRYING

hot H20

HF

SnCl2

inert


Table 2 shows the chemical characteristics of UO2F2 solution obtained from UF6 hydrolysis.

Table 2. Chemical characteristics of UO2F2 solution

#### **3.1.3 Chemical reduction of UF6 to UF4**

Uranium in its tetravalent state is very important in different technological processes. Essentially, the preparation process (aqueous way) from solutions containing uranyl ion (hexavalent) involves the reduction towards tetravalent state, and later precipitation as UF4 using HF solution. In aqueous solutions, these reductions can be carried out by chemical, electrochemical or photochemical methods.

All the trials for the preparation of UF4 using chemical reduction have been carried out using UO2F2 solution inside a stainless steel reactor, coated with Teflon. The solution has been heated under continuous stirring to reach a temperature set, and the reducing agent has been added. Next, the precipitating agent solution is slowly added to UO2F2 in solution with hydrofluoric acid (HF). Tests have been carried out using some reducing agents, such as SnCl2, CuCl, FeCl2, Na2S2O4.

$$\text{U}\text{O}\_2\text{F}\_2 + \text{SnCl}\_2 + 4\text{HF} \rightarrow \text{UF}\_4 + \text{SnClF}\_2 + 2\text{H}\_2\text{O} \tag{4}$$

$$\rm{UO\_2F\_2 + 4HF + Fe \to UF\_4 + FeF\_2 + 2H\_2O} \tag{5}$$

$$\text{UO}\_2\text{F}\_2 + \text{CuCl} + 4\text{HF} \rightarrow \text{UF}\_4 + \text{CuClF}\_2 + 2\text{H}\_2\text{O} \tag{6}$$

$$\text{U}\text{O}\_2\text{F}\_2 + \text{Na}\_2\text{S}\_2\text{O}\_4 + 2\text{HF} \rightarrow \text{UF}\_4 + \text{Na}\_2\text{SO}\_3 + \text{H}\_2\text{O} \tag{7}$$

Upon UF4 precipitation the suspension is left in rest up to reaching room temperature. After over 12 hours, it was performed the solid/liquid separation by vacuum filtration, washing and drying in a muffle kiln. The salts obtained were all identified as being uranium tetrafluoride. According to the results shown in Figure 3, it is evident that, from all used reducing agents, only SnCl2 and FeCl2 have shown significant results in regards of getting UF4. Nevertheless, SnCl2 is more consistent reducing agent at higher temperature of process.

The influence of the temperature upon UO2F2 and UO2 contents in obtained UF4 is shown in Figure 4. It was employed SnCl2 as the reducing agent in this study to precipitate UO2F2 solution. The residual moisture is dried at 130°C. The tin content in all obtained UF4 has shown to be in the range of 0.15 – 0.15%.

As shown previously, the process for obtaining UF4 by reduction precipitation using SnCl2 had the best results and achieved an yield of 98% of UF4 precipitation. The precipitation with HF solution is relatively slow and tends to accelerate as the temperature rises (17; 18). This is important, since it avoids excessive precipitate hydration and facilitates the

 UO2F2 + SnCl2 + 4HF → UF4 pp + SnCl2F2 + 2H2O (8) During the uranium processing stages, the goal is to achieve an end product with high purity and showing physical and chemical characteristics appropriate for the preparation of

Table 3 lists the suitable chemical and physical characteristics of UF4 for a later reduction to

at 130°C inert atmosphere at 400°C

Fe Cr Ni Mo Al Mn Cu Sn <20 <10 <10 <5 <10 <5 <5 0,1

sedimentation, filtration and drying operations. The full reaction is represented by:

Uranium (%) 74.20 75.0 Fluoride (%) 24.60 27.90 UF4 (%) 97.50 99.85 UO2F2 (%) 0.29 0.34 UO2 (%) 0.06 0.29 HF(%) 0.23 0.12 Moisture (%) 0.33 <0.03 Crystallization H2O 4.50 <100

Density (g/cm3 ) 6.70 Granulometry (m) 15.0

(m2/g) 0.21

metal (monel, inconel, nickel) which increases the cost of a plant.

Table 3. Chemical and Physical Properties of UF4 produced by an aqueous route

The UF4 obtained by reaction with UO2 with hydrofluoric acid is easily made. The reaction

 UO2 (s) + 4HF (aq) ↔ UF4 (s) + 2H2O (9) This process has some advantages over the other processes. Since the reaction occurs at low temperatures, the reactor can be constructed using materials as polyethylene, polypropylene or carbon steel with plastic coating, while other processes require equipment built with

In Figure 5, the x-ray diffractogram spectra are presented for UF4 produced by the method via NH4HF2 (bifluoride route) and by aqueous route. Typical SEM image of precipitated UF4 is presented in Figure 6. It displays a granular structure with relevant amount of porosity.

**3.1.4 Obtaining UF4** 

nuclear fuel.

obtain metallic uranium.

Met. Impurities

Specific Surface

**3.1.5 Preparation of UF4 from UO2** 

can be summarized as follows:

(µg/g)

Fig. 3. Influence of reducing agent as a function of obtaining UF4

Fig. 4. Influence of the temperature as a function of the contents of UO2 F2 and UO2 in UF4

#### **3.1.4 Obtaining UF4**

30 Radioisotopes – Applications in Physical Sciences

Fig. 3. Influence of reducing agent as a function of obtaining UF4

Fig. 4. Influence of the temperature as a function of the contents of UO2 F2 and UO2 in UF4

As shown previously, the process for obtaining UF4 by reduction precipitation using SnCl2 had the best results and achieved an yield of 98% of UF4 precipitation. The precipitation with HF solution is relatively slow and tends to accelerate as the temperature rises (17; 18). This is important, since it avoids excessive precipitate hydration and facilitates the sedimentation, filtration and drying operations. The full reaction is represented by:

$$\text{UO}\_2\text{F}\_2 + \text{SnCl}\_2 \uparrow + 4\text{HF} \rightarrow \text{UF}\_{4\text{FP}} + \text{SnCl}\_2\text{F}\_2 + 2\text{H}\_2\text{O} \tag{8}$$

During the uranium processing stages, the goal is to achieve an end product with high purity and showing physical and chemical characteristics appropriate for the preparation of nuclear fuel.


Table 3 lists the suitable chemical and physical characteristics of UF4 for a later reduction to obtain metallic uranium.

Table 3. Chemical and Physical Properties of UF4 produced by an aqueous route

#### **3.1.5 Preparation of UF4 from UO2**

The UF4 obtained by reaction with UO2 with hydrofluoric acid is easily made. The reaction can be summarized as follows:

$$\text{UO}\_2\text{(s)} + 4\text{HF (aq)} \leftrightarrow \text{UF}\_4\text{(s)} + 2\text{H}\_2\text{O}\tag{9}$$

This process has some advantages over the other processes. Since the reaction occurs at low temperatures, the reactor can be constructed using materials as polyethylene, polypropylene or carbon steel with plastic coating, while other processes require equipment built with metal (monel, inconel, nickel) which increases the cost of a plant.

In Figure 5, the x-ray diffractogram spectra are presented for UF4 produced by the method via NH4HF2 (bifluoride route) and by aqueous route. Typical SEM image of precipitated UF4 is presented in Figure 6. It displays a granular structure with relevant amount of porosity.

Fig. 6. SEM image of some UF4 particles, produced by the bifluoride(a) route e via SnCl2 (b).

The achievement of UF4 by this process was adopted in Canada, France, the former Czechoslovakia, South Africa, United States, Portugal, Brazil, Germany and Sweden (17;

The sequence of operations is to reduce UO3 by hydrogen, followed by treatment with HF

UO3 (s) + H2 (g) ↔ UO2 (s) + H2O (v) (10)

UO2 (s) + 4HF (g) ↔ F4 (s) + 2H2O (v) (11)

**3.2 Procedures for obtaining UF4 by dry process 3.2.1 Preparation of UF4 by fluorination of UO2** 

resulting UO2 anhydrous at atmospheric pressure.

21; 23).

Fig. 5. X-ray diffraction pattern of UF4 produced by the bifluoride route and from the aqueous route.

The UF4 fabrication using fluorination media with ammonium bifluoride is perfectly feasible. The ammonium bifluoride is a by-product effluent generated during the UF6 conversion to AUC1. UF4 obtained by this route has the same crystalline structure presented by the aqueous process, as demonstrated by the x-ray spectrum. Besides, it has the correct chemical and physical characteristics for metallothermic production of metallic uranium. Even presenting a lower relative tapped density; this property will not be a problem, because this is an alternative process that has as main goals the recovery of uranium, ammonium and the fluorides of the liquid effluents generated in the process of UF6 reconversion. This UF4 will be lately diluted in the UF4 charges produced by the aqueous route. The development of this process (bifluoride route) not only provides an efficient process for uranium recovery from secondary sources, as also eliminates the environmental pollution by discarding the bifluoride. It also provides a chemical compound with chemical and physical characteristics very similar to the aqueous route (SnCl2).

<sup>1</sup> Ammonium uranyl carbonate (UO2CO3·2(NH4)2CO3) is known in the uranium processing industry as AUC and is also called uranyl ammonium carbonate. Ammonium uranyl carbonate is one of the many forms called yellowcake in this case it is the product obtained by the heap leach process. This compound is important as a component in the conversion process of uranium hexafluoride (UF6) to uranium dioxide (UO2). In aqueous process uranyl nitrate is treated with ammonium bicarbonate to form ammonium uranyl carbonate as a solid precipitate and ammonium bifluoride as by-product (41).

32 Radioisotopes – Applications in Physical Sciences

Fig. 5. X-ray diffraction pattern of UF4 produced by the bifluoride route and from the

and physical characteristics very similar to the aqueous route (SnCl2).

The UF4 fabrication using fluorination media with ammonium bifluoride is perfectly feasible. The ammonium bifluoride is a by-product effluent generated during the UF6 conversion to AUC1. UF4 obtained by this route has the same crystalline structure presented by the aqueous process, as demonstrated by the x-ray spectrum. Besides, it has the correct chemical and physical characteristics for metallothermic production of metallic uranium. Even presenting a lower relative tapped density; this property will not be a problem, because this is an alternative process that has as main goals the recovery of uranium, ammonium and the fluorides of the liquid effluents generated in the process of UF6 reconversion. This UF4 will be lately diluted in the UF4 charges produced by the aqueous route. The development of this process (bifluoride route) not only provides an efficient process for uranium recovery from secondary sources, as also eliminates the environmental pollution by discarding the bifluoride. It also provides a chemical compound with chemical

1 Ammonium uranyl carbonate (UO2CO3·2(NH4)2CO3) is known in the uranium processing industry as AUC and is also called uranyl ammonium carbonate. Ammonium uranyl carbonate is one of the many forms called yellowcake in this case it is the product obtained by the heap leach process. This compound is important as a component in the conversion process of uranium hexafluoride (UF6) to uranium dioxide (UO2). In aqueous process uranyl nitrate is treated with ammonium bicarbonate to form ammonium uranyl carbonate as a solid precipitate and ammonium bifluoride as by-product

aqueous route.

(41).

Fig. 6. SEM image of some UF4 particles, produced by the bifluoride(a) route e via SnCl2 (b).

#### **3.2 Procedures for obtaining UF4 by dry process 3.2.1 Preparation of UF4 by fluorination of UO2**

The achievement of UF4 by this process was adopted in Canada, France, the former Czechoslovakia, South Africa, United States, Portugal, Brazil, Germany and Sweden (17; 21; 23).

The sequence of operations is to reduce UO3 by hydrogen, followed by treatment with HF resulting UO2 anhydrous at atmospheric pressure.

$$\text{UO}\_3\text{(s)} + \text{H}\_2\text{(g)} \leftrightarrow \text{UO}\_2\text{(s)} + \text{H}\_2\text{O}\text{(v)}\tag{10}$$

$$\text{UO}\_2\text{(s)} + 4\text{HF (g)} \leftrightarrow \text{F4 (s)} + 2\text{H}\_2\text{O (v)}\tag{11}$$

At this temperature, only 8 hours are necessary to promote the fluorination. The material is loaded into an aluminum container with calcium fluoride and heated inside a furnace. The furnace is fitted with a condensing tube with a relief valve, which releases the water and ammonia from the fluoridation reaction to a reservoir and retains the excess of sublimed

During the fluorination and/or decomposition, the formation of UO2F2 probably occurs. This is a significant happening, since it may reduce the efficiency of reduction in the next

In a second step of the process, under vacuum distillation, NH4UF5 is decomposed in UF4

NH4 UF5 ↔ UF4 + NH4F (18)

The UF4 can be prepared by reaction of ammonium fluoride or bifluoride with UO3

 3UO3 6NH4HF2 +9H2O ↔ 3UF4 +4NH3 +N2 (19) Although the United States have been among the first to study the process (34) Canada is

There are several possibilities to produce metallic uranium (41; 26; 42). Magnesiothermic reduction of UF4 is one of them and it is a known process since early 1940's (7; 8). The IPEN technology uses this route in 1970-80's for production 100kg ingots of natural uranium. For LEU U-production, it is necessary to handle safe mass (less than 2.2 kg U), to avoid possible criticality hazards. IPEN presently produces around of 1000g LEU ingots via magnesiothermic process and in future may produce 2000g or more. This range of uranium weight is rather small if compared to big productions of natural uranium. Metallic uranium is reported (9) to be produced with 94% metallic yield when producing bigger quantities. The magnesiothermic process downscaling to produce LEU has small possibilities to achieve this higher metallic yield. This is due to the design of crucibles, with relatively high proportion of surrounding area, which is more prone to withdraw evolved heat from the exothermic reaction during uranium reduction. Normally, calciothermic reduction of UF4 is preferred worldwide, since the exothermic heat is higher (-109.7 kcal/mol) compared to smaller amount of -49.85 kcal/mol using magnesium as the reducer (10). Nevertheless, IPEN chose magnesiothermic because it is easier to be done, avoiding no handling of toxic and pyrophoric calcium. Moreover, the magnesiothermic process is cheaper, so, it brings economical compensation for its worse metallic yield than calcium reduction process. In addition, the recycling of slag and operational rejects is highly efficient and there are

 UF4 + 2Mg = U + 2MgF2 ΔH= - 49.85 kcal/mol (at 640°C) (20) As magnesium thermodynamics is less prompt to ignite than calcium, the batch reactor is heated up to the temperature around 640°C. The routine shows that this ignition normally

**3.2.4. Preparation of UF4 by the reaction of ammonium bifluoride with UO3** 

bifluoride.

with the NH4F by this reaction:

according to the equation:

the country that developed this process (35; 36)

virtually insignificant LEU uranium is lost (23). The magnesiothermic reaction is given by:

**4. Production of metallic uranium** 

step.

The reduction of UO2 is performed at temperatures of 500-700oC. Another alternative is the reduction of U3O8 recommended when you have storage problems UO3, being extremely hygroscopic.

$$\text{U}\_{\text{3}}\text{O}\_{8}\text{(s)} + 2\text{H}\_{2}\text{(g)} \leftrightarrow 3\text{U}\text{O}\_{2}\text{(s)} + 2\text{H}\_{2}\text{O}\text{(v)}\tag{12}$$

In such a process is commonly used the moving bed or fluidized bed reactor type. Preparation of UF4 by reaction of the UO3 with NH3 and HF gaseous The process consists of only one step to produce UF4. The mixture consisting of NH3 and HF is treated with UO3 at 500-700oC. This reaction is fast and produces high purity UF4:

$$\text{3UO}\_3 + 2\text{NH}\_3 + 12\text{ HF} \leftrightarrow \text{9H}\_2\text{O} + \text{N}\_2 + \text{3UF}\_4 \tag{13}$$

The UF4 fabrication by the reaction of uranium oxides with fluorinated hydrocarbons (freon) is as follows:

$$\text{2CF}\_2\text{Cl}\_2 + \text{UO}\_3 \leftrightarrow \text{UF}\_4 \ + \text{CO}\_2 + \text{Cl}\_2 + \text{COCl}\_2 \tag{14}$$

The literature shows results of reactions of different freons with uranium oxides UO2, U3O8 and UO3 (27; 29; 33). The reactors used in this process cannot be constructed using nickel, copper, platinum and stainless steel, since they undergo chemical attack of reagents, besides this reaction promotes pyrolysis under carbon presence. The reactors are constructed with graphite or calcium fluoride, which may cause contamination to the obtained UF4. The advantages of this method are equipment simplicity and the possibility of applying this reaction to all the uranium oxides.

#### **3.2.2 Preparation of UF4 from metallic uranium or uranium hydride (UH3)**

By fluoridation at high temperatures uranium metal can be quickly converted into uranium tetrafluoride by the reaction below:

$$\text{U + 3 / 2H\_2 \leftrightarrow \text{UH}\_3} \tag{15}$$

$$\text{UH}\_3 + \text{4HF} \overset{200^\circ \text{C}}{\leftrightarrow} \text{UF}\_4 + 7 / 2\text{H}\_2 \tag{16}$$

Uranium metal is industrially manufactured from UF4. In the absence of advantage in obtaining first elemental uranium and transform it into UH3, then get to UF4.

#### **3.2.3 Procedures for obtaining UF4 by dry ammonium bifluoride with (NH4HF2)**

The fluorination of UO2 is made with NH4HF2, a white solid; it has low vapor pressure and can be operated freely since it is non-toxic. Initially, UO2 is mixed with bifluoride, 20% above the stoichiometric amount. The bifluoride crystal is easily crushed and the mixture of UO2 + NH4 HF2 is made in a monel 400 container to prevent contamination.

The conversion of bifluoride at room temperature occurs after approximately 24 hours, although under such conditions the water formed in the reduction may be retained in the precipitate. The elimination of NH3 and water is facilitated by the reaction of UO2 and NH4HF2 at 150°C:

$$\text{2UO}\_2 \text{+ 5NH}\_4\text{HF}\_2 \leftrightarrow \text{3NH}\_3 + \text{4H}\_2\text{O} + \text{2NH}\_4\text{UF}\_5 \tag{17}$$

The reduction of UO2 is performed at temperatures of 500-700oC. Another alternative is the reduction of U3O8 recommended when you have storage problems UO3, being extremely

U3O8 (s) + 2H2 (g) ↔ 3UO2 (s) + 2H2O (v) (12)

The process consists of only one step to produce UF4. The mixture consisting of NH3 and HF

 3UO3 + 2NH3 + 12 HF ↔ 9H2 O + N2 + 3UF4 (13) The UF4 fabrication by the reaction of uranium oxides with fluorinated hydrocarbons (freon)

 2CF2 Cl2 + UO3 ↔ UF4 + CO2 + Cl2 + COCl2 (14) The literature shows results of reactions of different freons with uranium oxides UO2, U3O8 and UO3 (27; 29; 33). The reactors used in this process cannot be constructed using nickel, copper, platinum and stainless steel, since they undergo chemical attack of reagents, besides this reaction promotes pyrolysis under carbon presence. The reactors are constructed with graphite or calcium fluoride, which may cause contamination to the obtained UF4. The advantages of this method are equipment simplicity and the possibility of applying this

By fluoridation at high temperatures uranium metal can be quickly converted into uranium

<sup>o</sup> 200 C

Uranium metal is industrially manufactured from UF4. In the absence of advantage in

The fluorination of UO2 is made with NH4HF2, a white solid; it has low vapor pressure and can be operated freely since it is non-toxic. Initially, UO2 is mixed with bifluoride, 20% above the stoichiometric amount. The bifluoride crystal is easily crushed and the mixture of

The conversion of bifluoride at room temperature occurs after approximately 24 hours, although under such conditions the water formed in the reduction may be retained in the precipitate. The elimination of NH3 and water is facilitated by the reaction of UO2 and

2UO2 + 5NH4 HF2 ↔ 3NH3 + 4H2O + 2NH4 UF5 (17)

<sup>o</sup> 250 C

U 3 /2H2 3 + ↔ UH (15)

UH 4HF 3 42 + ↔+ UF 7 / 2H (16)

In such a process is commonly used the moving bed or fluidized bed reactor type.

is treated with UO3 at 500-700oC. This reaction is fast and produces high purity UF4:

**3.2.2 Preparation of UF4 from metallic uranium or uranium hydride (UH3)** 

obtaining first elemental uranium and transform it into UH3, then get to UF4.

UO2 + NH4 HF2 is made in a monel 400 container to prevent contamination.

**3.2.3 Procedures for obtaining UF4 by dry ammonium bifluoride with (NH4HF2)** 

Preparation of UF4 by reaction of the UO3 with NH3 and HF gaseous

hygroscopic.

is as follows:

reaction to all the uranium oxides.

tetrafluoride by the reaction below:

NH4HF2 at 150°C:

At this temperature, only 8 hours are necessary to promote the fluorination. The material is loaded into an aluminum container with calcium fluoride and heated inside a furnace. The furnace is fitted with a condensing tube with a relief valve, which releases the water and ammonia from the fluoridation reaction to a reservoir and retains the excess of sublimed bifluoride.

During the fluorination and/or decomposition, the formation of UO2F2 probably occurs. This is a significant happening, since it may reduce the efficiency of reduction in the next step.

In a second step of the process, under vacuum distillation, NH4UF5 is decomposed in UF4 with the NH4F by this reaction:

$$\text{NH}\_4\text{ UF5}\_5 \leftrightarrow \text{UF4} + \text{NH}\_4\text{F} \tag{18}$$

#### **3.2.4. Preparation of UF4 by the reaction of ammonium bifluoride with UO3**

The UF4 can be prepared by reaction of ammonium fluoride or bifluoride with UO3 according to the equation:

$$\text{3UO}\_3\text{6NH}\_4\text{HF}\_2 + \text{9H}\_2\text{O} \leftrightarrow \text{3UF}\_4 + \text{4NH}\_3 + \text{N}\_2\tag{19}$$

Although the United States have been among the first to study the process (34) Canada is the country that developed this process (35; 36)

#### **4. Production of metallic uranium**

There are several possibilities to produce metallic uranium (41; 26; 42). Magnesiothermic reduction of UF4 is one of them and it is a known process since early 1940's (7; 8). The IPEN technology uses this route in 1970-80's for production 100kg ingots of natural uranium. For LEU U-production, it is necessary to handle safe mass (less than 2.2 kg U), to avoid possible criticality hazards. IPEN presently produces around of 1000g LEU ingots via magnesiothermic process and in future may produce 2000g or more. This range of uranium weight is rather small if compared to big productions of natural uranium. Metallic uranium is reported (9) to be produced with 94% metallic yield when producing bigger quantities. The magnesiothermic process downscaling to produce LEU has small possibilities to achieve this higher metallic yield. This is due to the design of crucibles, with relatively high proportion of surrounding area, which is more prone to withdraw evolved heat from the exothermic reaction during uranium reduction. Normally, calciothermic reduction of UF4 is preferred worldwide, since the exothermic heat is higher (-109.7 kcal/mol) compared to smaller amount of -49.85 kcal/mol using magnesium as the reducer (10). Nevertheless, IPEN chose magnesiothermic because it is easier to be done, avoiding no handling of toxic and pyrophoric calcium. Moreover, the magnesiothermic process is cheaper, so, it brings economical compensation for its worse metallic yield than calcium reduction process. In addition, the recycling of slag and operational rejects is highly efficient and there are virtually insignificant LEU uranium is lost (23).

The magnesiothermic reaction is given by:

$$\text{UF}\_4 + 2\text{Mg} = \text{U} + 2\text{MgF}\_2\\\text{\AA H= - 49.85 kcal/mol (at 640°C)}\tag{20}$$

As magnesium thermodynamics is less prompt to ignite than calcium, the batch reactor is heated up to the temperature around 640°C. The routine shows that this ignition normally

(a)

(b)

(c)

Fig. 7. Sequence of UF4+Mg charging in IPEN's magnesiothermic method to produce metallic uranium. (a) 10 layer preparation of UF4 (green) and Mg (metallic bright); (b)

blending of material; (c) full charge after tapping the 10 layers.

happens some degrees bellow this temperature (9). Nevertheless, several reactions may occur during heating of the UF4+Mg load. Moisture is normally present in the charge, either caught during UF4 handling after drying or during crucible charging. During heating, as the temperature crosses the water boiling point (>100°C), all moisture becomes water vapor. This vapor not only bores its passage through the load but easily oxidize the reactants in this pathway by the following reactions (30):

$$\text{UF4} + 2\text{H}\_2\text{O} \rightarrow \text{UO}\_2 + 4\text{HF} \tag{21}$$

$$\text{2UF}\_4 + \text{2H}\_2\text{O} \rightarrow \text{2UO}\_2\text{F}\_2 + \text{4HF (via UF}\_3\text{(OH) and UOF}\_2\text{ steps)}\tag{22}$$

As the loading of the charge is not fully sealed to avoid atmosphere contact, some O2 is entrapped in the system, leading also to reactants oxidation by:

$$\rm{2UF\_4 + O\_2 \to UF\_6 + UO\_2F\_2} \tag{23}$$

Producing some UF6 that transforms into UO2F2 by the following reaction:

$$\text{UF}\_6 \text{ +2H\_2O} \rightarrow \text{UO}\_2\text{F}\_2 + 4\text{HF} \tag{24}$$

and also occurring magnesium oxidation (very fast above 620°C) by:

$$2\text{Mg} + \text{O}\_2 \rightarrow 2\text{MgO} \tag{25}$$

The presence of the UO2 and UO2F2 in the produced UF4 accumulates with previous oxidized ones during the dehydration. All these compounds formation worsens the metallic yield of uranium production.

In this work, it is discussed the effect of LEU UF4 precipitated via hydrolyzed UF6 and its potential variability in reactivity. The chemical UO2F2 residual content in dried UF4 is also analyzed for its potential relevance in the uranium production. The tapped density of dehydrated and loaded UF4 is also commented as affecting the reactivity process of uranium production. The magnesiothermic ignition is also analyzed since the heating time of the charge may affect the reactivity of the load. The reaction sequence after ignition is theoretically proposed as a possible sequence of chemical and physical events. The evidences in the slag solidification on crucible wall, during the reaction process to reduce UF4 towards U°, is very enlightening to guide towards the interpretation of the reaction blast.

The IPEN's magnesiothermic reduction process of UF4 to metallic uranium (in the range of 1000g) could be synthesized as:


happens some degrees bellow this temperature (9). Nevertheless, several reactions may occur during heating of the UF4+Mg load. Moisture is normally present in the charge, either caught during UF4 handling after drying or during crucible charging. During heating, as the temperature crosses the water boiling point (>100°C), all moisture becomes water vapor. This vapor not only bores its passage through the load but easily oxidize the reactants in this

UF4 + 2H2O → UO2 + 4HF (21)

 2UF4 + 2H2O → 2UO2F2 + 4HF (via UF3(OH) and UOF2 steps) (22) As the loading of the charge is not fully sealed to avoid atmosphere contact, some O2 is

2UF4 + O2 → UF6 + UO2F2 (23)

UF6 +2H2O → UO2F2 + 4HF (24)

The presence of the UO2 and UO2F2 in the produced UF4 accumulates with previous oxidized ones during the dehydration. All these compounds formation worsens the metallic

In this work, it is discussed the effect of LEU UF4 precipitated via hydrolyzed UF6 and its potential variability in reactivity. The chemical UO2F2 residual content in dried UF4 is also analyzed for its potential relevance in the uranium production. The tapped density of dehydrated and loaded UF4 is also commented as affecting the reactivity process of uranium production. The magnesiothermic ignition is also analyzed since the heating time of the charge may affect the reactivity of the load. The reaction sequence after ignition is theoretically proposed as a possible sequence of chemical and physical events. The evidences in the slag solidification on crucible wall, during the reaction process to reduce UF4 towards U°, is very

The IPEN's magnesiothermic reduction process of UF4 to metallic uranium (in the range of

1. In preparation for the mass reduction of a single batch, it is used with a standard charge of reactants of 1815 ± 5g of the mixture Mg + UF4 (1540 ± 1g LEU UF4) containing 15% excess of stoichiometric Mg content. For purpose of homogenization, the charge of UF4 + Mg is divided into 10 layers, which are tapped one by one inside the crucible. All this operation is carried out inside a glovebox to prevent nuclear contamination. This

2. After placing the reactants inside the graphite crucible, a variable amount of CaF2 is tapped over the UF4+Mg load in the crucible to fully complete the reaction volume. This amount is dependent on tapped density and UF4+Mg blending, which varies in function to UF4 fabrication. The crucible is made of fully machined graphite volume with enough resistance to produce safe nuclear uranium amount around 1000g. This crucible was designed to withstand the blast impact of metallothermic reaction, as well as thermal cycles of heating and cooling without excessive wear in order to be used in several batches.

2Mg + O2 → 2MgO (25)

pathway by the following reactions (30):

yield of uranium production.

1000g) could be synthesized as:

sequence is illustrated in Figure 7.

entrapped in the system, leading also to reactants oxidation by:

Producing some UF6 that transforms into UO2F2 by the following reaction:

and also occurring magnesium oxidation (very fast above 620°C) by:

enlightening to guide towards the interpretation of the reaction blast.

(b)

(c)

Fig. 7. Sequence of UF4+Mg charging in IPEN's magnesiothermic method to produce metallic uranium. (a) 10 layer preparation of UF4 (green) and Mg (metallic bright); (b) blending of material; (c) full charge after tapping the 10 layers.

3. After closed with the top cover, the crucible is inserted inside a stainless steel cylindrical reactor vessel, made of ANSI 310, which allows argon fluxing during batch processing (1 L/min with 2 kgf/cm2 of pressure). As shown in Fig. 8 (a-b), the whole crucible + reactor are placed in resistor pit furnace with four programmable

4. The reaction vessel is set to heat up to 620°C. At this level, the reaction ignition is expected. The total heating time and waiting for ignition is about 180 minutes from heat

5. The reaction of UF4 with Mg produces an intense exothermic heat release inside the crucible. It is considered as an adiabatic reaction. It produces metallic uranium and MgF2 slag in liquid form. Both products deposit in the crucible bottom are easily taken apart after opening the crucible. Some products project over the crucible wall and freeze

6. This full reaction happens in a noticeable time between 800 and 1200ms from ignition to final deposit. This control is measured by sound waves, using an

7. After the reaction, 10 minutes is awaited for full solidification of reaction products inside the furnace. Then the furnace is turned off and the reactor vessel is lifted out of the furnace. There is a 16 hours for cooling before its opening. This avoids firing of

8. The disassembling of reduction set is performed inside a glove box. The top and bottom covers of the crucible are removed. By means of rubber soft hammering, it is able to withdraw the uranium ingot. The MgF2 slag is removed by mechanical cleaning. The metallic uranium is pickled in nitric acid 65%vol and the final mass of metallic uranium

The intermetallic U3Si2 is produced from metallic uranium (47). This alloy is produced from a uranium ingot and hyperstoichiometric silicon addition (7.9% Si). The induction furnace (15 kW) should be submitted to 2.10-3 mbar vacuum and flushed with argon-atmosphere. Then the melting is carried out. The blend is molten inside an induction furnace using zirconia crucible reaching more than 1750°C, as this intermetallic requests this level of temperature to be properly homogenized before solidification. No other crucibles, than a zirconia one could bear the aggressive environment created by uranium attack on linings. The load arrangement of uranium and silicon, as shown in Fig. 9, is then charged inside the crucible. It was planned to help the sequence of melting during the several stages that passes the alloy formation until reaching the final intermetallic composition. The quality of this intermetallic produced in this way normally meets the requirements as nuclear material. The X-ray diffractogram (Fig. 9) confirms the necessary proportion of phases presents in the produced powder of this alloy, which should be more than 80wt% of crystalline phases. As rule of thumb, the chemical amounts of boron, cadmium, cobalt, lithium should be less than 10µg/g individually. The other may reach hundreds of µg/g up 1000 µg/g. Carbon could reach up to 2000 µg/g. Isotopic concentration of 235U is 19.75±0.20wt%. The required density

zones having the possibility of raising the temperature up to 1200°C.

time to temperature setting point.

metallic uranium in contact with atmosphere.

**5. Production of uranium silicide** 

is measured and its density evaluated by Archimedes' method.

there.

is 11.7g/cm3.

accelerometer.

Fig. 8. (a) Schematic drawing of pit furnace, reactor vessel and crucible; (b) Charging of the reactor vessel inside the pit furnace; (c) Raw metallic uranium and upper deposited slag after removing from the crucible; (d) Metallic uranium after cleaning.

(a) (b)

(c) (d)

Fig. 8. (a) Schematic drawing of pit furnace, reactor vessel and crucible; (b) Charging of the reactor vessel inside the pit furnace; (c) Raw metallic uranium and upper deposited slag

after removing from the crucible; (d) Metallic uranium after cleaning.


#### **5. Production of uranium silicide**

The intermetallic U3Si2 is produced from metallic uranium (47). This alloy is produced from a uranium ingot and hyperstoichiometric silicon addition (7.9% Si). The induction furnace (15 kW) should be submitted to 2.10-3 mbar vacuum and flushed with argon-atmosphere. Then the melting is carried out. The blend is molten inside an induction furnace using zirconia crucible reaching more than 1750°C, as this intermetallic requests this level of temperature to be properly homogenized before solidification. No other crucibles, than a zirconia one could bear the aggressive environment created by uranium attack on linings. The load arrangement of uranium and silicon, as shown in Fig. 9, is then charged inside the crucible. It was planned to help the sequence of melting during the several stages that passes the alloy formation until reaching the final intermetallic composition. The quality of this intermetallic produced in this way normally meets the requirements as nuclear material. The X-ray diffractogram (Fig. 9) confirms the necessary proportion of phases presents in the produced powder of this alloy, which should be more than 80wt% of crystalline phases. As rule of thumb, the chemical amounts of boron, cadmium, cobalt, lithium should be less than 10µg/g individually. The other may reach hundreds of µg/g up 1000 µg/g. Carbon could reach up to 2000 µg/g. Isotopic concentration of 235U is 19.75±0.20wt%. The required density is 11.7g/cm3.

The reference industrial process to produce plate-type fuel involves roll-milling together the fissile core, or fuel meat (a blend of an uranium compound and aluminum powders), and the cladding (aluminum alloy plates). This process can draw on considerable feedback from experience, since nearly all research reactors use this type of fuel. The process has seen large-scale implementation with NUKEM, in Germany, UKAEA, in the United Kingdom,

In general, the MTR type fuel element fabrication process using silicide (U3Si2) can be divided into the following main steps: hydrolysis of UF6 through its reaction with water; production of uranium tetrafluoride (UF4); production of metallic uranium; U3Si2 powder production from uranium metal; production of fuel cores from U3Si2 and aluminum powders; production of fuel plates with U3Si2-Al dispersion; assembling of fuel elements;

The simplified block diagram of the fabrication process for silicide fuel elements is shown in Figure 10. The manufacturing process of the fuel begins with the UF6 processing. The UF6 is enriched to 19,75 wt% 235U, a enrichment level that categorize the fuel as LEU (low enriched

uranium). Bellow the main stages of manufacture of such fuel are discussed.

**6. Production of MTR nuclear fuel** 

CERCA, in France, and Babcock, in the United States.

Fig. 10. Fabrication process of silicide fuel elements.

recovery of uranium; effluent treatment; quality control.

Fig. 9. Crucible arrangement of before melting to produce the intermetallic. U3Si2 product and its x-ray diffractogram results compared to CERCA product and JPDF 47-1070 for pure U3Si2.
