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Advanced thermal denitration conversion processes for aqueous-based reprocessing and recycling of spent nuclear fuels
Advanced thermal denitration conversion processes for aqueous-based reprocessing and recycling of spent nuclear fuels
12
Emory D. Collins Oak Ridge National Laboratory, Oak Ridge, TN, USA
Acronyms ADU ASTM MDD MH MOX NOx ORNL RD&D UNH
ammonium diuranate (process) American Society for Testing and Materials modified direct denitration (process) microwave heating mixed oxide nitrogen oxide Oak Ridge National Laboratory research, development, and demonstration uranyl nitrate hexahydrate
12.1
Introduction
The aqueous nitrate solutions of uranium and plutonium resulting from the reprocessing of spent nuclear fuels must be denitrated and converted to oxide particles with good ceramic properties for fabrication of recycle fuels. Several processes are suitable for this purpose, including those that utilize precipitation: filtration-calcination, such as the ammonium diuranate (ADU) process; the plutonium oxalate precipitation process using plutonium in either the trivalent or the tetravalent state; the uranium or plutonium peroxide precipitation process; or the ammonium-uranium carbonate process. Also, the use of direct thermal denitration processes is possible and is the subject of this chapter.
12.2
History and concepts for improvement
Thermal denitration has been used for many years in the processing and recovery of uranium from mining ores in the United States. Simple thermal denitration was used in large-scale operations to convert aqueous uranyl nitrate hexahydrate (UNH) solution to a solid oxide (uranium trioxide: UO3) interim product, usually for subsequent conversion to uranium fluorides UF4 or UF6 (Haas, 1988). The conversion of UNH to UO3 is a two-step process in which the aqueous solution is first evaporated and Reprocessing and Recycling of Spent Nuclear Fuel. http://dx.doi.org/10.1016/B978-1-78242-212-9.00012-5 © 2015 Elsevier Ltd. All rights reserved.
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concentrated to eliminate most of the water and produce a syrupy liquid, followed by calcination and decomposition (denitration) to produce a solid caked oxide product. In the calcination step, decomposition of the UNH begins at temperatures less than 300 °C and is completed by 400-500 °C. These operations have been performed in heated pots, fluidized beds, and stirred troughs, either in batch or in continuous mode. In European refineries, the denitration and conversion is accomplished by means of the addition of ammonia to the UNH solution to precipitate ADU, which is then filtered, dried, and calcined to form uranium trioxide powder. The use of direct thermal denitration eliminates multiple precipitation-filtrationcalcination steps and enables the use of less-complicated process equipment (Lerch and Norman, 1984). Direct thermal denitration, however, does not provide any separation of impurity metals and relies on previous chemical separations to provide the required product purity. Although direct thermal denitration offers significant process simplification, the process has not been used widely due to the formation of a “mastic” phase during melting and decomposition of the metal salts, leading to poor handling characteristics and poor ceramic properties of the product oxide. Since the 1980s, research, development, and demonstration (RD&D) studies in Japan and the United States have focused on direct thermal denitration processes to prepare uranium oxide and mixed oxides (MOXs) for recycle nuclear fuel fabrication (Koizumi et al., 1983; Haas et al., 1981). The RD&D studies have been aimed at the use of more uniform heating equipment or the addition of modifier agents to avoid the mastic phase formation and to enable suitable product quality. A major breakthrough in this area occurred when Haas et al. (1981) discovered that, by introducing ammonium nitrate into a uranium solution, the mastic phase is eliminated by the formation and denitration of double salts of ammonium-uranium nitrate. The process is called modified direct denitration (MDD). Current reprocessing and recycle fuel fabrication operations rely on the production of separated uranium oxide and plutonium oxide and the use of ball-milling methods to produce a suitable MOX for use in recycle fuels. Separated plutonium is viewed as a safeguards concern that increases the “attractiveness level” (Bathke et al., 2009) for diversion of the fissile material to weapons of mass destruction. Therefore, coprocessing and co-conversion methods are being developed in several countries for use in future reprocessing and recycle fuel fabrication plants. A microwave heating (MH) process has been developed and demonstrated in Japan for codenitration of uranium-plutonium solutions without formation of the mastic phase to produce a homogeneous MOX. Both the Japanese MH process and the U.S. MDD process are suitable for codenitration. Description of these processes is the focus of this chapter.
12.3
Process chemistry
Thermal denitration of uranium and plutonium, either individually or from mixed uranium-plutonium, is believed to be represented by the following approximate calcination reactions: UO2 ðNO3 Þ2 6H2 O ! UO3 + 1:86NO2 + 0:14NO + 0:57O2 + 6H2 O
(12.1)
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PuðNO3 Þ4 5H2 O ! PuO2 + 3:4NO2 + 0:6NO + 1:3O2 + 5H2 O
(12.2)
The denitration begins at a temperature <300 °C and is completed at 350-400 °C. The reaction can be performed in a batch mode, as in the Japanese MH process, or in a continuous mode, as in the U.S. MDD process. The Japanese MH process proceeds through a sequence of four steps as the temperature is increased (Numao et al., 2007): 1. The mixed uranium-plutonium nitrate solution is concentrated at 120 °C, the azeotropic temperature of nitric acid and water. 2. Decomposition (denitration) of the fused salt occurs between 120 and 250 °C, during which nitrogen oxide (NOx) gases are generated, recovered by scrubbing of the off-gas, and subsequently recycled. 3. Above 250 °C, residual moisture evaporates and NOx evolves from the decomposition of uranium. 4. The denitration is completed at about 350 °C, and the temperature begins to drop as the UO3 begins to decompose to U3O8. The temperature at that point in the process is detected by means of an infrared thermometer, and the MH is stopped to prevent excessive decomposition of the UO3 product.
The reaction sequence occurs in a time frame of 30-40 min per batch. Slower heating was found to result in localized heating and temperature increase, creating a product powder that is less active and less suitable for MOX fuel fabrication (Numao et al., 2007). The process chemistry of the MDD process was studied by Notz and Haas (1981) for the denitration of the ammonium-uranium nitrate double salt, which proceeds through a sequence of reactions (Equations 12.3–12.5) as illustrated by the results of a thermogravimetric analysis shown in Figure 12.1.
171
(NH4)2UO2(NO3)4 . 2H2O
160
(NH4)2UO2(NO3)4 (160.6 mg)
60 NH4UO2(NO3)3 (137.4 mg)
137
40
Actual weight (mg)
Relative weight (mg)
80
20 UO3 (82.9 mg)
0
0
50
100
200 300 Temperature (⬚C)
400
84
Figure 12.1 Thermogravimetric analysis of (NH4)2UO2(NO3)42H2O (initial sample weight, 171 mg).
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The thermal decomposition proceeds in three distinct steps: 1. Dehydration at 40 °C: ðNH4 Þ2 UO2 ðNO3 Þ4 2H2 O ! ðNH4 Þ2 UO2 ðNO3 Þ4 + 2H2 O
(12.3)
2. Loss of 1 mole of ammonium nitrate (NH4NO3) at 170 °C ðNH4 Þ2 UO2 ðNO3 Þ4 ! NH4 UO2 ðNO3 Þ3 + N2 O + 2H2 O
(12.4)
3. Conversion to UO3 at 270 °C: NH4 UO2 ðNO3 Þ3 ! UO3 + ðN2 O5 Þ + N2 O + 2H2 O
(12.5)
The decomposition products for the latter two reactions were inferred on the basis that NH4NO3 decomposes under controlled conditions to H2O and N2O, which is a standard method for the preparation of N2O. However, a differential thermal analysis, shown in Figure 12.2, and effluent gas analysis, shown in Table 12.1, indicated that a combination of several other reactions, which are known to occur for decomposition of NH4NO3, are involved. Some are exothermic, but the overall effect is endothermic with a large endotherm occurring immediately after the exotherm, as indicated in Figure 12.2.
12.4
Process equipment and operation
The Japanese MH process equipment is centered on multiple-batch MH units, each contained in a glove box. Each batch of uranium-plutonium solution is loaded into a 50 cm diameter, 6 cm deep round denitration dish, which is constructed of Si3N4; a ceramic material that transmits the microwave energy uniformly from the surface to the bottom Figure 12.2 Differential thermal analyses. Upward deflection ¼ exothermic effect; downward deflection ¼ endothermic effect.
Reaction (3)
Reactions (2) and (3)
Reactions (1) and (2) and (3)
100 200 Temperature (°C)
300
Advanced thermal denitration conversion processes
Table 12.1 Mass no.
317
Effluent gas analysis Species
Relative quantity
Temperature of maximum rate (°C) 320
185
280
185
290
Decomposition products
14 15 16 17 18 28 30 32 44 46
N+ NH+ NH+2 NH+3 H2O+ N+2 NO+ O+2 N2O+ NO+2
UO2 (NO3)2 m m m m M m
NH4NO3 t M M M M
m M
NH4UO2 (NO3)3 M m m m M m M M M M
(NH4)2UO2
(NO3)4
m M M M m
M M m
m
M m M M M M
Abbreviations: m, minor component; t, trace component; M, major component.
of the solution. In addition, the denitration dish is slowly rotated on a turntable inside the glove box to heat the solution uniformly. Active tests have been made of full-size production units at the Rokkasho Reprocessing Plant (Numao et al., 2007). Uranium nitrate solution and plutonium nitrate solution are received from the chemical purification facility and are blended at a mass ratio of 1:1 U:Pu for codenitration and conversion to UO3-PuO2. The denitrated powder has a porous structure due to foam formation by evolution of NOx gas during boiling. Bulk density of the product is typically about 2 g/cm3. Accessory equipment includes a microwave generation and wave guide transmission to each glove box, plus process off-gas treatment for recovery and recycle of nitric acid. The capacity of the MH system at Rokkasho is 108 kg (U + Pu) per day (Numao et al., 2007). The MDD process equipment is centered on one or more rotary kilns, such as the development unit illustrated in Figure 12.3. Each kiln is basically a straight length of stainless steel tubing with a diameter selected for the capacity of oxide production needed. Typically, the kiln is mounted on steel wheels and rotates within a tube furnace, which surrounds about one-third or more of the pipe length. Feed solution is introduced at one end of the rotary pipe kiln, which is angled downward to enable the decomposing salt and solid oxide powder to move by a combination of gravity and tube rotation to the end of the kiln, where the oxide product is collected. One or more solid stainless steel rods are located within the kiln and roll freely to act as “breaker bars” to crush lumps or to remove solid oxide from the tube wall. An air purge gas enters the kiln at the powder discharge end and
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TE indicates thermocouple
220 V TC indicates temperature controller
All internal materials: stainless steel 2500-W tube furnace
TC TE
Scraper 3.5-in. IPS Sch 10 pipe Support wheels Solution in gases out TE
1.91-cm. (3/4-in.) O.D. tubing
TE
Variable-speed motor with chain and sprocket drive
Breaker bars
Solid product collection
3⬚
Figure 12.3 Rotary kiln cross section.
flows countercurrently through the kiln to remove the water, acid, and NOx gases that are formed. The off-gases leave the kiln at the liquid feed end and are routed to a scrubber and condenser system for recovery and recycle of nitric acid and for removal of entrained oxide. Most of the development work has been done with uranium in rotary kilns measuring up to 16 cm in diameter and has progressed to demonstration at the multi-kilogram-per-hour scale. More recently, small-scale units (Figure 12.4) have been built and installed in a glove box to permit testing of the process for denitration of ammonium-uranium-plutonium nitrate solutions and co-conversion to MOX powder at a rate of 100 g/h.
Off-gas
Condenser Feed tank Scrubber
Air purge Pump
Condensate collection Rotary kiln furnace Oxide product
Figure 12.4 The U.S. MDD process and small-scale experimental equipment.
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Coprocessed uranium-plutonium nitrate solution is mixed with ammonia gas or ammonium nitrate solution to provide the desired ratio of NH+4 to heavy metal, typically 2:1. The resulting ammonium-heavy metal double salt solution is fed to the rotary kiln at the rate needed for the production capacity. Inside the kiln, the solution goes through the sequence of (i) evaporation of water and nitric acid, (ii) denitration of the fused double salt, and (iii) formation of mixed uranium-plutonium powder. With the tube furnace operating at a typical temperature of 650 °C, measurements of the temperature profile and decomposition gas concentrations within the kiln have indicated that the denitration is completed within a short distance from the liquid feed entry point.
12.5
Conversion of UO3 to UO2
Thermal denitration of uranium produces a UO3 product; codenitration of mixed uranium and plutonium produces a mixed UO3-PuO2 product. Subsequently, a separate calcination and reduction step is required to convert the UO3 to UO2 or the UO3-PuO2 to a single-phase (U,Pu)O2 solid solution. The UO3 can be converted to UO2 by means of the following calcination and reduction reactions: 3UO3 ! U3 O8 + 0:5O2
(12.6)
U3 O8 + 2H2 ! 3UO2 + 2H2 O
(12.7)
The calcination occurs at temperatures up to 700 °C. The reduction reaction occurs at a temperature of 700-800 °C in a period of 4 h. Alternatively, by heating in a reducing atmosphere, uranium trioxide (UO3) can be directly converted to uranium dioxide (UO2). For the Japanese MH process at the Rokkasho plant, the dry MOX product from the denitration dish is crushed and placed into a rotary calcination furnace. Then, the calcined powder is transferred to a rotary reduction furnace, where it is treated with 95% N2-5% H2 to convert the powder to the final MOX form (U,Pu)O2 (Numao et al., 2007). The oxide powder produced in the MDD rotary kiln leaves the kiln in an agglomerated form, which requires screening through a 40-mesh sieve before calcination and reduction. Although not extensively developed, the calcination-reduction process and equipment would likely be similar to the process and equipment used for the MH process at Rokkasho. X-ray diffraction examination of MDD MOX has determined that the oxides in the MOX form a solid solution during reduction (Walker et al., 2007).
12.6
Product characteristics
Table 12.2 shows the endothermic heat of decomposition and typical UO2 product characteristics from direct thermal denitration in comparison with those from the MDD process.
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Comparison of typical oxide product characteristics produced by MDD and direct denitration
Table 12.2
Heat of decomposition, DH (kcal/gmol U) Characteristics Surface area (m2/g) Average particle size (mm) Bulk density (g/cm3) Appearance Sinterability (%)
Direct thermal denitration
Modified direct denitration
110
26
0.1-1.6 43 1.6-2.5 Hard glass slag-like chunks 70
8-12 2.5 mm 0.7-0.9 Fine granular solids >90
Much less heat is required to decompose and denitrate the ammonium-uranium double salt in the MDD process, and the MDD oxide product has greater surface area, better sinterability, and smaller average particle size. In appearance, the MDD oxide product is in the form of fine granular agglomerates; the oxide produced by direct thermal denitration is in the form of hard glass-like chunks. These characteristics show that significant improvements in product quality are achieved in the MDD process. In Table 12.3, key physical properties of MDD-produced UO2 are compared with those of industrial grades of UO2 produced by the ADU process. The physical properties of MOX produced by the MH process and the MDD process and the U.S. ASTM (American Society for Testing and Materials) standards for UO2 and PuO2 are also shown for comparison. The ADU process is already industrially deployed and produces UO2 suitable for fuel fabrication (Haas, 1988). The properties shown for MH- and MDD-produced MOX also indicate suitability for fuel fabrication.
Comparison of UO2 and coconverted MOX physical properties
Table 12.3
Density (g/mL) Surface area (m2/g)
Mean particle size (mm)
Source
Bulk
Tap
ADU UO2 (Source #1) ADU UO2 (Source #2) MDD UO2 MH (UO2-PuO2) MDD (92U-7.4Pu0.4Np)O2 Prior to reduction of UO3 UO2 (ASTM C 757-79) PuO2 (ASTM C 753-81)
2.36 2.3 1.0 2.2 1
3.21 2.65 1.3 3.7 1.4
1
1.3
10.6
5.51
– –
– –
>2.5 >2.5
<10 10
4.81 3.11 6.90 4.5 7.5
5.52 2.40 3.43 Milled 5.36
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Further tests of oxide powder rheology have been made to compare the producthandling characteristics of the MDD-produced UO2 with the industrial-grade UO2 products produced by the ADU process. The results indicated that: the MDD powder requires less energy to transport because it has less particle-to-particle friction during flow; the MDD powder agglomerates have a larger amount of interparticle void space and are more compressible; and the MDD powder is finer, more densely packed, and less permeable.
l
l
l
Thus, based on these analyses, the product-handling characteristics appear to be improved by the MDD process. Mailen et al. (1982) conducted tests with MDD-produced mixed uraniumplutonium oxide having a plutonium content of 22.4 wt%, which is typical of MOX fuel recycled to fast breeder reactors. Part of the MOX product from these tests was shipped to Pacific Northwest National Laboratory for determination of fuel fabrication characteristics. The results are shown in Table 12.4. Sintered densities of >93% of theoretical were obtained. Microstructure analyses made at Oak Ridge National Laboratory (ORNL) showed that the sintered pellets were free of significant porosity. Additional sintered pellets were used in dissolution tests at ORNL; the results showed that the pellets were completely dissolved in 15 h by refluxing in nitric acid (initially 7 M). Periodic samples were taken and showed that the dissolution rate was linear over the 15-h period, indicating the homogeneity of the MOX pellets.
Fabrication properties for MDD-produced MOX powder tests performed at Pacific Northwest National Laboratory
Table 12.4
MOX powder Tap density (g/cm3) Surface area (m2/g)
1.23 9.2
Calcined powder Calcining conditions Temperature (°C) Time (h) Weight loss (%) Tap density (g/cm3) Surface area (m2/g)
600 4 13.6 1.22 10.2
Pellet propertiesa Forming pressure (MPa) Green density (g/cm3) Sintered density (% of theoretical)b a
269 5.39 93.9
296 5.50 93.5
After being calcined, the powder was preslugged to 4.7 g/cm3 and was granulated through a 14-mesh screen. 35.4 wt% of the granules were 14 + 20 mesh. The remaining particles were less than 20 mesh in size. b Sintered at 1700 °C in 4% H2-Ar for 4 h.
322
12.7
Reprocessing and Recycling of Spent Nuclear Fuel
Co-conversion process comparisons
In comparison with precipitation and gelation processes, the MH and MDD advanced direct denitration processes require less-complicated process steps and equipment and do not require precise feed solution concentrations. For example, the one-step denitration replaces the multiple precipitation, filtration, and drying steps. In addition, less liquid waste and recycled product are generated. However, the advanced direct denitration processes do not remove any impurities, whereas coprecipitation has some impurity removal capability. Production of UO2-PuO2 MOX by means of oxalate coprecipitation requires a prereduction (of U6+ to U4+), whereas the MH and MDD processes require a postcalcination reduction of UO3 to UO2. The attributes of the MH and MDD processes are compared in the following list: l
l
l
l
l
l
The MH process has greater technical maturity. The MH process does not require addition of the ammonium nitrate denitration modifier and corresponding safety evaluation. The MH process is an automated multiple-batch process, whereas MDD is operated continuously in a rotary calciner. Both processes use simple one-step denitration but produce UO3, which requires an additional calcination-reduction process. The MDD oxide powder does not require crushing and milling as does the MH product. The MDD product requires only screening through a 20-40 mesh sieve to deagglomerate larger agglomerates. The MH product is 50% plutonium and requires a ball-milling step with UO2 to be down-blended to recycle fuel concentrations, whereas the MDD product may be able to use liquid blending of uranium and plutonium nitrate solutions to meet the precise recycle fuel concentrations.
12.8
Future trends
The MH process has been fully demonstrated at the Japanese Tokai Reprocessing Facility and is being deployed at full industrial scale in the Rokkasho Reprocessing Plant. The MH process product has been tested for production of both light water reactor recycle fuel and fast breeder recycle fuel (Numao et al., 2007). The MDD process has been demonstrated at industrial scale for UO2 production, but MOX production has only been tested at the engineering scale in research and development studies. As indicated in Table 12.3, the MDD process has been used to accomplish co-conversion of uranium-plutonium-neptunium nitrate. Moreover, in companion tests (Walker et al., 2007; Felker et al., 2008), co-conversion of uranium-plutoniumneptunium-americium nitrate was successfully demonstrated. Variations of the MDD process using different modifiers are used industrially for other applications. For example, at the French LaHague reprocessing plant, a similar process and similar equipment are used for codenitration and c-conversion of fissionproduct nitrate solutions to a solid oxide form prior to incorporation into vitrified highlevel waste (Brueziere et al., 2013). Because of its promising simplicity, the MDD process will likely be the focus of continued RD&D in the United States. Variations of the process are of interest in other countries.
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References Bathke, C., Ebbinghaus, B., Sleaford, B., Wallace, R., Collins, B., Hase, K., Robel, M., Jarvinen, G., Bradley, K., Ireland, J., Johnson, M., Prichard, A., Smith, B., 2009. An assessment of the attractiveness of material associated with a MOX fuel cycle from a safeguards perspective. In: Proceedings of 50th Annual Meeting of INMM. Brueziere, J., Chauvin, E., Piroux, J., 2013. World first in high level waste vitrification—a review of French vitrification industrial achievements. In: Proceedings of Global 2013 International Conference. Felker, L., Vedder, R., Walker, E., Collins, E., 2008. Product conversion: the link between separations and fuel fabrication. In: Proceedings of Atalante 2008 International Conference. Haas, P., 1988. A comparison of processes for the conversion of uranyl nitrate into ceramicgrade UO2. Nucl. Technol. 81, 383–406. Haas, P., Arthur, R., Stines, W., 1981. Development of thermal denitration to prepare uranium oxides and mixed oxides for nuclear fuel fabrication. ORNL-5735, Oak Ridge National Laboratory, Oak Ridge, TN. Koizumi, M., Katsuyuki, O., Isagawa, H., Akiyama, H., Todokoro, A., 1983. Development of a process for the co-conversion of Pu-U nitrate mixed solutions to mixed-oxide powder using a microwave heating method. Nucl. Technol. 61, 55–70. Lerch, R., Norman, R., 1984. Nuclear fuel conversion and fabrication chemistry. Radiochim. Acta 36, 75–88. Mailen, J., Pruett, D., McTaggart, D., 1982. Direct thermal denitration to prepare mixed oxides for nuclear fuel fabrication. ORNL/TM-8197, Oak Ridge National Laboratory, Oak Ridge, TN. Notz, K., Haas, P., 1981. Properties and thermal decomposition of the double salts of uranyl nitrate-ammonium nitrate. ORNL/TM-7820, Oak Ridge National Laboratory, Oak Ridge, TN. Numao, T., Nakayashiki, H., Arai, N., Miura, S., Nakamura, H., Tanaka, I., 2007. Results of active test of uranium-plutonium co-denitration Facility at Rokkasho Reprocessing Plant. In: Proceedings of Global 2007 International Conference. Walker, E., Vedder, R., Felker, L., Marschman, S., 2007. Application of modified direct denitration to support the ORNL coupled-end-to-end demonstration in production of mixed oxides suitable for pellet fabrication. In: Proceedings of Global 2007 International Conference.
12
Emory D. Collins Oak Ridge National Laboratory, Oak Ridge, TN, USA
Acronyms ADU ASTM MDD MH MOX NOx ORNL RD&D UNH
ammonium diuranate (process) American Society for Testing and Materials modified direct denitration (process) microwave heating mixed oxide nitrogen oxide Oak Ridge National Laboratory research, development, and demonstration uranyl nitrate hexahydrate
12.1
Introduction
The aqueous nitrate solutions of uranium and plutonium resulting from the reprocessing of spent nuclear fuels must be denitrated and converted to oxide particles with good ceramic properties for fabrication of recycle fuels. Several processes are suitable for this purpose, including those that utilize precipitation: filtration-calcination, such as the ammonium diuranate (ADU) process; the plutonium oxalate precipitation process using plutonium in either the trivalent or the tetravalent state; the uranium or plutonium peroxide precipitation process; or the ammonium-uranium carbonate process. Also, the use of direct thermal denitration processes is possible and is the subject of this chapter.
12.2
History and concepts for improvement
Thermal denitration has been used for many years in the processing and recovery of uranium from mining ores in the United States. Simple thermal denitration was used in large-scale operations to convert aqueous uranyl nitrate hexahydrate (UNH) solution to a solid oxide (uranium trioxide: UO3) interim product, usually for subsequent conversion to uranium fluorides UF4 or UF6 (Haas, 1988). The conversion of UNH to UO3 is a two-step process in which the aqueous solution is first evaporated and Reprocessing and Recycling of Spent Nuclear Fuel. http://dx.doi.org/10.1016/B978-1-78242-212-9.00012-5 © 2015 Elsevier Ltd. All rights reserved.
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concentrated to eliminate most of the water and produce a syrupy liquid, followed by calcination and decomposition (denitration) to produce a solid caked oxide product. In the calcination step, decomposition of the UNH begins at temperatures less than 300 °C and is completed by 400-500 °C. These operations have been performed in heated pots, fluidized beds, and stirred troughs, either in batch or in continuous mode. In European refineries, the denitration and conversion is accomplished by means of the addition of ammonia to the UNH solution to precipitate ADU, which is then filtered, dried, and calcined to form uranium trioxide powder. The use of direct thermal denitration eliminates multiple precipitation-filtrationcalcination steps and enables the use of less-complicated process equipment (Lerch and Norman, 1984). Direct thermal denitration, however, does not provide any separation of impurity metals and relies on previous chemical separations to provide the required product purity. Although direct thermal denitration offers significant process simplification, the process has not been used widely due to the formation of a “mastic” phase during melting and decomposition of the metal salts, leading to poor handling characteristics and poor ceramic properties of the product oxide. Since the 1980s, research, development, and demonstration (RD&D) studies in Japan and the United States have focused on direct thermal denitration processes to prepare uranium oxide and mixed oxides (MOXs) for recycle nuclear fuel fabrication (Koizumi et al., 1983; Haas et al., 1981). The RD&D studies have been aimed at the use of more uniform heating equipment or the addition of modifier agents to avoid the mastic phase formation and to enable suitable product quality. A major breakthrough in this area occurred when Haas et al. (1981) discovered that, by introducing ammonium nitrate into a uranium solution, the mastic phase is eliminated by the formation and denitration of double salts of ammonium-uranium nitrate. The process is called modified direct denitration (MDD). Current reprocessing and recycle fuel fabrication operations rely on the production of separated uranium oxide and plutonium oxide and the use of ball-milling methods to produce a suitable MOX for use in recycle fuels. Separated plutonium is viewed as a safeguards concern that increases the “attractiveness level” (Bathke et al., 2009) for diversion of the fissile material to weapons of mass destruction. Therefore, coprocessing and co-conversion methods are being developed in several countries for use in future reprocessing and recycle fuel fabrication plants. A microwave heating (MH) process has been developed and demonstrated in Japan for codenitration of uranium-plutonium solutions without formation of the mastic phase to produce a homogeneous MOX. Both the Japanese MH process and the U.S. MDD process are suitable for codenitration. Description of these processes is the focus of this chapter.
12.3
Process chemistry
Thermal denitration of uranium and plutonium, either individually or from mixed uranium-plutonium, is believed to be represented by the following approximate calcination reactions: UO2 ðNO3 Þ2 6H2 O ! UO3 + 1:86NO2 + 0:14NO + 0:57O2 + 6H2 O
(12.1)
Advanced thermal denitration conversion processes
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PuðNO3 Þ4 5H2 O ! PuO2 + 3:4NO2 + 0:6NO + 1:3O2 + 5H2 O
(12.2)
The denitration begins at a temperature <300 °C and is completed at 350-400 °C. The reaction can be performed in a batch mode, as in the Japanese MH process, or in a continuous mode, as in the U.S. MDD process. The Japanese MH process proceeds through a sequence of four steps as the temperature is increased (Numao et al., 2007): 1. The mixed uranium-plutonium nitrate solution is concentrated at 120 °C, the azeotropic temperature of nitric acid and water. 2. Decomposition (denitration) of the fused salt occurs between 120 and 250 °C, during which nitrogen oxide (NOx) gases are generated, recovered by scrubbing of the off-gas, and subsequently recycled. 3. Above 250 °C, residual moisture evaporates and NOx evolves from the decomposition of uranium. 4. The denitration is completed at about 350 °C, and the temperature begins to drop as the UO3 begins to decompose to U3O8. The temperature at that point in the process is detected by means of an infrared thermometer, and the MH is stopped to prevent excessive decomposition of the UO3 product.
The reaction sequence occurs in a time frame of 30-40 min per batch. Slower heating was found to result in localized heating and temperature increase, creating a product powder that is less active and less suitable for MOX fuel fabrication (Numao et al., 2007). The process chemistry of the MDD process was studied by Notz and Haas (1981) for the denitration of the ammonium-uranium nitrate double salt, which proceeds through a sequence of reactions (Equations 12.3–12.5) as illustrated by the results of a thermogravimetric analysis shown in Figure 12.1.
171
(NH4)2UO2(NO3)4 . 2H2O
160
(NH4)2UO2(NO3)4 (160.6 mg)
60 NH4UO2(NO3)3 (137.4 mg)
137
40
Actual weight (mg)
Relative weight (mg)
80
20 UO3 (82.9 mg)
0
0
50
100
200 300 Temperature (⬚C)
400
84
Figure 12.1 Thermogravimetric analysis of (NH4)2UO2(NO3)42H2O (initial sample weight, 171 mg).
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The thermal decomposition proceeds in three distinct steps: 1. Dehydration at 40 °C: ðNH4 Þ2 UO2 ðNO3 Þ4 2H2 O ! ðNH4 Þ2 UO2 ðNO3 Þ4 + 2H2 O
(12.3)
2. Loss of 1 mole of ammonium nitrate (NH4NO3) at 170 °C ðNH4 Þ2 UO2 ðNO3 Þ4 ! NH4 UO2 ðNO3 Þ3 + N2 O + 2H2 O
(12.4)
3. Conversion to UO3 at 270 °C: NH4 UO2 ðNO3 Þ3 ! UO3 + ðN2 O5 Þ + N2 O + 2H2 O
(12.5)
The decomposition products for the latter two reactions were inferred on the basis that NH4NO3 decomposes under controlled conditions to H2O and N2O, which is a standard method for the preparation of N2O. However, a differential thermal analysis, shown in Figure 12.2, and effluent gas analysis, shown in Table 12.1, indicated that a combination of several other reactions, which are known to occur for decomposition of NH4NO3, are involved. Some are exothermic, but the overall effect is endothermic with a large endotherm occurring immediately after the exotherm, as indicated in Figure 12.2.
12.4
Process equipment and operation
The Japanese MH process equipment is centered on multiple-batch MH units, each contained in a glove box. Each batch of uranium-plutonium solution is loaded into a 50 cm diameter, 6 cm deep round denitration dish, which is constructed of Si3N4; a ceramic material that transmits the microwave energy uniformly from the surface to the bottom Figure 12.2 Differential thermal analyses. Upward deflection ¼ exothermic effect; downward deflection ¼ endothermic effect.
Reaction (3)
Reactions (2) and (3)
Reactions (1) and (2) and (3)
100 200 Temperature (°C)
300
Advanced thermal denitration conversion processes
Table 12.1 Mass no.
317
Effluent gas analysis Species
Relative quantity
Temperature of maximum rate (°C) 320
185
280
185
290
Decomposition products
14 15 16 17 18 28 30 32 44 46
N+ NH+ NH+2 NH+3 H2O+ N+2 NO+ O+2 N2O+ NO+2
UO2 (NO3)2 m m m m M m
NH4NO3 t M M M M
m M
NH4UO2 (NO3)3 M m m m M m M M M M
(NH4)2UO2
(NO3)4
m M M M m
M M m
m
M m M M M M
Abbreviations: m, minor component; t, trace component; M, major component.
of the solution. In addition, the denitration dish is slowly rotated on a turntable inside the glove box to heat the solution uniformly. Active tests have been made of full-size production units at the Rokkasho Reprocessing Plant (Numao et al., 2007). Uranium nitrate solution and plutonium nitrate solution are received from the chemical purification facility and are blended at a mass ratio of 1:1 U:Pu for codenitration and conversion to UO3-PuO2. The denitrated powder has a porous structure due to foam formation by evolution of NOx gas during boiling. Bulk density of the product is typically about 2 g/cm3. Accessory equipment includes a microwave generation and wave guide transmission to each glove box, plus process off-gas treatment for recovery and recycle of nitric acid. The capacity of the MH system at Rokkasho is 108 kg (U + Pu) per day (Numao et al., 2007). The MDD process equipment is centered on one or more rotary kilns, such as the development unit illustrated in Figure 12.3. Each kiln is basically a straight length of stainless steel tubing with a diameter selected for the capacity of oxide production needed. Typically, the kiln is mounted on steel wheels and rotates within a tube furnace, which surrounds about one-third or more of the pipe length. Feed solution is introduced at one end of the rotary pipe kiln, which is angled downward to enable the decomposing salt and solid oxide powder to move by a combination of gravity and tube rotation to the end of the kiln, where the oxide product is collected. One or more solid stainless steel rods are located within the kiln and roll freely to act as “breaker bars” to crush lumps or to remove solid oxide from the tube wall. An air purge gas enters the kiln at the powder discharge end and
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TE indicates thermocouple
220 V TC indicates temperature controller
All internal materials: stainless steel 2500-W tube furnace
TC TE
Scraper 3.5-in. IPS Sch 10 pipe Support wheels Solution in gases out TE
1.91-cm. (3/4-in.) O.D. tubing
TE
Variable-speed motor with chain and sprocket drive
Breaker bars
Solid product collection
3⬚
Figure 12.3 Rotary kiln cross section.
flows countercurrently through the kiln to remove the water, acid, and NOx gases that are formed. The off-gases leave the kiln at the liquid feed end and are routed to a scrubber and condenser system for recovery and recycle of nitric acid and for removal of entrained oxide. Most of the development work has been done with uranium in rotary kilns measuring up to 16 cm in diameter and has progressed to demonstration at the multi-kilogram-per-hour scale. More recently, small-scale units (Figure 12.4) have been built and installed in a glove box to permit testing of the process for denitration of ammonium-uranium-plutonium nitrate solutions and co-conversion to MOX powder at a rate of 100 g/h.
Off-gas
Condenser Feed tank Scrubber
Air purge Pump
Condensate collection Rotary kiln furnace Oxide product
Figure 12.4 The U.S. MDD process and small-scale experimental equipment.
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319
Coprocessed uranium-plutonium nitrate solution is mixed with ammonia gas or ammonium nitrate solution to provide the desired ratio of NH+4 to heavy metal, typically 2:1. The resulting ammonium-heavy metal double salt solution is fed to the rotary kiln at the rate needed for the production capacity. Inside the kiln, the solution goes through the sequence of (i) evaporation of water and nitric acid, (ii) denitration of the fused double salt, and (iii) formation of mixed uranium-plutonium powder. With the tube furnace operating at a typical temperature of 650 °C, measurements of the temperature profile and decomposition gas concentrations within the kiln have indicated that the denitration is completed within a short distance from the liquid feed entry point.
12.5
Conversion of UO3 to UO2
Thermal denitration of uranium produces a UO3 product; codenitration of mixed uranium and plutonium produces a mixed UO3-PuO2 product. Subsequently, a separate calcination and reduction step is required to convert the UO3 to UO2 or the UO3-PuO2 to a single-phase (U,Pu)O2 solid solution. The UO3 can be converted to UO2 by means of the following calcination and reduction reactions: 3UO3 ! U3 O8 + 0:5O2
(12.6)
U3 O8 + 2H2 ! 3UO2 + 2H2 O
(12.7)
The calcination occurs at temperatures up to 700 °C. The reduction reaction occurs at a temperature of 700-800 °C in a period of 4 h. Alternatively, by heating in a reducing atmosphere, uranium trioxide (UO3) can be directly converted to uranium dioxide (UO2). For the Japanese MH process at the Rokkasho plant, the dry MOX product from the denitration dish is crushed and placed into a rotary calcination furnace. Then, the calcined powder is transferred to a rotary reduction furnace, where it is treated with 95% N2-5% H2 to convert the powder to the final MOX form (U,Pu)O2 (Numao et al., 2007). The oxide powder produced in the MDD rotary kiln leaves the kiln in an agglomerated form, which requires screening through a 40-mesh sieve before calcination and reduction. Although not extensively developed, the calcination-reduction process and equipment would likely be similar to the process and equipment used for the MH process at Rokkasho. X-ray diffraction examination of MDD MOX has determined that the oxides in the MOX form a solid solution during reduction (Walker et al., 2007).
12.6
Product characteristics
Table 12.2 shows the endothermic heat of decomposition and typical UO2 product characteristics from direct thermal denitration in comparison with those from the MDD process.
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Comparison of typical oxide product characteristics produced by MDD and direct denitration
Table 12.2
Heat of decomposition, DH (kcal/gmol U) Characteristics Surface area (m2/g) Average particle size (mm) Bulk density (g/cm3) Appearance Sinterability (%)
Direct thermal denitration
Modified direct denitration
110
26
0.1-1.6 43 1.6-2.5 Hard glass slag-like chunks 70
8-12 2.5 mm 0.7-0.9 Fine granular solids >90
Much less heat is required to decompose and denitrate the ammonium-uranium double salt in the MDD process, and the MDD oxide product has greater surface area, better sinterability, and smaller average particle size. In appearance, the MDD oxide product is in the form of fine granular agglomerates; the oxide produced by direct thermal denitration is in the form of hard glass-like chunks. These characteristics show that significant improvements in product quality are achieved in the MDD process. In Table 12.3, key physical properties of MDD-produced UO2 are compared with those of industrial grades of UO2 produced by the ADU process. The physical properties of MOX produced by the MH process and the MDD process and the U.S. ASTM (American Society for Testing and Materials) standards for UO2 and PuO2 are also shown for comparison. The ADU process is already industrially deployed and produces UO2 suitable for fuel fabrication (Haas, 1988). The properties shown for MH- and MDD-produced MOX also indicate suitability for fuel fabrication.
Comparison of UO2 and coconverted MOX physical properties
Table 12.3
Density (g/mL) Surface area (m2/g)
Mean particle size (mm)
Source
Bulk
Tap
ADU UO2 (Source #1) ADU UO2 (Source #2) MDD UO2 MH (UO2-PuO2) MDD (92U-7.4Pu0.4Np)O2 Prior to reduction of UO3 UO2 (ASTM C 757-79) PuO2 (ASTM C 753-81)
2.36 2.3 1.0 2.2 1
3.21 2.65 1.3 3.7 1.4
1
1.3
10.6
5.51
– –
– –
>2.5 >2.5
<10 10
4.81 3.11 6.90 4.5 7.5
5.52 2.40 3.43 Milled 5.36
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Further tests of oxide powder rheology have been made to compare the producthandling characteristics of the MDD-produced UO2 with the industrial-grade UO2 products produced by the ADU process. The results indicated that: the MDD powder requires less energy to transport because it has less particle-to-particle friction during flow; the MDD powder agglomerates have a larger amount of interparticle void space and are more compressible; and the MDD powder is finer, more densely packed, and less permeable.
l
l
l
Thus, based on these analyses, the product-handling characteristics appear to be improved by the MDD process. Mailen et al. (1982) conducted tests with MDD-produced mixed uraniumplutonium oxide having a plutonium content of 22.4 wt%, which is typical of MOX fuel recycled to fast breeder reactors. Part of the MOX product from these tests was shipped to Pacific Northwest National Laboratory for determination of fuel fabrication characteristics. The results are shown in Table 12.4. Sintered densities of >93% of theoretical were obtained. Microstructure analyses made at Oak Ridge National Laboratory (ORNL) showed that the sintered pellets were free of significant porosity. Additional sintered pellets were used in dissolution tests at ORNL; the results showed that the pellets were completely dissolved in 15 h by refluxing in nitric acid (initially 7 M). Periodic samples were taken and showed that the dissolution rate was linear over the 15-h period, indicating the homogeneity of the MOX pellets.
Fabrication properties for MDD-produced MOX powder tests performed at Pacific Northwest National Laboratory
Table 12.4
MOX powder Tap density (g/cm3) Surface area (m2/g)
1.23 9.2
Calcined powder Calcining conditions Temperature (°C) Time (h) Weight loss (%) Tap density (g/cm3) Surface area (m2/g)
600 4 13.6 1.22 10.2
Pellet propertiesa Forming pressure (MPa) Green density (g/cm3) Sintered density (% of theoretical)b a
269 5.39 93.9
296 5.50 93.5
After being calcined, the powder was preslugged to 4.7 g/cm3 and was granulated through a 14-mesh screen. 35.4 wt% of the granules were 14 + 20 mesh. The remaining particles were less than 20 mesh in size. b Sintered at 1700 °C in 4% H2-Ar for 4 h.
322
12.7
Reprocessing and Recycling of Spent Nuclear Fuel
Co-conversion process comparisons
In comparison with precipitation and gelation processes, the MH and MDD advanced direct denitration processes require less-complicated process steps and equipment and do not require precise feed solution concentrations. For example, the one-step denitration replaces the multiple precipitation, filtration, and drying steps. In addition, less liquid waste and recycled product are generated. However, the advanced direct denitration processes do not remove any impurities, whereas coprecipitation has some impurity removal capability. Production of UO2-PuO2 MOX by means of oxalate coprecipitation requires a prereduction (of U6+ to U4+), whereas the MH and MDD processes require a postcalcination reduction of UO3 to UO2. The attributes of the MH and MDD processes are compared in the following list: l
l
l
l
l
l
The MH process has greater technical maturity. The MH process does not require addition of the ammonium nitrate denitration modifier and corresponding safety evaluation. The MH process is an automated multiple-batch process, whereas MDD is operated continuously in a rotary calciner. Both processes use simple one-step denitration but produce UO3, which requires an additional calcination-reduction process. The MDD oxide powder does not require crushing and milling as does the MH product. The MDD product requires only screening through a 20-40 mesh sieve to deagglomerate larger agglomerates. The MH product is 50% plutonium and requires a ball-milling step with UO2 to be down-blended to recycle fuel concentrations, whereas the MDD product may be able to use liquid blending of uranium and plutonium nitrate solutions to meet the precise recycle fuel concentrations.
12.8
Future trends
The MH process has been fully demonstrated at the Japanese Tokai Reprocessing Facility and is being deployed at full industrial scale in the Rokkasho Reprocessing Plant. The MH process product has been tested for production of both light water reactor recycle fuel and fast breeder recycle fuel (Numao et al., 2007). The MDD process has been demonstrated at industrial scale for UO2 production, but MOX production has only been tested at the engineering scale in research and development studies. As indicated in Table 12.3, the MDD process has been used to accomplish co-conversion of uranium-plutonium-neptunium nitrate. Moreover, in companion tests (Walker et al., 2007; Felker et al., 2008), co-conversion of uranium-plutoniumneptunium-americium nitrate was successfully demonstrated. Variations of the MDD process using different modifiers are used industrially for other applications. For example, at the French LaHague reprocessing plant, a similar process and similar equipment are used for codenitration and c-conversion of fissionproduct nitrate solutions to a solid oxide form prior to incorporation into vitrified highlevel waste (Brueziere et al., 2013). Because of its promising simplicity, the MDD process will likely be the focus of continued RD&D in the United States. Variations of the process are of interest in other countries.
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References Bathke, C., Ebbinghaus, B., Sleaford, B., Wallace, R., Collins, B., Hase, K., Robel, M., Jarvinen, G., Bradley, K., Ireland, J., Johnson, M., Prichard, A., Smith, B., 2009. An assessment of the attractiveness of material associated with a MOX fuel cycle from a safeguards perspective. In: Proceedings of 50th Annual Meeting of INMM. Brueziere, J., Chauvin, E., Piroux, J., 2013. World first in high level waste vitrification—a review of French vitrification industrial achievements. In: Proceedings of Global 2013 International Conference. Felker, L., Vedder, R., Walker, E., Collins, E., 2008. Product conversion: the link between separations and fuel fabrication. In: Proceedings of Atalante 2008 International Conference. Haas, P., 1988. A comparison of processes for the conversion of uranyl nitrate into ceramicgrade UO2. Nucl. Technol. 81, 383–406. Haas, P., Arthur, R., Stines, W., 1981. Development of thermal denitration to prepare uranium oxides and mixed oxides for nuclear fuel fabrication. ORNL-5735, Oak Ridge National Laboratory, Oak Ridge, TN. Koizumi, M., Katsuyuki, O., Isagawa, H., Akiyama, H., Todokoro, A., 1983. Development of a process for the co-conversion of Pu-U nitrate mixed solutions to mixed-oxide powder using a microwave heating method. Nucl. Technol. 61, 55–70. Lerch, R., Norman, R., 1984. Nuclear fuel conversion and fabrication chemistry. Radiochim. Acta 36, 75–88. Mailen, J., Pruett, D., McTaggart, D., 1982. Direct thermal denitration to prepare mixed oxides for nuclear fuel fabrication. ORNL/TM-8197, Oak Ridge National Laboratory, Oak Ridge, TN. Notz, K., Haas, P., 1981. Properties and thermal decomposition of the double salts of uranyl nitrate-ammonium nitrate. ORNL/TM-7820, Oak Ridge National Laboratory, Oak Ridge, TN. Numao, T., Nakayashiki, H., Arai, N., Miura, S., Nakamura, H., Tanaka, I., 2007. Results of active test of uranium-plutonium co-denitration Facility at Rokkasho Reprocessing Plant. In: Proceedings of Global 2007 International Conference. Walker, E., Vedder, R., Felker, L., Marschman, S., 2007. Application of modified direct denitration to support the ORNL coupled-end-to-end demonstration in production of mixed oxides suitable for pellet fabrication. In: Proceedings of Global 2007 International Conference.