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Concentration of radioactive components in liquid low-level radioactive waste by membrane distillation
Journal of Membrane Science 163 (1999) 257–264
Concentration of radioactive components in liquid low-level radioactive waste by membrane distillation G. Zakrzewska-Trznadel ∗ , M. Harasimowicz, A.G. Chmielewski Institute of Nuclear Chemistry and Technology, Dorodna 16, 03-195 Warsaw, Poland Received 9 February 1999; received in revised form 3 May 1999; accepted 4 May 1999
Abstract The paper addresses some aspects of liquid low-level radioactive waste (LLLW) purification. Since the volume of liquid low-level wastes is usually large and the activity is rather low, the radioactive substances separated from the non-active portion have to be concentrated into the small volume for subsequent conditioning and disposal. The need for the improvement of decontamination and minimisation of the costs have led to new specific methods being under examination and development. The method proposed in the paper is membrane distillation. The experimental work described below supports the statement that membrane distillation can be an attractive alternative for liquid radioactive waste treatment. The advantages of membrane distillation over the other processes commonly used for the processing of LLLW are discussed in the paper. ©1999 Elsevier Science B.V. All rights reserved. Keywords: Membrane; Membrane distillation; Radioactive waste treatment; Decontamination
1. Introduction Radioactive wastes are generated in a number of different kinds of facilities. They arise in a wide range of concentrations of radioactive materials and a variety of physical and chemical forms. Low-level waste is defined as a waste which, because of its low radionuclide content, does not require shielding during normal handling and transportation. In Poland, one assumes a specific activity of liquid low-level waste (LLLW) < 107 Bq/m3 . There is a variety of methods for the treatment and conditioning of wastes prior to disposal. The selec∗ Corresponding author. Fax: +4822-811-15-32 E-mail address: [email protected] (G. ZakrzewskaTrznadel)
tion of a treatment process for an aqueous waste is determined by the physical, chemical and radiological properties of the waste. The processing of liquid waste aims at separating the radionuclides from the liquid phase and concentrating them in a solid waste form. The separation is pursued until the residual concentration or total amount of radionuclides in liquid phase is below the limits set by the regulatory body for the discharge of liquid waste from a nuclear facility as an effluent. The treatment technologies employed in nuclear industry include chemical precipitation, ion exchange, evaporation, membrane processes, solvent extraction and biotechnology. A number of factors have to be taken into account before the selection of a treatment process, including capital and operational costs, costs associated with the disposal of secondary wastes, etc.
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Fig. 1. MD pilot plant for radioactive waste concentration. 1: spiral-wound module; 2: distillate reservoir; 3: retentate tank; 4, 5: pumps; 6: heat supplier; 7, 8: heat exchangers; 9, 10: pre-treatment filters.
Membrane processes have already been employed successfully in nuclear industry. Reverse osmosis has been applied for a large scale operation in Chalk River Laboratory [1,2]. The plant is composed of a microfiltration stage, 50 spiral wound RO modules in a three-stage cascade, and tubular modules. Ultrafiltration is used for the separation of dissolved salts or acids from particulate and colloidal material (Sallefield, Paks, Harwell) [3]. The process can also be combined with physical and chemical processes of binding small ions with macromolecular ligands or complexation, known as ‘seeded ultrafiltration’ [4]. Microfiltration is often employed as a pre-treatment stage for RO. The pressure-driven membrane processes are very effective, especially when they are realised as a multistage process combining MF or UF and RO units, but they also have many limitations as fouling phenomena often resulting in cleaning operations which decrease membrane resistance and produce secondary wastes, or a need for the design of rather complex plants operating at high pressure. Some of these disadvantages can be avoided by membrane distillation (MD), non-isothermal process employing hydrophobic porous membrane, developed for water solutions concentration or for the production of pure water [5–11]. As the process is characterised by high retention of non-volatile solutes, large decontamination factors are expected in the separation of radionuclides which are present in liquid radioactive effluents in mainly ionic form. Laboratory tests have
shown that membrane distillation can be used for a concentration of radioactive solutions [12]. The results of laboratory experiments have led to MD pilot plant design and construction.
2. Experimental 2.1. Apparatus A scheme of the pilot plant is shown in Fig. 1. The productivity of the plant is up to ∼0.05 m3 /h of distillate which is a water of radiochemical purity (<10 Bq/dm3 for  and ␥ emitters). The installation consists of spiral-wound PTFE module G-4.0-6-7 (SEP GmbH) 1, distillate reservoir 2 (V = 80 dm3 ), retentate tank (V = 80 dm3 ) 3, two PRI 505M Ebara pumps 4 and 5, Omega heat supplier 6, two heat exchangers 7, 8 (JAD X-6/50 and JAD 6/50) and ceramic filters for pre-treatment (9,10). The process is conducted as a direct-contact membrane distillation. In such an arrangement, two streams at different temperatures directly contact the porous, hydrophobic membrane. The permeate condenses in a cold stream (distillate) which is pumped from reservoir 2 by PRI 505M Ebara pump through heat exchangers, pre-treatment units with ceramic filters to the PTFE module.The warm stream (retentate) is circulated from tank 3 by pump through the heat exchanger, heater and pre-treatment filters. Both streams — cold and warm — are returned to the reservoirs. The
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with an NaI/Tl scintillation crystal or an 800-channel analyser DIDAC with the spectrometric probe SCINTIBLOC (Intertechnique, France). For more precise concentration measurements at the beginning and at the end of the experiment, ionic chromatograph (DIONEX) and DSPEC Digital Gamma-Ray Spectrometer with germanium detector were used. 2.3. Experiments Fig. 2. Spiral-wound membrane module — cross-section of the spiral. 1: distillate channel; 2: retentate channel; 3: distillate inlet; 4: distillate outlet; 5: feedwater inlet; 6: feed water outlet; 7: membrane.
Table 1 Characteristic of spiral-wound MD module Overall dimensions φ × h Effective surface area Volume flow of feed streams Distillate flux Recommended feed stream temperature Average energy consumption per 1 m3 of the distillate PTFE membrane Pore size Porosity
0.45 m × 0.62 m 4 m2 300–1500 dm3 /h 10–60 dm3 /h 60–80◦ C ∼600 kW hth 0.2 m 80%
flows in the module are operated on the counter-flow principle. Fig. 2 shows the construction of the spiral membrane in the module and the flow arrangement. The flow rates in the module are set in such a way that sufficient heat-transfer coefficients and required distillate transport through the membrane are obtained. The process parameters are adjusted to ensure that the outlet temperature of warm stream is lower than that of the cold one and the recovery of the part of lost heat is possible in heat exchanger 7. Additionally, the distillate is cooled down in heat exchanger 8. The characteristics of the module as per the manufacturer is shown in Table 1.
2.3.1. The performance characteristic of PTFE module The experiments were conducted in the temperature range 35–80◦ C at the feed inlet and 5–30◦ C at the cold stream (distillate) inlet, with the feed and distillate flow rates set in the range 300–1500 dm3 /h. Under these conditions, the permeate stream 10–50 dm3 /h (60–300 dm3 /m2 per day) was obtained. The permeate flow rate was measured periodically with a graduated cylinder collecting the excess of the liquid flowing out of the distillate reservoir. The influence of process parameters (inlet temperatures, flow rates of both streams, cooling water flow rate, etc.) on the productivity of the installation was checked. A number of experimental cycles were run in which one of these parameters was changed. 2.3.2. Raising the concentration of non-active solutions of inorganic salts The preliminary tests were performed using solutions of pure inorganic salts: NaCl, NaNO3 and KCl. The initial concentration of salts in the feed solution was changed from 1 to 55 g/dm3 . The experiments were conducted at different temperatures in such a way that the cold stream rise returned to the retentate tank by the over-fall pipe; so the concentration in the warm stream is stable if the process is run properly. The flow rate of cold and warm streams was 960 l/h. Every 0.5–1 h, the samples from warm and cold streams were drawn and the conductivity of both solutions was measured.
2.2. Analysis The amount of inorganic solute in both streams was controlled with conductometer TDScan20 and the concentration of radioactive substances via specific activity measurements using standard probe equipped
2.3.3. Concentration of radioactive solutions The experiments were run in the same way as the tests with non-active solutions. The excess of water from the distillate tank was turned back into the retentate reservoir. As a feed, the model solu-
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Fig. 3. Productivity of MD installation vs. inlet feed temperature. Flow rates VD = VR = 720, 960, 1260 dm3 /h.
tions of caesium and cobalt salts with admixtures of radioactive 137 Cs and 60 Co (total specific activity ∼5000 Bq/dm3 ) were used and the samples of the original liquid low-level radioactive wastes of specific activity up to 104 Bq/dm3 were used as well. Every 1–2 h, the specific activity of samples drawn from the two streams using analyser DIDAC with NaI/Tl probe was measured. Additionally, the specific conductivity and pH of the samples were controlled. At the beginning and at the end of the experiment, the composition of both solutions was determined by ion chromatography and radioactivity was determined by use of DSPEC Digital Gamma-Ray Spectrometer with germanium detector.
3. Results and discussion The highest permeate volumes were obtained at high inlet feed temperature and high distillate and retentate flow rates. In Fig. 3, the dependence of permeate flow Vp on the inlet feed temperature for different cold and warm stream flow rates is shown. The flow rates of both feed streams circulated in the unit were set at the same level: VD = VR (720, 960, 1260 dm3 /h). The cooling water flow rate in heat exchanger 8 influenced the productivity of MD installation as well. The experiments showed that the increase in the flow rate of cooling water caused a slight increase in the permeate flux; but it induced the rise in the energy consumption for additional heating of the retentate too. In some cases, the significant increase in cooling seems to be uneconomic.
Fig. 4. Energy consumption per unit of the product.
Fig. 5. MD experiment with the solution of sodium nitrate.
The best conditions of running the process were at higher inlet feed temperatures (70–80◦ C). In this temperature range, the energy consumption per unit of the product was lower than for lower inlet feed temperatures (Fig. 4). Experiments with sodium chloride, sodium nitrate and potassium chloride showed that the ions of dissolved salts did not pass the porous membrane and that all of them were retained in the retentate. An example of the experimental results is shown in Fig. 5. The upper curve represents the conductivity of the retentate which was stable during the experiment for each initial concentration of the feed. The concentration of the feed solution was changed from 1.3 to 18.5 g/dm3 . For the same time period, specific conductivity of the distillate decreased from 250 S/cm at the beginning of the experiment to 40 S/cm at the end. The increase in feed concentration did not significantly influence the
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Fig. 6. The influence of salt concentration on permeate flux for different inlet feed temperatures.
permeate flow rate and did not change the retention factor. Fig. 6 presents the concentration dependence of permeate flux for different inlet feed temperatures. Up to the concentration 18.5 g/dm3 , no significant change was observed in the permeate flow rate. Preliminary experiments with radioactive compounds were done using the model solutions of 60 Co and 137 Cs. The experimental temperatures were as follows: Temperature of the retentate at the inlet to the module: T1 = 67.4◦ C Temperature of the distillate at the inlet to the module: T2 = 28.8◦ C Temperature of the retentate at the module outlet: T3 = 40.4◦ C Temperature of the distillate at the module outlet: T4 = 51.7◦ C The flow rates of the distillate and retentate were fixed at 960 dm3 /h. Under these conditions, the permeate flow was 27 dm3 /h (flux 6.75 dm3 /m2 h). During the experiment, the activity of the retentate did not change (the concentrated retentate was diluted by the water returned from the distillate reservoir) and the activity of the distillate was stable on the level of the natural background radioactivity. The retention of radioactive ions in the retentate was complete. Similar results were obtained for the samples of the original laboratory radioactive waste. During the experiment conducted under the same conditions as the experiments with model solutions, the specific conductivity, radioactivity and pH of samples drawn from
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two streams were controlled every hour. The results of the measurements are presented in Figs. 7 and 8. The specific conductivity of the retentate (Fig. 7) did not change during the experiment. The conductivity of the distillate decreased from 130 to 45 S/cm. Small sorption of radionuclides took place at the beginning of the test (Fig. 8: retentate curve), and after 12 h, it reached the fixed level. The radioactivity of the distillate did not change during the 20 h experiment and the pH too of both streams remained constant. The composition of the feed solution and the distillate at the end of experiment is shown in Table 2. Three parameters describe the efficiency of purification of radioactive liquid waste: • the volume reduction factor (VRF), which is the ratio of the waste volume before treatment to the volume of the residues containing the bulk of the radioactivity, • the decontamination factor defined as DF =
af vf total activity in feed = total activity in effluent a e ve
(1)
where af and ae are the specific activities of the feed and the effluent, respectively, and f and e are the volumes of the feed and the effluent, respectively, and • the retention coefficient (φ) that refers to the concentrations of the non-active ballast in retentate (cR ) and permeate (cP ): φ = 1−
cP cR
(2)
Decontamination factors and retention coefficients calculated for the data collected in the experiment are presented in Table 2. All retention coefficients were higher than 0.9, most of them close to 1. Decontamination factors for most of the radioisotopes present in the feed stream were very high. Actually, in the effluent, only Co-60 and Cs-137 were detected. For cobalt, DF was 4336.5, while for caesium, it was about 44. The rest of the radioisotopes were not detected in the distillate, so, according to Eq. (1), the decontamination factors for them were close to infinity. The rather poor results for caesium were expected: this radioisotope is hardly removable from the processed waste by most of the known methods. Employing seeded ultrafiltration in which caesium
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Fig. 7. Conductivity of the retentate and the distillate in time.
Table 2 Chemical composition of the waste sample used in the experiments and the effluent after the membrane distillation plant Ion
Concentration in the feed (mg/dm3 )
Concentration in the effluent (mg/dm3 )
Retention coefficient (φ)
Na+
SO42−
1060.6 207.1 21 33.7 87.2 5.7 744.2 1832.9 37.6
3.269 14.584 0.212 not detected 2.375 0.442 1.485 0.065 0.186
0.9969 0.9296 0.9899 1 0.9728 0.9225 0.9980 0.9999 0.9950
Radioisotope
Activity of the feed (Bq/dm3 )
Activity of the effluent (Bq/dm3 )
Decontamination factor (DF)
140 La
<6.53 E − 01 2.99 E + 03 5.26 E + 02 8.62 E + 01 3.73 E + 01 1.04 E + 01 3.39 E + 03 7.84 E + 00 2.95 E + 01 4.51 E + 03
not detected not detected not detected not detected not detected not detected not detected not detected 6.73 E − 01 1.04 E + 00
→∞ →∞ →∞ →∞ →∞ →∞ →∞ →∞ 43.8 4336.5
NH4+ K+ Mg2+ Ca2+ F− Cl− NO 3
133 Ba 170 Tm 114m In 192 Ir 110m Ag 65 Zn 134 Cs 137 Cs 60 Co
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Fig. 8. Radioactivity of the retentate and the distillate in time.
ions were bound with cyanoferrates (Cu2 [Fe(CN)]6 , Ti[Fe(CN)]6 ), decontamination factors equal to 32–45 were reached [13]. The decontamination factor obtained by use of thin-film composite membrane element NTR 739HF (NITTO) was 30 for caesium in model solutions, and 23 in original waste samples, while for cobalt, even DF = 515 were reached in the same plant [14]. The specific activity of the radioisotopes presented in Table 2 was measured using DSPEC Digital Gamma Ray Spectrometer with Germanium detector which has a much more higher geometrical efficiency than the NaI/Tl detector. Due to that, the activity of radioisotopes in Table 2 is higher than the total activity of the waste shown in Fig. 8. One of the more often-risen questions about membrane distillation reliability is the danger of membrane hydrophobicity loss after long-time operation. The membrane made of porous polytetrafluoroethylene, used in the experiments, resists the entry of pure water at 0.28 MPa, which is sufficiently large with respect to the operating pressures of the membrane distillation plant. The pressures at the retentate and distillate entrances to the module were kept on almost the same level, 0.01–0.02 MPa according to the manufacturer’s demand that they should not exceed 0.07 MPa. Usually, the pressure on the distillate side was adjusted as a few percentage units higher than that at the retentate entrance to diminish the risk of wetting. Stabilisation of flows in the unit was done before the membrane module was switched on; so the flow through the module was uniform and stable.
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In long-duration experiments performed for up to 80 h, wetting of the membrane was not observed. In fact, as model feed solutions, inorganic salts dissolved in water, which would increase the surface tension and interfacial tension, were used. Such compounds (sulphates, nitrates, chlorides) are mainly present in liquid radioactive wastes. All substances which decrease the surface tension (oils, surfactants, etc.) are rather dangerous for hydrophobic membrane, and have to be removed before the MD unit if they are present in processed waste. Most of the methods applied for liquid radioactive wastes processing require the removal of organic compounds at the pre-treatment stage. These substances need a special and different treatment by solvent extraction, oxidation or incineration. Small amounts of organic admixtures can be removed by adsorption, oxidation, phase separation by adduct formation, etc. Laboratory experiments in which the solutions of inorganic salts were filtered via MD showed that the achievement of 25% solute concentration in the retentate is possible. This concentration is the upper limit for solutions immobilised in concrete. In fact, usually, the limit of solute concentration is not reached because the dose limit is exceeded first. The problems of blockage and scaling are coupled with the application of almost all membrane processes. They can be overcome by special pre-treatment of the feed solution before membrane installation and addition of appropriate anti-scalant. The problem of scale formation is more serious in evaporators that are still used in the nuclear industry for the reduction of the volume of liquid wastes and operating at higher temperatures than membrane distillation. In such conditions, the proposed process run at moderate temperatures has particular value.
4. Conclusions The pilot plant experiments have proved the statement that membrane distillation can be applied for liquid low-level radioactive waste treatment. In one stage installation, all radionuclides are retained by the membrane. The distillate obtained in the process was pure water which can be re-utilised or safely discharged into the environment. The process avoids various problems inherent in normal evaporation, such as corro-
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sion, scaling or foaming. The entrainment of droplets is not possible because two phases are separated by the membrane. Operation at a low evaporation temperature can decrease the volatility of some volatile nuclides present in the waste, such as tritium, some forms of iodine and ruthenium. Preliminary economic analysis has proved an advantage of membrane distillation for low capacity plants utilising waste heat, e.g. MD employed in the nuclear power industry could utilise the waste heat from the cooling system of nuclear reactor. The data collected in pilot plant experiments justifies the conclusions: 1. Membrane distillation can replace or supplement other methods applied for the concentration of liquid low-level radioactive wastes. 2. The process, in some cases, can compete with other methods used for LLLW treatment because of a high retention coefficient, moderate process conditions, simple apparatus and the possibility of the utilisation of cheap energy sources. 3. Moderate process conditions enable the use of plastics instead of expensive stainless steel, eliminating corrosion and minimising the total costs of installation. 4. Contrary to evaporation, membrane distillation enables the removal of all radioisotopes from the distillate stream in a one-stage process. Pure distillate can be directly discharged into the environment or utilised as technological water. Membrane distillation has an advantage over reverse osmosis which has been already introduced into nuclear industry. The benefit of MD is a consequence of • the possibility of running the process to the high solute concentration required for fossilisation in concrete (up to 25% of solute), • gaining the high concentration in a one-stage process, • avoidance of the sorption of such ions as 50 Co2+ , 137 Cs+ , 134 Cs+ inside the membrane pores, • elimination of high pressures required by reverse osmosis and expensive high-pressure pumps, and
• less frequent washing cycles because of the elimination of fouling and sorption phenomena and minimisation of secondary waste generation. References [1] S.K. Sen Gupta, J.A. Slade, W.S. Tulk, Liquid Radwaste Processing with Crossflow Microfiltration and Spiral Wound Reverse Osmosis, AECL-11270, February 1995, Chalk River, Ont. [2] S.K. Sen Gupta, L.P. Buckley, S. Rimpelainen, A.Y. Tremblay, Liquid radwaste processing with spiral wound reverse osmosis, in: Proc. WM ’96 Conf., Tuscon, Arizona, 25–29 February 1996. [3] Advances in Technologies for Treatment of Low and Intermediate Level Radioactive Liquid Wastes, Technical Report Series No. 370, IAEA, Vienna, 1994. [4] R. Barnier, S. Caminade, L. Loudenot, M. Maurel, F. Courtois, Ultrafiltration treatment of laundry liquid wastes from a nuclear research centre, in: Proc. Conf. on Waste Management, Kyoto, 1989. [5] M. Tomaszewska, M. Gryta, A.W. Morawski, Study of concentration of acids by membrane distillation, J. Membr. Sci. 102 (1995) 113. [6] E. Drioli, Y. Wu, V. Calabro, Membrane distillation in the treatment of aqueous solutions, J. Membr. Sci. 33 (1987) 277. [7] K. Shoji, N. Shin-Ichi, S. Shun-Ichi, Transport phenomena in membrane distillation, J. Membr. Sci. 33 (1987) 285. [8] K. Schneider, T.J. Gassel, Membrane distillation, Chem. Ing. Tech. 56 (1984) 514. [9] G.C. Sarti, C. Gostoli, S. Matulli, Low energy cost desalination process using hydrophobic membranes, Desalination 56 (1985) 277. [10] S.J. Anderson, N. Kjellander, B. Rodesjo, Design and field tests of a new membrane distillation process, Desalination 56 (1985) 345. [11] P.A. Hogan, F.A.G. Sudjito, G.L. Morrison, Desalination by solar heated membrane distillation, Desalination 81 (1991) 81. [12] A.G. Chmielewski, M. Harasimowicz, G. ZakrzewskaTrznadel, Purification of radioactive wastes by low temperature evaporation, Sep. Sci. Technol. 32(1–4) (1997) 709–720. [13] A.G. Chmielewski, M. Harasimowicz, Application of ultrafiltration, Application of ultrafiltration and complexation to the treatment of low-level radioactive effluents, Sep. Sci. Technol. 30(7–9) (1995) 1779–1789. [14] A.G. Chmielewski, M. Harasimowicz, G. ZakrzewskaTrznadel, J. Palige, Membrane processes for liquid radioactive waste treatment, in: Proc. Third Int. Symp. ECCEE, Warsaw, 10–13 September 1996.
Concentration of radioactive components in liquid low-level radioactive waste by membrane distillation G. Zakrzewska-Trznadel ∗ , M. Harasimowicz, A.G. Chmielewski Institute of Nuclear Chemistry and Technology, Dorodna 16, 03-195 Warsaw, Poland Received 9 February 1999; received in revised form 3 May 1999; accepted 4 May 1999
Abstract The paper addresses some aspects of liquid low-level radioactive waste (LLLW) purification. Since the volume of liquid low-level wastes is usually large and the activity is rather low, the radioactive substances separated from the non-active portion have to be concentrated into the small volume for subsequent conditioning and disposal. The need for the improvement of decontamination and minimisation of the costs have led to new specific methods being under examination and development. The method proposed in the paper is membrane distillation. The experimental work described below supports the statement that membrane distillation can be an attractive alternative for liquid radioactive waste treatment. The advantages of membrane distillation over the other processes commonly used for the processing of LLLW are discussed in the paper. ©1999 Elsevier Science B.V. All rights reserved. Keywords: Membrane; Membrane distillation; Radioactive waste treatment; Decontamination
1. Introduction Radioactive wastes are generated in a number of different kinds of facilities. They arise in a wide range of concentrations of radioactive materials and a variety of physical and chemical forms. Low-level waste is defined as a waste which, because of its low radionuclide content, does not require shielding during normal handling and transportation. In Poland, one assumes a specific activity of liquid low-level waste (LLLW) < 107 Bq/m3 . There is a variety of methods for the treatment and conditioning of wastes prior to disposal. The selec∗ Corresponding author. Fax: +4822-811-15-32 E-mail address: [email protected] (G. ZakrzewskaTrznadel)
tion of a treatment process for an aqueous waste is determined by the physical, chemical and radiological properties of the waste. The processing of liquid waste aims at separating the radionuclides from the liquid phase and concentrating them in a solid waste form. The separation is pursued until the residual concentration or total amount of radionuclides in liquid phase is below the limits set by the regulatory body for the discharge of liquid waste from a nuclear facility as an effluent. The treatment technologies employed in nuclear industry include chemical precipitation, ion exchange, evaporation, membrane processes, solvent extraction and biotechnology. A number of factors have to be taken into account before the selection of a treatment process, including capital and operational costs, costs associated with the disposal of secondary wastes, etc.
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Fig. 1. MD pilot plant for radioactive waste concentration. 1: spiral-wound module; 2: distillate reservoir; 3: retentate tank; 4, 5: pumps; 6: heat supplier; 7, 8: heat exchangers; 9, 10: pre-treatment filters.
Membrane processes have already been employed successfully in nuclear industry. Reverse osmosis has been applied for a large scale operation in Chalk River Laboratory [1,2]. The plant is composed of a microfiltration stage, 50 spiral wound RO modules in a three-stage cascade, and tubular modules. Ultrafiltration is used for the separation of dissolved salts or acids from particulate and colloidal material (Sallefield, Paks, Harwell) [3]. The process can also be combined with physical and chemical processes of binding small ions with macromolecular ligands or complexation, known as ‘seeded ultrafiltration’ [4]. Microfiltration is often employed as a pre-treatment stage for RO. The pressure-driven membrane processes are very effective, especially when they are realised as a multistage process combining MF or UF and RO units, but they also have many limitations as fouling phenomena often resulting in cleaning operations which decrease membrane resistance and produce secondary wastes, or a need for the design of rather complex plants operating at high pressure. Some of these disadvantages can be avoided by membrane distillation (MD), non-isothermal process employing hydrophobic porous membrane, developed for water solutions concentration or for the production of pure water [5–11]. As the process is characterised by high retention of non-volatile solutes, large decontamination factors are expected in the separation of radionuclides which are present in liquid radioactive effluents in mainly ionic form. Laboratory tests have
shown that membrane distillation can be used for a concentration of radioactive solutions [12]. The results of laboratory experiments have led to MD pilot plant design and construction.
2. Experimental 2.1. Apparatus A scheme of the pilot plant is shown in Fig. 1. The productivity of the plant is up to ∼0.05 m3 /h of distillate which is a water of radiochemical purity (<10 Bq/dm3 for  and ␥ emitters). The installation consists of spiral-wound PTFE module G-4.0-6-7 (SEP GmbH) 1, distillate reservoir 2 (V = 80 dm3 ), retentate tank (V = 80 dm3 ) 3, two PRI 505M Ebara pumps 4 and 5, Omega heat supplier 6, two heat exchangers 7, 8 (JAD X-6/50 and JAD 6/50) and ceramic filters for pre-treatment (9,10). The process is conducted as a direct-contact membrane distillation. In such an arrangement, two streams at different temperatures directly contact the porous, hydrophobic membrane. The permeate condenses in a cold stream (distillate) which is pumped from reservoir 2 by PRI 505M Ebara pump through heat exchangers, pre-treatment units with ceramic filters to the PTFE module.The warm stream (retentate) is circulated from tank 3 by pump through the heat exchanger, heater and pre-treatment filters. Both streams — cold and warm — are returned to the reservoirs. The
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with an NaI/Tl scintillation crystal or an 800-channel analyser DIDAC with the spectrometric probe SCINTIBLOC (Intertechnique, France). For more precise concentration measurements at the beginning and at the end of the experiment, ionic chromatograph (DIONEX) and DSPEC Digital Gamma-Ray Spectrometer with germanium detector were used. 2.3. Experiments Fig. 2. Spiral-wound membrane module — cross-section of the spiral. 1: distillate channel; 2: retentate channel; 3: distillate inlet; 4: distillate outlet; 5: feedwater inlet; 6: feed water outlet; 7: membrane.
Table 1 Characteristic of spiral-wound MD module Overall dimensions φ × h Effective surface area Volume flow of feed streams Distillate flux Recommended feed stream temperature Average energy consumption per 1 m3 of the distillate PTFE membrane Pore size Porosity
0.45 m × 0.62 m 4 m2 300–1500 dm3 /h 10–60 dm3 /h 60–80◦ C ∼600 kW hth 0.2 m 80%
flows in the module are operated on the counter-flow principle. Fig. 2 shows the construction of the spiral membrane in the module and the flow arrangement. The flow rates in the module are set in such a way that sufficient heat-transfer coefficients and required distillate transport through the membrane are obtained. The process parameters are adjusted to ensure that the outlet temperature of warm stream is lower than that of the cold one and the recovery of the part of lost heat is possible in heat exchanger 7. Additionally, the distillate is cooled down in heat exchanger 8. The characteristics of the module as per the manufacturer is shown in Table 1.
2.3.1. The performance characteristic of PTFE module The experiments were conducted in the temperature range 35–80◦ C at the feed inlet and 5–30◦ C at the cold stream (distillate) inlet, with the feed and distillate flow rates set in the range 300–1500 dm3 /h. Under these conditions, the permeate stream 10–50 dm3 /h (60–300 dm3 /m2 per day) was obtained. The permeate flow rate was measured periodically with a graduated cylinder collecting the excess of the liquid flowing out of the distillate reservoir. The influence of process parameters (inlet temperatures, flow rates of both streams, cooling water flow rate, etc.) on the productivity of the installation was checked. A number of experimental cycles were run in which one of these parameters was changed. 2.3.2. Raising the concentration of non-active solutions of inorganic salts The preliminary tests were performed using solutions of pure inorganic salts: NaCl, NaNO3 and KCl. The initial concentration of salts in the feed solution was changed from 1 to 55 g/dm3 . The experiments were conducted at different temperatures in such a way that the cold stream rise returned to the retentate tank by the over-fall pipe; so the concentration in the warm stream is stable if the process is run properly. The flow rate of cold and warm streams was 960 l/h. Every 0.5–1 h, the samples from warm and cold streams were drawn and the conductivity of both solutions was measured.
2.2. Analysis The amount of inorganic solute in both streams was controlled with conductometer TDScan20 and the concentration of radioactive substances via specific activity measurements using standard probe equipped
2.3.3. Concentration of radioactive solutions The experiments were run in the same way as the tests with non-active solutions. The excess of water from the distillate tank was turned back into the retentate reservoir. As a feed, the model solu-
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Fig. 3. Productivity of MD installation vs. inlet feed temperature. Flow rates VD = VR = 720, 960, 1260 dm3 /h.
tions of caesium and cobalt salts with admixtures of radioactive 137 Cs and 60 Co (total specific activity ∼5000 Bq/dm3 ) were used and the samples of the original liquid low-level radioactive wastes of specific activity up to 104 Bq/dm3 were used as well. Every 1–2 h, the specific activity of samples drawn from the two streams using analyser DIDAC with NaI/Tl probe was measured. Additionally, the specific conductivity and pH of the samples were controlled. At the beginning and at the end of the experiment, the composition of both solutions was determined by ion chromatography and radioactivity was determined by use of DSPEC Digital Gamma-Ray Spectrometer with germanium detector.
3. Results and discussion The highest permeate volumes were obtained at high inlet feed temperature and high distillate and retentate flow rates. In Fig. 3, the dependence of permeate flow Vp on the inlet feed temperature for different cold and warm stream flow rates is shown. The flow rates of both feed streams circulated in the unit were set at the same level: VD = VR (720, 960, 1260 dm3 /h). The cooling water flow rate in heat exchanger 8 influenced the productivity of MD installation as well. The experiments showed that the increase in the flow rate of cooling water caused a slight increase in the permeate flux; but it induced the rise in the energy consumption for additional heating of the retentate too. In some cases, the significant increase in cooling seems to be uneconomic.
Fig. 4. Energy consumption per unit of the product.
Fig. 5. MD experiment with the solution of sodium nitrate.
The best conditions of running the process were at higher inlet feed temperatures (70–80◦ C). In this temperature range, the energy consumption per unit of the product was lower than for lower inlet feed temperatures (Fig. 4). Experiments with sodium chloride, sodium nitrate and potassium chloride showed that the ions of dissolved salts did not pass the porous membrane and that all of them were retained in the retentate. An example of the experimental results is shown in Fig. 5. The upper curve represents the conductivity of the retentate which was stable during the experiment for each initial concentration of the feed. The concentration of the feed solution was changed from 1.3 to 18.5 g/dm3 . For the same time period, specific conductivity of the distillate decreased from 250 S/cm at the beginning of the experiment to 40 S/cm at the end. The increase in feed concentration did not significantly influence the
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Fig. 6. The influence of salt concentration on permeate flux for different inlet feed temperatures.
permeate flow rate and did not change the retention factor. Fig. 6 presents the concentration dependence of permeate flux for different inlet feed temperatures. Up to the concentration 18.5 g/dm3 , no significant change was observed in the permeate flow rate. Preliminary experiments with radioactive compounds were done using the model solutions of 60 Co and 137 Cs. The experimental temperatures were as follows: Temperature of the retentate at the inlet to the module: T1 = 67.4◦ C Temperature of the distillate at the inlet to the module: T2 = 28.8◦ C Temperature of the retentate at the module outlet: T3 = 40.4◦ C Temperature of the distillate at the module outlet: T4 = 51.7◦ C The flow rates of the distillate and retentate were fixed at 960 dm3 /h. Under these conditions, the permeate flow was 27 dm3 /h (flux 6.75 dm3 /m2 h). During the experiment, the activity of the retentate did not change (the concentrated retentate was diluted by the water returned from the distillate reservoir) and the activity of the distillate was stable on the level of the natural background radioactivity. The retention of radioactive ions in the retentate was complete. Similar results were obtained for the samples of the original laboratory radioactive waste. During the experiment conducted under the same conditions as the experiments with model solutions, the specific conductivity, radioactivity and pH of samples drawn from
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two streams were controlled every hour. The results of the measurements are presented in Figs. 7 and 8. The specific conductivity of the retentate (Fig. 7) did not change during the experiment. The conductivity of the distillate decreased from 130 to 45 S/cm. Small sorption of radionuclides took place at the beginning of the test (Fig. 8: retentate curve), and after 12 h, it reached the fixed level. The radioactivity of the distillate did not change during the 20 h experiment and the pH too of both streams remained constant. The composition of the feed solution and the distillate at the end of experiment is shown in Table 2. Three parameters describe the efficiency of purification of radioactive liquid waste: • the volume reduction factor (VRF), which is the ratio of the waste volume before treatment to the volume of the residues containing the bulk of the radioactivity, • the decontamination factor defined as DF =
af vf total activity in feed = total activity in effluent a e ve
(1)
where af and ae are the specific activities of the feed and the effluent, respectively, and f and e are the volumes of the feed and the effluent, respectively, and • the retention coefficient (φ) that refers to the concentrations of the non-active ballast in retentate (cR ) and permeate (cP ): φ = 1−
cP cR
(2)
Decontamination factors and retention coefficients calculated for the data collected in the experiment are presented in Table 2. All retention coefficients were higher than 0.9, most of them close to 1. Decontamination factors for most of the radioisotopes present in the feed stream were very high. Actually, in the effluent, only Co-60 and Cs-137 were detected. For cobalt, DF was 4336.5, while for caesium, it was about 44. The rest of the radioisotopes were not detected in the distillate, so, according to Eq. (1), the decontamination factors for them were close to infinity. The rather poor results for caesium were expected: this radioisotope is hardly removable from the processed waste by most of the known methods. Employing seeded ultrafiltration in which caesium
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Fig. 7. Conductivity of the retentate and the distillate in time.
Table 2 Chemical composition of the waste sample used in the experiments and the effluent after the membrane distillation plant Ion
Concentration in the feed (mg/dm3 )
Concentration in the effluent (mg/dm3 )
Retention coefficient (φ)
Na+
SO42−
1060.6 207.1 21 33.7 87.2 5.7 744.2 1832.9 37.6
3.269 14.584 0.212 not detected 2.375 0.442 1.485 0.065 0.186
0.9969 0.9296 0.9899 1 0.9728 0.9225 0.9980 0.9999 0.9950
Radioisotope
Activity of the feed (Bq/dm3 )
Activity of the effluent (Bq/dm3 )
Decontamination factor (DF)
140 La
<6.53 E − 01 2.99 E + 03 5.26 E + 02 8.62 E + 01 3.73 E + 01 1.04 E + 01 3.39 E + 03 7.84 E + 00 2.95 E + 01 4.51 E + 03
not detected not detected not detected not detected not detected not detected not detected not detected 6.73 E − 01 1.04 E + 00
→∞ →∞ →∞ →∞ →∞ →∞ →∞ →∞ 43.8 4336.5
NH4+ K+ Mg2+ Ca2+ F− Cl− NO 3
133 Ba 170 Tm 114m In 192 Ir 110m Ag 65 Zn 134 Cs 137 Cs 60 Co
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Fig. 8. Radioactivity of the retentate and the distillate in time.
ions were bound with cyanoferrates (Cu2 [Fe(CN)]6 , Ti[Fe(CN)]6 ), decontamination factors equal to 32–45 were reached [13]. The decontamination factor obtained by use of thin-film composite membrane element NTR 739HF (NITTO) was 30 for caesium in model solutions, and 23 in original waste samples, while for cobalt, even DF = 515 were reached in the same plant [14]. The specific activity of the radioisotopes presented in Table 2 was measured using DSPEC Digital Gamma Ray Spectrometer with Germanium detector which has a much more higher geometrical efficiency than the NaI/Tl detector. Due to that, the activity of radioisotopes in Table 2 is higher than the total activity of the waste shown in Fig. 8. One of the more often-risen questions about membrane distillation reliability is the danger of membrane hydrophobicity loss after long-time operation. The membrane made of porous polytetrafluoroethylene, used in the experiments, resists the entry of pure water at 0.28 MPa, which is sufficiently large with respect to the operating pressures of the membrane distillation plant. The pressures at the retentate and distillate entrances to the module were kept on almost the same level, 0.01–0.02 MPa according to the manufacturer’s demand that they should not exceed 0.07 MPa. Usually, the pressure on the distillate side was adjusted as a few percentage units higher than that at the retentate entrance to diminish the risk of wetting. Stabilisation of flows in the unit was done before the membrane module was switched on; so the flow through the module was uniform and stable.
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In long-duration experiments performed for up to 80 h, wetting of the membrane was not observed. In fact, as model feed solutions, inorganic salts dissolved in water, which would increase the surface tension and interfacial tension, were used. Such compounds (sulphates, nitrates, chlorides) are mainly present in liquid radioactive wastes. All substances which decrease the surface tension (oils, surfactants, etc.) are rather dangerous for hydrophobic membrane, and have to be removed before the MD unit if they are present in processed waste. Most of the methods applied for liquid radioactive wastes processing require the removal of organic compounds at the pre-treatment stage. These substances need a special and different treatment by solvent extraction, oxidation or incineration. Small amounts of organic admixtures can be removed by adsorption, oxidation, phase separation by adduct formation, etc. Laboratory experiments in which the solutions of inorganic salts were filtered via MD showed that the achievement of 25% solute concentration in the retentate is possible. This concentration is the upper limit for solutions immobilised in concrete. In fact, usually, the limit of solute concentration is not reached because the dose limit is exceeded first. The problems of blockage and scaling are coupled with the application of almost all membrane processes. They can be overcome by special pre-treatment of the feed solution before membrane installation and addition of appropriate anti-scalant. The problem of scale formation is more serious in evaporators that are still used in the nuclear industry for the reduction of the volume of liquid wastes and operating at higher temperatures than membrane distillation. In such conditions, the proposed process run at moderate temperatures has particular value.
4. Conclusions The pilot plant experiments have proved the statement that membrane distillation can be applied for liquid low-level radioactive waste treatment. In one stage installation, all radionuclides are retained by the membrane. The distillate obtained in the process was pure water which can be re-utilised or safely discharged into the environment. The process avoids various problems inherent in normal evaporation, such as corro-
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sion, scaling or foaming. The entrainment of droplets is not possible because two phases are separated by the membrane. Operation at a low evaporation temperature can decrease the volatility of some volatile nuclides present in the waste, such as tritium, some forms of iodine and ruthenium. Preliminary economic analysis has proved an advantage of membrane distillation for low capacity plants utilising waste heat, e.g. MD employed in the nuclear power industry could utilise the waste heat from the cooling system of nuclear reactor. The data collected in pilot plant experiments justifies the conclusions: 1. Membrane distillation can replace or supplement other methods applied for the concentration of liquid low-level radioactive wastes. 2. The process, in some cases, can compete with other methods used for LLLW treatment because of a high retention coefficient, moderate process conditions, simple apparatus and the possibility of the utilisation of cheap energy sources. 3. Moderate process conditions enable the use of plastics instead of expensive stainless steel, eliminating corrosion and minimising the total costs of installation. 4. Contrary to evaporation, membrane distillation enables the removal of all radioisotopes from the distillate stream in a one-stage process. Pure distillate can be directly discharged into the environment or utilised as technological water. Membrane distillation has an advantage over reverse osmosis which has been already introduced into nuclear industry. The benefit of MD is a consequence of • the possibility of running the process to the high solute concentration required for fossilisation in concrete (up to 25% of solute), • gaining the high concentration in a one-stage process, • avoidance of the sorption of such ions as 50 Co2+ , 137 Cs+ , 134 Cs+ inside the membrane pores, • elimination of high pressures required by reverse osmosis and expensive high-pressure pumps, and
• less frequent washing cycles because of the elimination of fouling and sorption phenomena and minimisation of secondary waste generation. References [1] S.K. Sen Gupta, J.A. Slade, W.S. Tulk, Liquid Radwaste Processing with Crossflow Microfiltration and Spiral Wound Reverse Osmosis, AECL-11270, February 1995, Chalk River, Ont. [2] S.K. Sen Gupta, L.P. Buckley, S. Rimpelainen, A.Y. Tremblay, Liquid radwaste processing with spiral wound reverse osmosis, in: Proc. WM ’96 Conf., Tuscon, Arizona, 25–29 February 1996. [3] Advances in Technologies for Treatment of Low and Intermediate Level Radioactive Liquid Wastes, Technical Report Series No. 370, IAEA, Vienna, 1994. [4] R. Barnier, S. Caminade, L. Loudenot, M. Maurel, F. Courtois, Ultrafiltration treatment of laundry liquid wastes from a nuclear research centre, in: Proc. Conf. on Waste Management, Kyoto, 1989. [5] M. Tomaszewska, M. Gryta, A.W. Morawski, Study of concentration of acids by membrane distillation, J. Membr. Sci. 102 (1995) 113. [6] E. Drioli, Y. Wu, V. Calabro, Membrane distillation in the treatment of aqueous solutions, J. Membr. Sci. 33 (1987) 277. [7] K. Shoji, N. Shin-Ichi, S. Shun-Ichi, Transport phenomena in membrane distillation, J. Membr. Sci. 33 (1987) 285. [8] K. Schneider, T.J. Gassel, Membrane distillation, Chem. Ing. Tech. 56 (1984) 514. [9] G.C. Sarti, C. Gostoli, S. Matulli, Low energy cost desalination process using hydrophobic membranes, Desalination 56 (1985) 277. [10] S.J. Anderson, N. Kjellander, B. Rodesjo, Design and field tests of a new membrane distillation process, Desalination 56 (1985) 345. [11] P.A. Hogan, F.A.G. Sudjito, G.L. Morrison, Desalination by solar heated membrane distillation, Desalination 81 (1991) 81. [12] A.G. Chmielewski, M. Harasimowicz, G. ZakrzewskaTrznadel, Purification of radioactive wastes by low temperature evaporation, Sep. Sci. Technol. 32(1–4) (1997) 709–720. [13] A.G. Chmielewski, M. Harasimowicz, Application of ultrafiltration, Application of ultrafiltration and complexation to the treatment of low-level radioactive effluents, Sep. Sci. Technol. 30(7–9) (1995) 1779–1789. [14] A.G. Chmielewski, M. Harasimowicz, G. ZakrzewskaTrznadel, J. Palige, Membrane processes for liquid radioactive waste treatment, in: Proc. Third Int. Symp. ECCEE, Warsaw, 10–13 September 1996.