Removal of Radioactive Ions. from Nuclear Waste Solutions by Electrodialysis

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1 Journal of NUCLEAR SCIENCE and TECHNOLOGY, 15(10), pp (October 1978). 753 Removal of Radioactive Ions from Nuclear Waste Solutions by Electrodialysis Sen-ichi SUGIMOTO Radia Industries Co., Ltd.* Received March 18, 1978 Revised June 21, 1978 Removal of radioactive ions was studied from low and medium level radioactive waste solutions by electrodialysis using ion exchange membranes. The test solutions contained "'Cs+, sosrz2+, 106Ru3+ or fission products (F.P.) as active ions and NaCl, Na2SO4 or Ca(NO3)2 as inactive coexisting salts. The decontamination factor of the active ions was in the order : 137Cs+ (greater than 99%) >90Sr2+>F.P.>106Ru3f. The dialysis time required to attain the saturation was the shortest for monovalent cations K+, Cs+ and Na+, intermediate for divalent cation Sr2+, and the longest for trivalent cation Ru3+. The ratio of the decontamination factor of an active ion ea to the desalination factor of an inactive ion eb was nearly equal to unity for "Na, '"Cs and 90Sr. On the other hand, the apparent selective permeability of an active ion (A+) against Na+ ion, Tha+ was higher than unity for all the active ions tested, and was in the order of 13ICs>90Sr>42K>24Na, where T.1,,''1+ is defined by the ratio of g A to gna+ with TA being the ratio of dilution of A in the diluate and gna+ being that of Na+ in the same diluate. The decontamination factor of the active ions did not depend significantly on the species and concentrations of the coexistent salts or on the concentration of the active ions. KEYWORDS: electrodialysis, ion exchange materials, membranes, low-level radioactive wastes, radioactive ions, decontamination foctor, liquid wastes, valence, radioisotopes I. INTRODUCTION When applied to treatment of low and medium level radioactive waste solutions from nuclear facilities, the electrodialysis method using ion exchange membranes has the following advantages : (1) The treatment operation is simple. (2) No regeneration processes are necessary. (3) The interference due to the coexistent ions is minor. These characteristics are those which cannot be obtained by ion exchange resin method. Electrodialysis therefore has a practical potentiality not only for small scale waste treatment of radioactive solutions, but also for other desalting processes and industrial waste water disposal. Generally there are fairly large amount of inactive ions in addition to a trace amount of radioactive ions, contained in the low and medium level radioactive waste solutions discharged from a nuclear power plant. The electrical conductivity of the liquid, which is required for electrodialysis, is mainly supported by the presence of these inactive ions. Therefore it is of practical interest to acquire knowledges about the effect of coexistent inactive ions on the decontamination process of the radioactive ions by electrodialysis. Although there have been several studies performed on electrodialysis of radioactive ions with the aid of ion exchange membranes")-"), few studies have been done from the aspects how coexistent salts affect the decontamination process or how this process or desalting * Oyagi -cho, Takasaki-shi. 35

2 754 J. Nucl. Sci. Technol., process depends on the time duration of dialysis. The present study therefore aimed to comparatively examine the degree of removal of active ions and that of inactive ions when they are coexistent in a solution. The typical hazardous and long-lived nuclides ''Cs, "'Sr, '"'Ru and the fission products (F.P.) were used as the active ions, while NaCI and Na,SO which are commonly found in the effluent waste of regeneration process of ion exchange resin, and also Ca(NO,), were used as the inactive salts. II. EXPERIMENTAL Figure 1 shows the flow sheet of the testing apparatus for electrodialysis and Fig. 2 schematically illustrates the cross section of the dialysis cell stack. Two solutions, diluate and concentrate, of the same initial composition were stored in the two reservoirs. Each of them was pumped into the dialysis cell in which they were subjected to dialysis treatment and then circulated back into the respective reservoir. The volume of the solo Fig. 1 Flow sheet of electrodialysis with ion exchange membrane used for removal of radioactive Waste Fig. 2 Schematic diagram of electrodialysis cell stack 36

3 Vol. 15, No. 10 (Oct. 1978) 755 tions was 10 1 for diluate and 1 1 for the concentrate. Since the treatment was done thus continually, the solution in the diluate reservoir was desalinated and decontaminated while that in the concentrate reservoir was enriched during the electrodialysis process. As the electrodialysis cell, a small scale unit, type Du Ob (Asahi Glass Co.) was used. Two types of membranes were employed : Ionics Inc., CMV-10 cationic and AMT-10 anionic membranes. Nine diluting and ten concentrating chambers were constructed by inserting eleven pairs of cation and anion exchange membranes alternatively into the cell. Across these chambers, the DC voltage of V was applied. The two solutions entered the cell at the bottom, flew through the respective chambers in parallel and left the cell at the top. While they flew in the cell from the bottom to the top, cations in the solution migrated through the cationic membranes (denoted with C in Fig. 2) whereas anions migrated through the anionic membranes (a). Sample solutions of 2~15 ml each were collected at the biginning of the experiment from the initial two solutions, at the end from the concentrate and at every 1 hr during the dialysis process from the diluate. The activity concentration was determined with a 2p-gas flow counter after the sample solutions were dried. Concentration of the salts was measured by flame spectroscopy. Designating the concentration of radioactive ion A before dialysis with A0 and that at time t with A the decontamination factor of this ion is given in percentage as )A=100 x (Ao-At)IAo Correspondingly, the desalination factor of inactive ion B is expressed as 2B=100x(Bo-Bt)/Bo The radioactive nuclides 'Cs+, "Sr" and 100Ru" were used in the chemical form of chlorides. Fission products were used in the form of nitrates after cooling for 2 yr. Since the composition of F.P. depends significantly upon their history, the treatment conditions and the cooling period, it may be said that the composition of F.P. used in this experiments lacks of the generality. But since the presence of F.P. is usually inevitable in the waste solutions from a nuclear reprocessing plant, it is highly desired to outline their behaviors in the decontamination process from the standpoint of nuclear waste disposal. III. RESULT AND DISCUSSION Table 1 shows the test results for solutions of and "K+ respectively in the presence of inactive NaCI. Table 2(a)-(c) carry the results for solutions of 'Cs', "Sr", '"Ru" and F.P. in the presence of inactive salt ; NaCl, Na2S0, or Ca(NO3)2. The preliminary results obtained for "Na+ and given in Table 1 indicates that their decontamination factors approach to the saturation values, both greater than 99% in Table 1 Dependence of 7,-A and 72B on dialysis period in solutions of "NaCI, NaCI and 42KC1 37

4 756 Table 2 Dependence of 72A and r/b on dialysis period in solutions of '"CsCI, 9-SrCl2, 10"Ru, F.P. and various inactive salts 38

Vol. 15, No. 10 (Oct. 1978) 757 about 3 hr. The decontamination factor for 24Na and the desalination factor for inactive NaCl agrees within the experimental error just as would naturally be expected.

5 Vol. 15, No. 10 (Oct. 1978) 757 about 3 hr. The decontamination factor for 24Na and the desalination factor for inactive NaCl agrees within the experimental error just as would naturally be expected. The results obtained for solutions of the radioactive ions in Table 2(a)-(c) may be summarized as follows : (1) Differences in the coexistent salt species and their concentrations do not give any significant influences on the decontamination factor of these radioactive ions. (2) The decontamination factor for "'Cs+ and "Sr' averaged over all the solutions irrespective of the difference in the coexistent salt species is 99.2 and 98.2% respectively, indicating that "'Cs+ can be decontaminated slightly easier than "Sr". Among the four active nuclides tested, the readiness of decontamination is in the order 4"Cs+>"Sr">F.P.> 40GRu3+, which corresponds with the increasing order of ionic valency except for F.P. (3) The decontamination factor gradually approaches to the saturation value in about 2-3 hr for Na+, K+ and Cs', whereas 3-4 hr are required for Sr". It keeps increasing even after 5~6 hr for the case of Ru'. (4) The ratio of the decontamination factor for active ion to the desalination factor of inactive ion )2A/72B is 1.00, 1.00, 1.01 and 1.00 respectively for 24Na', "K+, and "Sr". On the other hand, this ratio for "Tu" and for F.P. which also includes 106Ru", is less than unity for both the cases. The relatively small values of the ratio for the latter two cases seem to be attributable to the formation of complex ions formed around the heavy metal ions, about which will be discussed later. (5) Comparing the results given in Table 2 (b), (c) concerning to the degree of desalination 728 of inactive coexistent ions Na+ and Ca", it is understood that Ca' has slightly larger value of 72, than does Na+. This seems to be due to the higher selectivity of the membranes employed to Ca" than to Na+. (6) As seen in Table 2(b), the decontamination factor for '"Ru"- is significantly small. This is probably related to the fact that RuC13 tends to be hydrolyzed, forming various chloro aquo-complexes of positive, neutral and even negative charges which are difficult to permeate through the membrane. (7) The chemical form of 106Ru included in the solution of F.P. is different from the other ones in order to simulate the waste solutions from nuclear fuel reprocessing. An wet process, consisting of dissolution of the spent fuels with nitric acid and of extraction with TBP, is widely utilized for nuclear fuel reprocessing. Therefore in the waste solutions from a fuel reprocessing plant usually are contained F.P. among which '"Ru is present in the form of nitrate complex. The amount of '"Ru frequently exceeds 50% of the total activity concentration in the waste solutions. In Table 3 is listed a typical composition of radioactive wastes from nuclear fuel reprocessing plants"'. Ruthenium-106 included in these waste solutions is regarded to be in the chemical form of nitrato nitrosyl complex ions of charge 0 to 3+ which induce proton dissociation at dissolution or dilution with water"""). Therefore '"Ru discharged from fuel reprocessing plants has different complex form and ionic charge from 10GRu in the chloride test solutions and hence is expected to raise a difference in the decontamination factor. The reason that the decontamination factor of F.P. is higher Table 3 Typical composition of radioactive nuclides in waste solutions from fuel reprocessing plant 39

6 758 J. Nucl. Sci. Technol., in Table 2(c) than in Table 2(b) seems to be mainly due to the fact that F.P. in Table 2(c) contain the larger amount of ''Cs and alkali earth ions, e.g. "Sr which are more readily decontaminated than l"ru. But the difference in the chemical form of 10URu may be raising an additional effect. Selective Permeability: So far the dialysis efficiency has been discussed in terms of the decontamination factor )2, and the desalination factor 77B. But as can be seen in Tables 1,-2(a) these factors are very close to unity and hence are inconvenient to examine the difference in the dialysis behaviors among the active ions to be removed. Therefore the efficiency is discussed here with the aid of selective permeability Ta defined by the dilution factors ra and as follows : where Ao, A, B0 and B, are the concentrations of A and B ions in the diluate at the dialysis periods 0 and t, respectively. Usually the selective permeability is rather taken as Ta=- (1//,)/(gB) because the purpose of dialysis is to concentrate the solutionco'"" in contrast to the present case in which the purpose is the dilution. In Table 4 thus calculated Tg values are compared for 24Na+, 42K+, 437cs+ and "Sr" as A ions and inactive Na+ as B ion for the standard. Since the dialysis process can be regarded to have achieved the saturation at least in 4 hr, the average concentrations over the three periods, 4, 5 and 6 hr, are used as A, and B, values, which consequently stand for the final saturation values. In contrast to the results given in Tables 1~2(a), where ea/728 values are almost the same : 1.00(24Na+), 1.00(12K+) 1.01(137Cs+) and I.00(90Sr2+), T'1,1a+ values in Table 3 are significantly different from each other. Although the reproducibility of the experiment was not as good as to make a detailed discussion useful, it is clear that Tf.la+ values for "K+, '"Cs+ and "Sr"- are at least greater than that for "Na+. The selective permeability of dialysis depends on the selectivity of membranes used and on the mobilities of ions in them""3). It becomes larger as these two factors increase. Comparing mono- and di-valent cations, the former has the larger mobility but the poorer selectivity in general"". The ions H+ and K+, especially H+, possess very large mobilities and hence exhibit high values of the selective permeability. On the other hand, Na+ ion has relatively small mobility""" in spite of its being a mono-valent cation, and hence has small selective permeability. These seem to be the reason why "K+, '"Cs and "Sr raise higher selective permeabilities than "Na+ does in Table 4. Table 4 Apparent selective permeabilities of some radioactive ions against Na+ 40

7 Vol. 15, No. 10 (Oct. 1978) 759 I V. SUMMARY When electrodialysis with ion exchange membrane was applied to treatment of low and medium level radioactive solutions containing inactive electrolyte salts, the removal of both active and inactive ions proceeded rapidly in the initial period but then slowed down to saturation. The behaviors of removal of the ions may be summarized as follows : (1) The degree of decontamination decreased in the order, "Vs+>"Sr">F.P.> 100Ru3'. It was 98 to 99% for 'Cs+ and "Sr' when saturation was attained. (2) The degree of decontamination increased with time until it reached saturation. The length of time to attain the saturation was in the order, mono-valent ions (K+, Cs+ and Nat) <divalent (Sr2+) < trivalent ion (Ru3+). (3) The decontamination factor of the active ions was essentially independent of the coexistent inactive salt species and of their concentrations. (4) The decontamination factor was not affected by the radioactive concentration of the solution over the range of 10-1~ 10-4 pci/m/. (5) The ratio of the degree of decontamination of active ions "M.+, 421(*, '"Cs and "Sr to the degree of removal of various coexistent inactive ions was close to unity. But the apparent selective permeability values Tf.',0- were all greater than unity and in the order of '"Cs+> "Sr'>121:+>21Nat ACKNOWLEDGMENTS The author would like to thank Prof. M. Muramatsu for his helpful discussions and Mr. H. Aikawa for his assistance in the experiment. The paper is partially based on the experiments carried out at Japan Atomic Energy Research Institute, Tokai Laboratory. REFERENCES (1) SUGIMOTO, S., et al.: JAERI-memo 2670, (1967). (2) JAERI Nucl. Eng. Div.: JAERI-1012, (3) SHIMOKA W A, J., et al.: J. At. Energy Soc. of Japan, (in Japanese), 1(4), 225 (1959), (4) NISHIDOI, M., et al.: ibid., 2(8), 460 (1960). (5) ITo, M., et al.: JAERI (6) KAWASHIMA, T., et al.: Nagoya Koshi., (in Japanese), 6(9), 499 (1957). (7) KA W ASHIMA, T.: ibid., 9(5), 235 (1960). (8) KISO, I.: Kagaku no Ryoiki, (in Japanese), 28 4], 496 (1974). [ (9) ISHIKAWA, M. : ibid., 28(5), 405 (1974). (10) FLETCHER, J. M.: J. Inorg. Nucl. Chem., 1, 378 (1955). (11) JENKINS, I.L., et al.: ibid., 12, 346 (1957). (12) ODA, K., et al.: J. Elect. Chem., 25, 330 (1957). (13) YAMABE, T., et al.: J. Ind. Chem., (Japan), (in Japanese), 63(8), 1342 (1960). (14) YAMABE, T., et al.: ibid., 63(11), 1907 (1960). 41

Uranium (IV)-(VI) Electron Exchange Reactions in

Journal of NUCLEAR SCIENCE and TECHNOLOGY, 5[4], F. 179-486 (April 1968) 179 Uranium (IV)-(VI) Electron Exchange Reactions in Anion Exchange Resin, Tri-n-Octyl Amine and Tri-Butyl Phosphate Kozo GONDA*,