Patent Number: 051436540
Section: description

DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS An embodiment of the present invention will be explained with reference to FIGS. 1 and 2. In this embodiment, a concentrated radioactive liquid waste, such as a radioactive waste generated from an atomic power plant, is dried into the form of a powder, and then granulated into pellets. The pellets are charged into a container and solidified by a solidifying agent that is poured into the container to cover the pellets. A flowchart of the process of an embodiment of the invention is shown in FIG. 1. FIG. 2 shows a schematic representation of an apparatus for performing the process. In a first step 21, radioactive liquid waste from an atomic power plant, for example, preferably having radioactive nuclide(s) of known type is stored in a tank 1. The liquid waste is transferred from tank 1 to dryer 2, which may be a centrifugal thin-film dryer, for example. In step 21 the liquid waste is concentrated by drying it in dryer 2 to form a powder. It is preferred that the powder is further pelletized in a pelletizer 3 in a step labeled 22 in FIG. 1. Thereafter, the pellets are charged in container 4, as shown in step 23. Alternatively, as shown in step 23 in FIG. 1, the dried powdered waste can be charged in container 4 without the intermediate step of pelletizing. In accordance with the present invention, a solidifying agent is introduced into container 4 for solidifying the pelletized waste. In preparing the solidifying agent, first a concentration ratio o is determined in step 24. The concentration ratio o is determined by estimating what the concentration of the radioactive liquid waste will be with respect to its present state after concentrating the waste by drying it in dryer 2 and converting it into powder or pellet form for charging it in container 4. The distribution coefficient Kd of the solidifying agent is then determined on the basis of the estimated concentration ratio .alpha. in step 25. The solidifying agent with the desired distribution coefficient Kd is prepared in step 26 from one or more of a plurality of solidifying agent components selected according to the type of radioactive substances present in the waste and based upon each solidifying agent component's coefficient of distribution with respect to the type of radioactive substances present in the waste. In FIG. 2, for example, two solidifying agent components are shown as being contained in tanks 6a and 6b, respectively. The mixture of these solidifying agent components is controlled by a controller 5 in accordance with the desired distribution coeffioient Kd. Controller 5 controls the opening and closing of valves 10a and 10b, respectively, to deliver the appropriate proportions of the solidifying agent components from tanks 6a and 6b into solidifying agent tank 7. Then, the solidifying agent 7 is mixed with water from tank 8 in a mixing tank 9. The solidifying agent in tank 9 is then poured into the container 4 in step 27, and thereafter the contents of container 4 are hardened to a solidified body in step 28. After hardening, a final solidified waste is obtained. The final solidified waste contains approximately 8 to 10 times as great an amount of radioactive substances as a conventional cement-solidified waste having the same solidified volume because the conventional cement-solidified waste is produced merely by solidifying a radioactive liquid waste with cement in a container as it is without subjecting the waste to prior volume-reduction processing. Therefore, the container of solidified waste reduced according to the present invention has an 8 to 10 times greater radioactive concentration than that of the conventional cement-solidified waste of the same quantity. Table 2 shows the measured value of the distribution coefficient of each solidifying agent component with respect to the ions of a plurality of radioactive nuclides found in the radioactive waste of an atomic power plant. TABLE 2 __________________________________________________________________________ Measured Value of Distribution Coefficient of Solidifying Agent Components with Respect to Nuclides (Saturated Na.sub.2 SO.sub.4 solution, 25.degree. C.) Solidifying agent ml/g Sodium Calcium*.sup.1 Oxine-added Ion Cement silicate Zeolite Bentonite salt charcoal __________________________________________________________________________ Cs 1 90 20 100 50 1 C 70 10 0 0 500 -- Co 930 600 50 20 50 27000 Sr 20 4300 -- 5 50 300 Ni 2000 2000 50 20 50 27000 .alpha. waste 2000 2000 -- 200 -- -- __________________________________________________________________________ *.sup.1 calcium hydroxide --: no measured data The measurement of a distribution coefficient is explained with reference to the following example. Assuming that a concentrated radioactive liquid waste is a regenerated liquid waste of a desalting ion exchange resin (the main ingredient thereof being Na.sub.2 SO.sub.4) generated from an atomic power plant, 50 ml of saturated aqueous Na.sub.2 SO.sub.4 solution is charged into the tank. To this solution are added 0.01 .mu.Ci/ml of the ions of one of the six nuclides shown in Table 2 and thereafter 1 g of the articles of one of the solidifying agent components shown in Table 2 obtained by pulverizing the solidified component. After the elapse of time sufficient for reaching the adsorption equilibrium, the solution is separated from the solidifying agent component, and the concentration (.mu.Ci/ml) of the nuclide in the solution and the concentration (.mu.Ci/g) of the nuclide in the solidifying agent component are measured by X-ray measurement. The value obtained by dividing the measured value of the latter concentration by the measured value of the former concentration is the distribution coefficient with respect to the solidifying agent component. The distribution coefficient varies greatly in accordance with different radioactive nuclides and solidifying agent components. In the present invention, the composition of the solidifying agent is adjusted to obtain the desired distribution coefficient according to the concentration of the radioactive nuclide of a solidified radioactive waste having its volume reduced so that the amount of leaching of the solidified waste is equal to or smaller than that of a conventional cement-solidified waste of the same type and quantity. The solidifying agent comprises one or more of the solidifying agent components shown in Table 2. To determine the most effective solidifying agent component or mixture of components in preparing the solidifying agent, the various distribution coefficients shown in Table 2 are noted with respect to the type of radioactive substance contained in the waste to be solidified. An analysis of the considerations involved in preparing the desired solidifying agent is discussed as follows. Any given nuclide of the six nuclides shown in Table 2 is selected as a noticeable nuclide represented by j, and any given solidifying agent component shown in Table 2 is represented by k. The distribution coefficient of k with respect to j is represented by Kd.sub.jk. In the preparation of the solidifying agent, two cases are considered. In the first case, a single solidifying agent component is used for solidifying the radioactive waste. In the second case, a solidifying agent comprising a plurality of mixed solidifying agent components is used to solidify the radioactive waste. (1) The case of using a single solidifying agent component Let the amount of nuclide leached from a solid body be ##EQU1## wherein C.sub.j represents the concentration of the nuclide j in the solid waste. The intended condition is ##EQU2## If the concentration ratio of the radioactive nuclide j powdered or further pelletized from its original state as a liquid waste is .alpha..sub.j, formula (2) is represented as follows: ##EQU3## wherein Kd.sub.j1 represents the distribution coefficient of cement (i.e., represented by k=1). In the case (1) of using a single solidifying agent component, the single solidifying agent used is not ordinarily conventional cement, such as Portland cement and blast furnace cement, namely k.noteq.1. Although the distribution coefficients vary with respect to different solidifying agent components and radioactive nuclides, generally there is almost no nuclide dependence of the concentration ratio .alpha..sub.j obtained by volume reduction. In other words, .alpha..sub.j substantially has the same value with respect to any nuclide j. EXAMPLE 1 In the case of solidifying a dried powder of Cs, which has been concentrated by 10 times by volume reduction, with sodium silicate, the condition of formula (4) holds and is represented as follows when the data of Table 2 is substituted: ##EQU4## Additionally, in the case of using a single solidifying agent component, the amount of Cs or Co leached is not reduced with any solidifying agent component shown in Table 2 as compared with that of a conventional cement-solidified waste. However, it is advantageous to reduce the elution ratio, as shown in Example 1, by paying special attention to C.sub.s, which is a nuclide having a long half life. (2) The case of using a solidifying agent comprising a plurality of mixed solidifying agent components In this case, the general formula corresponding to formula (4) is represented as follows: ##EQU5## wherein Kd.sub.ja, Kd.sub.jb . . . represent the distribution coefficients of the respective solidifying agent components used: a (k=a), b (k=b), . . . ; W.sub.a, W.sub.b, . . . represent the mixing ratios by weights of the respective solidifying agent components; and the following relationship holds: EQU W.sub.a +W.sub.b +. . . =1 (1) EXAMPLE 2 In the case of solidifying a dried powder of Cs, which is concentrated by 10 times by volume reduction, with a solidifying agent obtained by mixing sodium silicate with cement, formula (6) is represented as follows: ##EQU6## wherein k=I means cement and k=b represents sodium silicate. Since Kd.sub.jl =1 and Kd.sub.jb =90 from Table 2, formula (78) is represented as follows: ##EQU7## Since W.sub.1 +W.sub.b =1, if W.sub.1 =0.89 and W.sub.b =0.11, the condition of formula (9) is represented by the following expression, and sufficiently holds: EQU 0.89+90.times.0.11=10.8&gt;10 EXAMPLE 3 In the case of solidifying a dried powder of Co and Cs, which are concentrated by 10 times by volume reduction, with a solidifying agent obtained by mixing sodium silicate and oxine-added charcoal with cement, formula (6) relating to Co and Cs is represented as follows: ##EQU8## wherein k =1 means cement, k =b represents sodium silicate and k =c represents oxine-added charcoal. From the data of Table 2, Kd.sub.jl =1, Kd.sub.jb =90 and Kd.sub.jc =1 with respect to Cs; and Kd.sub.jl =930, Kd.sub.jb =600 and Kd.sub.jc =27000 with respect to Co, and the conditions of the following three formulas hold when the data is substituted: ##EQU9## If W.sub.1 =0.6, W.sub.b =0.1 and W.sub.c =0.3 by solving the conditions of these three formulas, the formulas (11) and (12) hold and it is possible to greatly reduce the amount of Cs and Co leached as compared with that of a conventional cement-solidified waste. In Example 1, the result of formula (5) is 90, which leaves too much margin for the limit 10. When a solidifying agent is expensive, for example, it is more desirable from the point of view of cost to use a satisfactory amount of solidifying agent as in Examples 2 and 3 than to leave too much margin. In order to actually obtain the concentration ratio .alpha..sub.j in carrying out the present invention, a concentrated liquid waste is sampled from a storage tank or the supply tank and the concentration of the solid content (the portion which is to be powdered or pelletized as a result of the drying process) therein is measured, thereby calculating the concentration ratio .alpha. obtained by powdering and pelletization. As described above, there is actually almost no nuclide dependence of the concentration ratio o and, in fact, .alpha..sub.j takes almost the same value with respect to any nuclide j. In a standard concentrated liquid waste (the main ingredient is Na.sub.2 SO.sub.4, 20 wt %), .alpha.=6 to 8 in the case of powdering, and .alpha.=8 to 10 in the case of pelletization. The nuclide concentration C.sub.j is determined by .gamma.-ray measurement or by .beta.-ray measurement at the time of the above-described sampling measurement. A solidifying agent is prepared as a general rule by using the above-described formulas on the basis of the concentration ratio o obtained by measurement of the sampled liquid waste from the storage tank or the supply tank 1 (or from the drier 2) at every solidification process. Actually, however, since the concentration ratio o is substantially determined by the particular volume reduction process and the solidifying system that is used, as described above, it is more practical to use a solidifying agent prepared in advance that corresponds with that system. For example, .alpha. is about 10 in the case of pelletization, so a solidifying agent containing sodium silicate as the main ingredient is prepared in advance. An example thereof is the solidifying agent (called cement glass) prepared by mixing cement and sodium silicate described in Example 2. As the noticeable nuclide j, the six nuclides shown in Table 2 are fundamentally selected, but it may be more convenient or practical to use one of the following three nuclides contained in a liquid waste. ______________________________________ Cs-137 Representative nuclide Same group: generated due to the .alpha. waste, breakage of atomic fuel Sr-90 Co-60 Representative nuclide Same group: generated due to corrosion Ni-63 C-14 Not belonging to the above two groups ______________________________________ More simply, it is possible to select only Cs-137 as the noticeable nuclide which has a long half life (about 30 years) and radiates .gamma. rays, thereby facilitating measurement. Additionally, it is more logical in actual execution of the present invention to take the concentration, the content, the half life, etc. of a nuclide into consideration as well as the concentration ratio .alpha. when selecting the solidifying agent components and the mixing ratio thereof. For example, even if the concentration of Co-60 (half period: 5.8 years) mixed with Cs-137 (half period: 30 years) is about 10 times as high as that of Cs-137, the concentrations of both nuclides are on the same level in about 20 years and thereafter Cs-137 has a higher concentration. Therefore, if the control period (300 years in Japan) of the final disposal facility is taken into consideration, it can be said to be more logical to select a solidifying agent while selecting Cs-137 as the noticeable nuclide. In FIG. 3, a comparison is shown between the amounts of leaching of solidified wastes produced according to the present invention (Comparative Example I), and according to a conventional cement-solidified waste process (Comparative Example II). The amount of radioactive nuclide leached is represented as a value standardized on the basis of the amount of Cs leached in Comparative Example I as "1". The solidified waste in Comparative Example I is an embodiment of the present invention produced by drying a concentrated liquid waste to form powder, pelletizing the powder and solidifying the pellets with sodium silicate as a solidifying agent, while the solidified waste in Comparative Example II is a conventional cement-solidified waste produced by homogeneously solidifying a concentrated liquid waste with cement as the solidifying agent without first subjecting the waste to volume reduction processing. It is clear that according to the embodiment of the present invention, the effect of preventing leaching of the solidified waste is superior to that of the conventional cement-solidified waste. Further in accordance with another embodiment of the invention, the solidifying agent can be prepared so that the amount of leaching for the solidified body is restricted to a permitted value, such as one generally considered acceptable by the industry or set by an ordinance. If a permitted amount of leaching of a radioactive nuclide j is P.sub.j (Ci/year.multidot.ton) and the radioactive concentration of the nuclide is C.sub.j (Ci/ton), and the distribution coefficient of the solidifying agent with respect to the nuclide j is Kd.sub.jk, the condition of the following formula must hold in order that the permitted value is not exceeded. ##EQU10## That is, for keeping the amount of leaching nuclide lower than the permitted amount, the distribution coefficient of the solidifying agent must satisfy the condition of the following formula. ##EQU11## wherein A is a value determined by several factors, including the proportion of the solidifying agent and radioactive waste contained in the container, the density of the solidifying agent, and so forth. Assuming that the amount of leaching nuclide is regulated by the distribution balance between the nuclide and the solidifying agent, the value A is obtained by the following formula: EQU A=1/(r.times..rho.) (16) Wherein r is a proportion of the solidifying agent in the solidified radioactive waste in the container, and .rho. is the density of the solidifying agent. The radioactive concentration Cj in the solidified radioactive waste may be estimated beforehand by the radioactive concentration of the nuclide j in the tank and the concentration ratio .alpha.. In the case of solidifying radioactive waste with a solidifying agent obtained by mixing more than two solidifying agent components together, the solidifying agent may be prepared on a way similar to that practiced when meeting the conditions of formulas (6) and (7). That is, the solidifying agent is prepared by using the following formulas: ##EQU12## Further, in the case of a radioactive waste having a plurality of noticeable nuclides, the solidifying agent may also be prepared in the same way as disclosed in Example 3. Example 4 An example in which the noticeable nuclide is Cs-137 will be explained. The permitted amount of leaching nuclide of Cs-137 is assumed to be 0.3 Ci/year.multidot.to". The radioactive concentration of Cs-137 and the concentration of the solids content in the tank 1 are measured in a conventional manner. The concentration ratio .alpha. is obtained in accordance with the measured concentration of the solids content and in consideration of the particular concentration steps, e.g., the drying and pelletizing steps. Therefore, if the measured radioactive concentration in the tank 1 is 2 Ci/ton and the concentration ratio .alpha. is 5, the radioactive concentration of Cs-137 in the solidified radioactive waste is estimated to be 10 Ci/ton. Next, if the proportion of the solidifying agent in the container is 0.45 and the density of the solidifying agent (e.g., the mixture of cement and sodium silicate) is 1.7 ton/m.sup.3 (the density of the inorganic solidifying agent, e.g., cement or sodium silicate is about 1.5-2.5 ton/m.sup.3) the value of A becomes 1.3 (m.sup.3 /ton.multidot.y) according to formula (14). Therefore, the distribution coefficient of the solidifying agent must be larger than the following value. ##EQU13## The solidifying agent component is selected based upon the distribution coefficients shown in Table 2. If sodium silicate (50 wt%) and cement (50 wt%) are selected and mixed, the distribution coefficient is 46. Therefore, the mixture thus produced satisfies the condition that the amount of leached nuclide be less than the permitted level. Although in the examples of an embodiment of the invention given above, the liquid waste is concentrated by drying and forming the waste into a powder, pelletizing the powder, and solidifying the powder or pellets with a solidifying agent, the method and apparatus of the present invention are not restricted to these examples, but is also applicable to the volume reduction and solidification of a used ion-exchanged resin slurry that is concentrated into a liquid waste sludge. As a result or the present invention, it is possible to increase the amount of radioactive waste that can be charged into a solidified waste container since solid waste having a higher volume reduction ratio than that of conventional cement-solidified waste is contained within the container. As a result, overhead expenses incurred with respect to the waste disposal cost and storage thereof are reduced. While a preferred embodiment has been described with variations, further embodiments, variations and modifications are contemplated within the spirit and scope of the following claims.