EP0030404B1 - Procédé pour diminuer nettement la quantité de déchets radioactifs nucléaires dangereux - Google Patents

Procédé pour diminuer nettement la quantité de déchets radioactifs nucléaires dangereux Download PDF

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EP0030404B1
EP0030404B1 EP80201147A EP80201147A EP0030404B1 EP 0030404 B1 EP0030404 B1 EP 0030404B1 EP 80201147 A EP80201147 A EP 80201147A EP 80201147 A EP80201147 A EP 80201147A EP 0030404 B1 EP0030404 B1 EP 0030404B1
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nuclides
waste
neutron
neutrons
transmutation
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EP0030404A1 (fr
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Richard Marriott
Frank S. Henyey
Adolf R. Hochstim
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Perm Inc
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Perm Inc
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    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21F—PROTECTION AGAINST X-RADIATION, GAMMA RADIATION, CORPUSCULAR RADIATION OR PARTICLE BOMBARDMENT; TREATING RADIOACTIVELY CONTAMINATED MATERIAL; DECONTAMINATION ARRANGEMENTS THEREFOR
    • G21F9/00—Treating radioactively contaminated material; Decontamination arrangements therefor

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  • the invention is in the field of nuclear waste control and is particularly directed toward the elimination of long-lived radioactive nuclides of nuclear reactor waste.
  • radioactive materials are stored in the containers in the ground or under the sea. With adequate safeguards, storage for about 30 years suffices to remove the harm from relatively short-lived radioactive nuclides, but the situation is quite different for the long lived wastes. Fortunately, the majority of the fission wastes have half-lives less than one year, which means that at worst they must be stored for 33 years to be reduced to 10- 10 of their original amount. However, eighteen fission waste products as well as all the actinide waste products have half-lives greater than one year, but less than 10" years, and it is these products that pose the long term storage problem.
  • the burial solution to the waste problem is based on the assumption that the geological formation will remain stable for the necessary containment period. While this assumption is reasonable for plutonium, for example, it is not evident for the longer lived wastes including the fission products Pd107, T c 99 , 1 129 , CS 135 and Zr 93 , as well as the actinides.
  • a more attractive technique involves the direct transmutation of the dangerous waste materials by neutron bombardment into innocuous materials, or at worst short lived radioactive species. Such a transmutation can be achieved, for example, by recycling waste products back into the reactor which produced them.
  • Such nuclear transformations have been discussed in the literature but have been found only applicable for effective elimination of the actinides produced by neutron capture e.g. "Advanced Waste Management Studies Progress Report", 8, BNWL-B-223 (1973); H. C. Claiborne, "Neutron Induced Transmutation of High-Level Radioactive Wastes", ORNL-TM-3964, 1, 24; and "High Level Radioactive Waste Management Alternatives", 4, 9, BNWL 1900 (1974).
  • the applicability of transmuting long-lived fission products as well as the actinides by neutron capture in reactors has not been regarded as practical since such a procedure reputedly produces more long term waste than it removes. '
  • This method will have the same disadvantages as mentioned above: the method will consume a rather high amount of neutrons and produce more long-term waste than it removes.
  • a more specific object of the invention is to propose an improved method for processing radio-active waste materials, whereby the rate of transmutation of the radio-active elements can be increased in excess of their natural decay rates for a more rapid conversion to stable nuclides.
  • the invention provides a method of decreasing the amount of long-lived fission products in radio-active waste materials, wherein these waste materials, after removal of stable and other constituents, are exposed to a neutron flux, in order to produce transmutations therein, and wherein the resulting product, after removal of stable and short-lived radio-active nuclides is re-exposed to a neutron flux.
  • This method is characterized by the following sequence of steps:
  • the invented method makes use of the steps of transmuting long-lived fission products by neutron capture, together with previous separation of stable and short-lived radio-active nuclides, just like in the method of GB-A-802,971.
  • a difference with that prior art method is, however, that the long-lived. nuclides to be exposed to a neutron flux are separated into components before the exposure to that flux and that each component is individually exposed to the flux. This brings about a better economy of neutron consumption since each component may be irradiated with an appropriate amount of neutrons.
  • the method will not produce more long-term waste since the exposure rates can be chosen individually.
  • transmutation may be defined as the change of one nuclide into another nuclide of the same or a different element by any nuclear process, natural or artificial
  • a beneficial transmutation can be defined as any transmutation which leads, or is part of a sequence of transmutation which leads, in a reasonably short time, from a long lived radioactive nuclide to a stable nuclide.
  • radioactive waste materials are re-cycled in a region of a high-flux of thermal neutrons to permit neutron induced transmutation.
  • Chemical and/or physical and/or isotope separation of the waste is performed both prior to and after neutron irradiation.
  • FIG. 1 A block diagram of the process in accordance with the invention is shown in Figure 1.
  • U 235 or other fissile material undergoes fission, splitting into various fission fragments and producing neutrons. Some of these neutrons are used up in maintaining the chain reaction, while others are used in transmuting the waste.
  • the waste products including the fission fragments and actinides produced by neutron irradiation of Uranium, Plutonium, and/or Thorium, are separated into various components, each component comprising one or more different elements of the waste nuclei. This separation is either chemical or physical or a combination of the two and may further include isotope separation.
  • isotope separation as for example employing a mass spectrometer, could be utilized for separation of all isotopes. Economic consideration would, however, dictate primarily a combination of chemical and physical processing. Those "good” components which include only short-lived and stable elements and which do not include long-lived hazardous radioactive substances are stored to allow the decay of short-lived substances. Those "bad” components containing long-lived radioactive substances are exposed to a high flux of neutrons in order to induce transmutation. After a certain amount of exposure, these wastes are recycled through the separation/irradiation loop.
  • the high neutron flux may be produced by any of a number of methods that are often referred to as flux-trapping. These methods allow the flux in some regions of the fission reactor to be significantly higher than in other parts, making use of the strong decrease of cross sections of increasing neutron energy from thermal to MeV regime neutrons. Flux-trap reactor designs are described in, for example, U.S. Patents 3,255,083; 3,341,420; 3,276,963; 3,175,955; and 2,337,475.
  • the high flux may, in the future, be produced independently of fission reactors, most notably by fusion reactors. In this case economy of reaction utilization is not critical as copious supplies of neutrons can be produced with little accompanying radioactive waste.
  • Figure 1 illustrates the inventive method generally.
  • reaction economy is an important factor i.e. fission produced sources
  • the preferred chemical/physical separation techniques is to be carried out as a two-stage process as illustrated in Figure 2.
  • stage 1 reactor products are separated into components designated, for the sake of illustration A, B, D and D. Each component, once separated is maintained in a separate channel isolated from other components and fed to the high flux region. After transmutation in the high flux region the output of any given channel will generally contain some smaller amount of the original component remaining together with additional elements. These additional elements may be "good" products designated G i , G 2 ....G 6 , or other components which are long-lived and require further processing. The original component of each channel is thus separated in stage 2 from these additional elements as illustrated in figure 2. The recycling then occurs from the output of the stage 2 separation to the high flux region. Isotope separation may be part of stage 1 and/or stage 2 separation. Further, a specific rest stage may be provided before and/or after exposure to the neutron flux to permit ⁇ decay where desired prior to further neutron exposure.
  • Figure 3 shows a general format utilized in describing the decay/transmutation sequence
  • figure 4 illustrates, as an example, a portion of a chart of some nuclides illustrating the transmutation possibilities.
  • Natural ⁇ decay transmutations change a nuclide into another shown directly above it, while artificial neutron induced transmutations take a nuclide into another immediately to the right.
  • a and (3+decay are not significant, and for simplicity, only one isomer of each nuclide has been considered.
  • the values shown on the vertical lines connecting nuclides are the half-life of the transmutation in hours, while the values on the horizontal line are neutron cross-sections in barns.
  • Fission yields per 100 fissions are also given in the chart.
  • the direct fission yield is almost completely to neutron rich nuclides not shown, which would occur below those shown. These neutron rich nuclides rapidly undergo a series of (3 decays, as tabulated by Rose, P. F. and Burrows, T. W., ENDF/B Fission Product Decay Data, Aug. 1976, BNL-NCS-50545 (ENDF-243), to those nuclides which are illustrated on the chart.
  • the yield shown on the charts of Figure 3 and 4 is therefore the same yield as the direct fission products of the same atomic weight.
  • Sr 9 ° is a long-lived radioactive nuclide which is desired to be removed. Therefore, the transmutation from Sr 90 to Sr 91 is a beneficial transmutation. Sr 91 naturally transmutes in a short time to stable Zr 91. On the other hand, the other neutron induced transmutations shown are not beneficial and must be minimized by choice of the separation/irradiation loop. Y 89 ⁇ Y 90 , for example, does not involve long-lived nuclides at all and therefore the induced neutron transformation simply wastes neutrons. Sr 89 ⁇ Sr 90 not only wastes neutrons but also procudes a long-lived nuclide.
  • the Sr89 is allowed to naturally transmute to Y 89 prior to insertion of Sr into the high neutron flux region.
  • Y is then chemically separated from the Sr to prevent its otherwise neutron usage, and the Sr is exposed to the high neutron flux to transmute to Sr 90 to Sr 91 .
  • Table 1 lists 18 long-lived radioactive fission products of concern. These "bad" nuclides are broken-up into two groups, the first group having half-lives less than 100 years, and the second group having half-lives greater than 30,000 years. In addition, there are actinide wastes notlisted. In reference to figure 2, there may be up to 18 separate separation/irradiation loops for the fission products and an appropriate number of loops for the actinides, one loop for each substance.
  • the "bad" nuclides considered for elimination are listed in Table 1 and are defined primarily by the amount of radio-activity they are responsible for in the waste, after the waste has been stored for a certain length of time. Their half-life is not too long, else they provide very little radioactivity. Their half-life is not too short, else they decay during the storage period. They must be present, or at least have the possibility of being present, in a sufficiently high concentration to contribute significant radioactivity.
  • a conservative level of activity at which a substance can be considered nearly safe is half the activity of an equal amount of U 238 .
  • This criterion is in agreement with the cutoff in half lives of 10 10 years, twice the half life of U 238 (and also twice the age of the Earth).
  • the required storage time as a function of half life is shown in figure 5, with the bad nuclides indicated by dots.
  • this criterion requires a storage time of 33 years.
  • the lower group of bad nuclides requires up to 3,000 years of storage, while the upper group requires at least a million years for every nuclide in that group, and up to 1/30 the age of the earth.
  • the transmutation process must safisfy at leasuhreecriteria; (1) it must consume less energy than was produced when the waste was created, (2) it must generate of itself less hazardous waste than that destroyed, and (3) must eliminate waste materials at a rate significantly greater than their natural decay rate.
  • Previous studies reported in ERDA-76-43, vol: 4 indicate that only neutron absorption processed can satisfy the first criterion.
  • the major source of neutrons at present are the fission power reactors themselves. Therefore the issue of the second criterion is whether the number of neutrons produced in the power reactor is sufficient to transmute all or a substantial amount of the long lived waste produced along with those neutrons.
  • the computer program further allows initial separation and periodic separation between exposures to a high thermal neutron flux.
  • a strategy for each nuclide is presented which is sufficient to meet the three criteria for successful transmutation.
  • each reactor is responsible for precessing its own waste or (2) several "power” reactors send their waste to one "transmutation” reactor.
  • the second alternative is most probably not viable since it wastes the neutrons.
  • the first possibility allows the exchange of waste between different reactors; one, for example, might handle all the cesium while another handles all the zirconium, with appropriate design differences between the reactors. The important consideration is that neutrons not be wasted.
  • any given nuclide which we label by its atomic weight A and its atomic number Z, may undergo ⁇ decay to the nuclide (Z+1,A) or it may undergo neutron absorption to the nuclide (Z,A+1).
  • the first two terms determine the loss of the nuclide due to its decay and neutron absorption while the last two terms determine its gain due to the decay or transmutation of other species.
  • nuclide which has a lower nuclide in the chain with a smaller value of ⁇ .
  • Z,A a nuclide accompanied by a lighter isotope (Z, A-1) such that ⁇ Z,A -1 ⁇ Z,A
  • the transmutations of the 18 bad isotopes have been analyzed for periods of up to about 100,000 hours (about 11-1/2 years) of irradiation in a flux of 10 16 neutrons/cm2 sec. From Table 2, one can see that this time period ranges from orders of magnitude more than enough time to remove a nuclide to less than one half-life. The improvements of removal rate over natural decay varies from a few percent to a factor of over 1 08 .
  • the isotope separation provides an absolute optimum situation, and provides a measure of the inefficiency of chemical separation.
  • chemical separation two possibilities are treated for some waste components. In the first case, it is assumed that there is control over the time between fission and chemical separation. This situation is possible in liquid fission fuel reactors where the fuel and waste may, for example, be continuously cycled without shut-down of the reactor, In the second, the time between fission and separation is assumed long, as for example, in solid fuel fission reactors. The extra control allows one to separate nuclides which would otherwise decay into another element.
  • Table 3 shows the results of the computer analysis for chemical separation and for isotope separation.
  • the two cases of chemical separation are labeled a and b for separation with and without timing respectively.
  • the table is ordered by increasing half life and divided into the first and second groups previously defined in relation to Table 1.
  • a very rough measure of the hazard of nuclear waste is the total amount of each of the two groups of bad nuclide. (This measure neglects differences in biological activity, in ease of storage (i.e., geochemical effects, in half life within a group, and in the nature of the radiation emitted).
  • the waste starts with about 15 atoms per hundred fissions for the low group and 20 atoms of the high group.
  • the high group would contain a small amount of Sn 126 , an amount of Sell depending on its cross section but less than .055 atoms, and a trace of Zr 93 . All other bad nuclides would be removed. About 32 of the up to 127 neutrons would be used.
  • the remaining elements most notably the large amount of Cs 137 , can have the partially processed part of the element combined with the part of that element freshly separated from the recent fission waste, as indicated by the solid lines in Figure 2.
  • the bad nuclides are discussed in order of increasing atomic number and increasing atomic weight. It is noted that in developing a strategy for each individual bad nuclide, there is no interdependency between the individual bad nuclides except in the case of Pr 147 and Sm 151 .
  • the decay/transmutation chain for Se 79 is shown in Figure 7.
  • the thermal neutron absorption cross section for Se 79 is shown in Figure 7.
  • the thermal neutron absorption cross section for Se 79 is not reported (probably owing to its low abundance) and may be assumed small. Thus one may assume that no significant reduction of this isotope is possible. If however the neutron absorption cross section is found to be significant, the separation/irradiation process may be utilized with Br and Kr removed periodically to enhance neutron economy.
  • Figure 8 shows the decay/transmutation chain for Kr 85 .
  • the remainder ends up as Rb 85 , because the rapid ⁇ decay chain passes through an excited isomer of Kr 85 which ⁇ decays to Rb 85 .
  • Kr 83 has by far the largest neutron absorption cross section.
  • Kr 84 and Kr 86 each have a cross section about 1/20 of Kr 85 .
  • Kr 83 When the Kr is subject to the neutron flux Kr 83 is converted into Kr 84 . At this point the ratio of Kr 84 to Kr 85 is about 5. This ratio cannot exceed 20 (the ratio of the cross sections of Kr 85 and Kr 84 ). Therefore, after about 75% of the Kr 85 has been transmuted, it becomes difficult to convert any more Kr 85 . For every 20 atoms of Kr 84 converted to Kr 85 , only 21 are converted to Kr 86 . Meanwhile about 30 Kr 86 's are transmuted. Thus it takes about 70 neutrons to gain 1 Kr 85 .
  • Figure 9 The results of the actual circulation are shown in Figure 9 to 48,000 hours (somewhat over 5 years) at which time 75% of the Kr 85 is gone. At about this time, the natural decay of Kr 85 actually removes if faster than continued exposure to neutrons. Only isotope separation would improve matters.
  • Figure 9 also shows the removal of Kr 85 as compared to its natural decay. Also shown are the average and marginal usage of neutrons showing the effects to Kr 83 at very small times and Kr 84 at large times. Different scales are used for amounts and neutron usage.
  • Sr 90 is transmuted according to the exponential law, since the only other isotope of Strontium in the waste is stable Sr 88 which transmutes to Sr 89 with a very small cross section. Sr 89 is initially present in the waste (as well as being produced from Sr 88 ), but it decays with a half life of 1250 hours, short compared to the relevant time scale for the transmutation of Sr 90 .
  • the program WASTE was utilized with chemical processing every 3000 hours. No indication of undesirable effects of the build-up of Yttrium and Zirconium were noted. Clearly a much less frequent chemical processing for separation of Y and Zr from Sr would suffice. Only 1.06 neutrons are required for each transmutation of the first 96% of the Sr 90 . The effective half life of Sr 90 in a flux of 10 16 neutrons per square cm per second is 2-1/4 years.
  • More of the nuclear waste is Zirconium (15%) than any other single element.
  • the stable isotopes in the waste are Zr 91 , Zr 92 , Zr 94 , and Zr 96 .
  • Zr 93 has a larger neutron absorption cross section than any of the stable isotopes, it is not large enough to make transmutation easy.
  • Figure 10 shows a portion of the decay/transmutation chain including Zr 93 .
  • the initial chemical separation is carried out in a time short compared to two months after the fission process.
  • the Yttrium is separated from the Zirconium.
  • the y 91 with a half life of about two months, decays into stable Zr 91 , which therefore is isolated from the Zr 93 .
  • the Zr 92 , Zr 93 , Zr 94 , Zr 95 and Zr 96 are then allowed to stand, allowing the Zr 95 to decay (with a half life also of about two months).
  • the Zr 91 is included. After 50,000 hours (about 6 years) the Zr 91 has added an extra 6% of the original Zr 93 by the sequential transmutation chain, Zr 91 ⁇ Zr 92 ⁇ Zr 93 atom removed. In both cases, neutron economy is enhanced by periodically separating Zr from Nb and Mo.
  • Tc 99 whose decay/transmutation chain is shown in Figure 12, provides one of the most favorable cases for transmutation. It has a reasonably large cross section for neutron absorption and there is only the single isotope, Tc 99 , in the waste. Therefore the removal follows an exponential curve (with an effective half life of 42 days). After chemical processing, all the Tc can be combined, since it is all Tc 99 . With chemical processing every 300 hours, the neutron usage is 1.03 neutrons per transmutation. The extra 3% comes from absorption in Ru 100 which builds up for the 300 hours.
  • Ru 106 whose decay/transmutation chain is shown in Figure 13, has a half life of 1.01 years, just above the cutoff of 1 year. It requires 33.4 years to decay to half the activity of U 238. It has a very small neutron absorption cross section, 146 barns, requiring an exposure to neutrons for 31.4 years to reduce the activity to the same level. The saving of 2 years is not deemed worth the trouble and expense of cycling and processing. Therefore Ru 116 may be treated as a short-lived isotope, storing it for at least 33-1/2 years before allowing it to enter the environment.
  • the decay/transmutation chain for Pd 107 is shown in Figure 13. There is not much Pd 107 in the waste since it is on the high side of the lighter bump in the fission yield curve.
  • Pd 105 with an atomic weight smaller by only 2, has 6 times the fission yield.
  • Ru 106 has an intermediate yield, and decays to Pd 106 (in two steps) with a half life of 1 year. Ruthenium is assumed to be separated from the Palladium before a significant amount of it has been allowed to decay (if processing occurs within half a year after fission, the results are not substantially modified).
  • the Pd 105 has a cross section 40% larger than Pd 107 , which when multiplied by the factor of 6 in yield gives a conversion rate 8.4 times that of Pd 107 Pd 107 transmuted to Pd 108 , which, with a roughly comparable cross section, converts to Pd 109 which rapidly decays.
  • Pd 107 has a cross section 40% larger than Pd 107 , which when multiplied by the factor of 6 in yield gives a conversion rate 8.4 times that of Pd 107 Pd 107 transmuted to Pd 108 , which, with a roughly comparable cross section, converts to Pd 109 which rapidly decays.
  • Neutron economy would dictate removal of Pd from Ag and Cd prior to recycling into each new irradiation step.
  • Sn 126 and Sb 125 are shown in Figure 14. These isotopes occur at the minimum of the yield curve, and are present in very small amounts. As a result, neutron economy is not of paramount importance and the products I and Te need not generally be separated. They are treated together because exposure of tin to neutrons produces Sb 125 .
  • the cross section for Sn 126 is very small, so transmutation is very slow. After about 12 years in a flux of 10 16 neutrons/cm 2 sec., one third of the original Sn 126 still remains. However, this corresponds to .3% of any one of the five most common bad isotopes. Sb 125 is easily removed.
  • 1 129 (Figure 15) is removed following an exponential curve, since 1 128 is highly unstable.
  • the neutron use does not exceed 1.2 neutrons per 1 129 atom transmuted, as the 1 129 is accompanied by 1/5 as much I 127 .
  • the situation is even more favorable since the cross section for I 127 is smaller.
  • the average use was 1.1 neutrons per transmutation, as only about half of the I 127 was removed.
  • Iodine may readily be separated from the fission waste and is thus a very favorable element for waste transmutation.
  • C S 135 and Cs 137 are somewhat separate problems, and are discussed separately.
  • Cs 134 is not a direct fission product and therefore occurs in small amounts in the waste. It has a large cross section and is easily removed in the treatment of Cs 135 and Cs 137 .
  • Figure 16 shows a portion of the decay/transmutation chain including Cesium.
  • Cs 135 The major problem with the treatment of Cs 135 is the large amount (6.75 atoms/100 fissions) of stable Cs 133 in the waste.
  • Cs 133 has a considerably higher neutron absorption cross section than Cs 135 , and must absorb 3 neutrons before again becoming a stable nuclide.
  • the chain is Cs 133 ⁇ nCs 135 ⁇ nCs 136 ⁇ Ba 136 .
  • Xe 133 has a half life of over 5 days, and Xe 135 has a half life of over 9 hours.
  • Xe 135 has an extremely high neutron absorption cross section (3x10 6 barns) and stable Xe 136 has a very small cross section (.16 barns).
  • Stable Xe 134 also has a rather small cross section (1.73 barns).
  • Cesium may be efficiently treated.
  • a liquid fuel reactor would clearly be desirable in achieving these short processing times, since the processing may be continuous.
  • Xenon As soon as Xenon is separated out, it is exposed to a high neutron flux for a short time. At 10 16 neutrons/cm 2 sec., the optimum time is 11 minutes. In the example of Table 3, 20 minutes was used. After this irradiation the Xenon is removed from the flux and stored for, say, two months for the Xe 133 to cecay (this is about 30 half lives of Xe 133 ). After this, the Xenon left is not radioactive. There may be a further separation of the Cesium produced in the first two hours after fission. Thus there are three, possibly four, places in which Cesium is produced.
  • the Cesium produced before separation consists of some or most of the Cs 137 and a small amount of Cs 135 and Cs 133 .
  • the amounts depend on the time before separation.
  • the Cesium produced during the irradiation is some or most of the Cs 137 , and a small amount of C S 135 and Cs 133 .
  • Cs 137 is produced, and it might be desirable to keep it with the Cesium produced in the first two steps.
  • most of the Cs 133 is produced. This Cs 133 would not be subject to further irradiation.
  • the amount of Cs 135 that is contained in with this Cs 133 is 4.4x10 ⁇ 6 atoms per 100 fissions (22 parts per billion of the waste), coming from the equilibrium between Xe 134 and Xe 135 .
  • Figure 17 shows the amount of Cs 133 and C S 135 produced up to end of the time of irradiation as a function of the time of separation. If, for example, separation is accomplished in 6 minutes, there will be 0.055 atoms of Cs 135 per 100 fissions. In the first two hours after irradiation 0.08 atoms of Cs 133 per 100 fissions out of a total of 6.75 are produced.
  • Cs 137 has the smallest neutron absorption cross section of any of the bad nuclides (with the possible exception of Se 79 ). Irradiation in a flux of 10 16 neutrons/cm 2 sec. only brings the effective half life to 12 years, compared to 30 years for natural decay. Aside from a small amount of C S 137 generated from Cs 137 removal follows an exponential curve (most Cs 136 decays to Ba 136 ).
  • Pm 147 (figure 19) is the lightest isotope of Promethium in the waste, and the only one with a large half life. It is transmuted following an exponential curve with an effective half life of 4-1/3 days in a flux of 10 16 neutrons/cm 2 sec.
  • Sm 151 there is also a small amount of the bad isotope Sm 151 created from the Sm 150 , the amount depending on the frequency of chemical separation of the Samarium from the Promethium (the Samarium is then not exposed to any more neutrons). In the example of Table 3, chemical processing was assumed to occur every 2 hours. The amount of Sm 151 produced is comparable to the amount of Sm 151 left after the irradiation of the Samarium waste. It is not feasible, without isotope separation, to transmute this Sm 151 .
  • the Sm 151 (figure 19) is accompanied by a larger amount of Sm 149 .
  • Sm 151 has a large neutron absorption cross section (1.4x10 4 barns) but Sm 149 has an even larger cross section by nearly a factor of five.
  • Europium (figure 19 & 21) is one of the heaviest elements in the waste, and occurs in small amounts. Eu 152 and Eu 154 do not occur directly as fission products. What little Eu 152 does occur will be rapidly transmuted while the Eu 155 is being removed, as will the Eu 151 coming from the Sm 151 that decayed before it was transmuted.
  • Transmutation of Eu 155 per se is very simple, as indicated by the "isotope separation" columns of table 3.
  • the waste contains some Eu 153 , which is converted to radioactive Eu 154 . Therefore, in order to remove the Eu 155 it is necessary to convert the Eu 153 by the chain There is 5 times as much Eu's3 as Eu 155 , and it requires 5 neutrons for conversion to Gd 158 leading to neutrons per atom of Eu 155 removed.
  • the Eu 156 would have time to decay, terminating the chain at Gd, saving up to 40% of the neutrons.
  • Figure 22 shows the time development of the amounts of Eu 154 and Eu 155 .
  • the initial Eu 155 is rapidly transmuted away, while the Eu 154 builds up almost as fast as the Eu 155 is removed.
  • the Eu 154 continues to build up while it, in turn, transmutes to Eu 155 .
  • the Eu 154 and Eu 155 are in equilibrium with the remaining Eu 153 and then are removed at a rate determined by the Eu 153 cross section.
  • the amounts and composition of the actinides depends on the parameters of the reactor system such as the enrichment of the Uranium and the integrated flux to which it has been exposed.
  • U236 Fifteen percent of the U 235 on absorbing a neutron does not fission but produces U236. This U236 would be only moderately expensive in neutrons to transmute except that it is mixed in with all the U 236 in the spent fuel. It is impossible from the point of view of neutron economy to put the U 238 in the high flux region. If the U 236 produced from all U 235 by neutron irradiation is mixed in with the amount of U 238 accompanying that much U 231 in natural uranium, the radioactivity is double that of U 238 . Therefore, U 236 does not pose a serious hazard if combined with the U 238 .
  • the U 236 will absorb a neutron, giving the transmutation chain If this Np 237 is exposed to as high a flux as possible, it first transmutes to Np 238 which then fissions if it absorbs a neutron or decays to P U 238 . The fissioning is preferable on the grounds of neutron economy.
  • U 238 if it absorbs a neutron, becomes PU239
  • This plutonium (as well as the Pu 238 discussed above) is, on separation, used as a fissionable substance. In thermal fission, 3/4 is fissioned and 1/4 becomes Pu 240 .
  • the Pu 240 absorbs a neutron becoming Pu 241 , Three-fourths of Pu 241 ' fissions and 1/4 becomes Pu 242 .
  • Pu 242 and heavier isotopes are not fissionable with high probability and form part of the heavy actinide waste. By sequential neutron absorption, these eventually lead to fission.
  • Fissionable isotopes include Cm 245 , Cm 247 , Bk 250 , Cf 249 , Vg 251 , and Es 254 .
  • the number of neutrons required per atom of Pu 242 is large, the very small quantities involved makes the neutron usage not have a serious effect on the total neutron economy. This is consistent with the conclusion of earlier studies that actinides can be reduced by transmutation.
  • the neutron economy is approximately a net loss of one neutron for each atom of Np 237 (including the absorption on U 236 ) and approximately a balance (a small net loss) for each atom of U 238 transmuted.
  • the fission wastes from Plutonium and other actinides increase the amount of fission wastes that must be processed by the excess neutrons from the fission of U 235 .

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  • Solid-Sorbent Or Filter-Aiding Compositions (AREA)
  • Treatment Of Sludge (AREA)
  • Monitoring And Testing Of Nuclear Reactors (AREA)
  • Treatment Of Liquids With Adsorbents In General (AREA)

Claims (3)

1. Procédé pour réduire la quantité de produits fission à longue durée de vie dans des matières de déchets radioactifs, dans lequel ces matières de déchets après retrait des constituants stables et autres, sont exposées à un flux de neutrons afin de produire des transmutations dans ceux-ci, et dans lequel le produit résultant, après retrait des nucléides radioactifs stables et à courte durée de vie, est réexposé à un flux de neutrons, caractérisé par la séquence de phases suivantes:
- on sépare les nucléides radioactifs ayant une durée de vie relativement courte et les nucléides stables de la matière de déchets initiale et on stocke au moins certains d'entre eux, on sépare la matière de déchets restante en constituants individuels, chaque constituant comprenant essentiellement environ un nucléide radioactif de durée de vie relativement longue ou une combinaison de nucléides sélectionnés dans le groupe comprenant Se79, Kr85, Sr9o, Zr93, Tc99, Pd107, Sb125+Sn126, l129, Cs135, Cs137, Pm'147, Sm151+Eu et les actinides, on expose les constituants précités à l'exception éventuellement du Se79, Sn126+Sb125, CS137 et Pm147 individuellement à un flux de neutrons afin d'induire des transmutations dans ceux-ci, on sépare les nucléides radioactifs de durée de vie relativement courte et les nucléides stables de chacun des constituants après exposition et on stocke au moins certains d'entre eux, on sépare la matière restante de chaque constituant individuel après exposition en d'autres constituants, chaque autre constituant comprenant environ un nucléide radioactif de durée de vie relativement longue ou une combinaison de nucléides selectionnés à partir du groupe précité, on combine dans la mesure du possible ceux des autres constituants qui contiennent les nucléides correspondants, on répète les phases d'exposition et de séparation au moins une fois avec au moins certains des dits autres constituants, et on stocke les constituants après qu'ils aient atteint un niveau réduit de radioactivité au cours de leur décroissance naturelle.
2. Procédé suivant la revendication 1, caractérisé en ce qu'on sépare Y9' du constituant comprenant Y93 avant exposition.
3. Procédé suivant la revendication 1, caractérisé par la phase supplémentaire de séparation du xénon gazeux à partir des matières de déchets dans un état initial et par l'exposition au même flux de neutron pour empêcher la formation de CS135,
EP80201147A 1979-12-05 1980-12-03 Procédé pour diminuer nettement la quantité de déchets radioactifs nucléaires dangereux Expired EP0030404B1 (fr)

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AT80201147T ATE19561T1 (de) 1979-12-05 1980-12-03 Verfahren zur deutlichen verringerung gefaehrlicher, radioaktiver, nuklearer abfallmaterialien.

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US10065879A 1979-12-05 1979-12-05
US100658 1993-07-30

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EP0030404B1 true EP0030404B1 (fr) 1986-04-30

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WO1985004752A1 (fr) * 1984-04-09 1985-10-24 Gerrit Berdinus Engelen Procede de transformation systematique de nucleides
JPH073474B2 (ja) * 1990-07-13 1995-01-18 動力炉・核燃料開発事業団 放射性廃棄物の消滅処理方法
RU2003191C1 (ru) * 1993-01-18 1993-11-15 Игорь Петрович Еремеев Способ трансмутации изотопов
ITCO20130066A1 (it) * 2013-12-16 2015-06-17 Wow Technology S P A Metodo per trattare una soluzione/dispersione acquosa contenente almeno una sostanza radioattiva ed impianti che lo realizzano
JP6106892B2 (ja) * 2015-03-20 2017-04-05 株式会社東芝 放射性廃棄物の処理方法

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EP0030404A1 (fr) 1981-06-17
AU539393B2 (en) 1984-09-27
AU6435380A (en) 1981-06-11
DE3071586D1 (en) 1986-06-05
ZA807201B (en) 1982-02-24
ATE19561T1 (de) 1986-05-15

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