Environ. Sci. Technol. $903, 17, 425-431
NOTES High-Level Radioactive Waste from Fusion Reactors Joachim Gruber Hahn-Meitner-Institut fur Kernforschung GmbH, D 1000 Berlin 39, Federal Republic of Germany
Highly radioactive waste is generated in fusion reactors by neutron irradiation of the reactor blanket. All major elements of construction, Le., the alloy constituents of stainless steel, the moderator graphite, and the neutron multipliers, beryllium, bismuth, or lead, as well as the impurities in each, are likely to be activated. Since every element might be present, the activity produced during the expected operating lifetime of a typical fusion reactor is estimated for each element. A measure of the maximum concentration of each element that can be tolerated in drinking water is defined by integrating the expected activity and International Commission on Radiological Protection guidelines. Finally, these tolerance levels are compared to ranges of concentrations found in natural waters to identify elements that may be serious environmental hazards. Of all elements, nickel, silver, molybdenum, niobium, chlorine, and calcium are most suspect.
Introduction The nuclear fusion reactor is a potential energy source of the future (I). The process providing the energy will be the fusion of two hydrogen isotopes, deuterium and tritium, which yields a helium atom and a neutron. The neutron emerges from the nuclear reaction zone, the socalled plasma, with a high speed carrying an energy of 14 MeV, the larger part of the nuclear energy released by the fusion reaction. Since tritium cannot be obtained in sufficient quantities except from nuclear reactors, fusion reactors will be designed to generate (“breed”) their own tritium supply. Since for every tritium nucleus consumed in the plasma one neutron is emitted, sufficient tritium is bred if every neutron produces one tritium nucleus. Thii is achieved in an assembly (“blanket”) of three materials: a neutron multiplier, e.g., lead, beryllium, or bismuth, a neutron moderator, e.g., graphite, and the breeding material, which consists mainly of lithium. When lithium atoms are exposed to neutrons they break apart, yielding helium and tritium atoms. The neutrons must have a low enough speed, and the graphite serves the purpose of slowing them down. In this process the graphite is in turn heated up by the neutrons. The purpose of the addition of the multiplier to the blanket is to compensate for neutrons that fail to fission lithium atoms. Ultimately, after having penetrated the blanket the neutrons are absorbed in a shield composed of steel and borcarbide. On their way through blanket and shield the neutrons will change the matter from inactive to highly radioactive. If stainless steel is employed as structural material in fusion reactors, its activity will not differ much from the activity of the stainless steel of the core barrel of a fission reador since thermal neutron irradiation is the main source of activation in both types of reactors, and the fluxes are for the most part similar. Moreover, the weights of the 0013-936X/83/09 17-0425$01.50/0
core barrel on the one hand and the most active part of a fusion reactor, the wall surrounding the plasma, on the other hand, are similar, Le., 6-8 tons for reactors having a thermal power of 500-1000 MW. The most hazardous part of the activity within a fission reactor is considered to be the irradiated fuel element, and its activity is practically invariably connected with the fission process. In contrast, the amount and composition of future fusion reactor waste will depend largely on the choice of the materials of which it is made. This provides fusion new degrees of freedom with respect to the objectives and strategies of waste management, because the choice of the blanket materials is still open. There are materials that could be used in construction of reactor blankets that would produce only short-lived activity (if all impurities of these materials could be eliminated that produce long-lived activity), but unfortunately they are apparently impractical at present; stainless steels are the most likely choice. This paper will identify elements likely to be present in activated steels and the moderator and multiplier materials that, if released to groundwaters, would excessively contaminate it for drinking. For the elements that may result in higher than permitted concentrations if brought in contact with groundwater directly after shutdown of the reactor, a graph will show which will dominate the concentration and how long they will need isolation.
Construction of Blanket and Shield In order to give an impression of what the radioactive parts will look like, the arrangement of the blanket and shield around the plasma is shown, taking as an example the INTOR (International Tokamak Reactor). INTOR has been designed at the headquarters of the International Atomic Energy Agency (IAEA) in Vienna, Austria, by a group of European, Japanese, US., and USSR scientists during the past few years. It belongs to the group of intermediate devices that will be operable by the turn of the century, between the present-day fusion experiments like the Tokamak Fusion Test Facility in Princeton, NJ, and the demonstration reactors that will show the technological feasibility to generate energy by fusion. In the blanket of INTOR 500 MW will be generated. In Tokamaks the plasma is confined within a horizontal torus by magnetic forces. In Figure 1 (2) one can see the right half of a vertical cross section through the torus and its surrounding parts. It shows the first wall, blanket, and shield. Through the opening in the blanket the plasma will be heated. The torus is built of an array of wedgeshaped sectors, one of which is shown in Figure 2 (2). Also the blanket itself is composed of modules (2), one of which is depicted in Figure 3. The blanket has a thickness of
0 1983 American Chemlcal Society
Envlron. Sci. Technol., Vol. 17, No. 7, 1983 425
CENTER LINE OF REACTOR FIRST WALL?
COOLANT OUTLET LINE
,-BLANKET SHIELD
HEATING PORT COOLANT INLET
INLET PLENUM
DIVERTOR TO
-
0
OHMIC HEATING '-TOROIDAL FIELD COIL COIL
1
2m
Flgure 1. Simplified cross section through INTOR (after Abdou et al.
(2)). FIR ST WALL,
,-BLANKET
HEATING PORT
DIVERTOR SLOT
TOROIDAL SHIELD SECTOR
Figure 2. Flrst wall-blanket-shield sector of INTOR (after Abdou et at. (2)).
50 cm; therefore, the height of the steel box in Figure 3
is 50 cm.
Neutron Flux The amount of activity generated by a neutron flux depends on its spectrum, i.e., the number and energy of neutrons in the flux. Although the spectra of different reactor designs differ considerably,they also have common features, which reflect the compromise between the technologically possible and the economically desirable. (a) 14-MeV Neutrons. 14-MeV neutrons are produced in the plasma chamber. Economics calls for a high power output of the reactor, i.e., of the order of several thousand megawatts. Engineers, on the other hand, want the dimensions of the plasma chamber to be as small as possible. These demands taken together drive up the neutron flux that penetrates the first wall. Material scientists predict the failure of a steel wall after a critical exposure of several megawatt years per square meter (1 MW = 4.5 X lo1' neutrons/s). Today the neutron flux chosen is several megawatts per square meter (which implies a short lifetime of the first wall). Since the neutron flux decreases by 2 orders of magnitude within the 50-cm thick blanket, the radiation damage and activation is concentrated in the first wall. (b) Slow Neutrons. Within the blanket, 14-MeV neutrons are moderated by graphite, producing slow neutrons that are absorbed by lithium to produce tritium and do not cause excessive radiation damage. Blanket designers maximize that flux by optimizing the balance be428
Environ. Scl. Technol., Vol. 17, No. 7, 1983
P~ASMA
Figure 3. Blanket module of INTOR (after Abdou et at. (2)).
tween fast-neutron moderation on the one hand and absorption and leakage of slow neutrons on the other. The slow-neutron flux is likely to be between 10l2 and 1013 neutrons/(cm2 9). Those high fluxes are seldom apparent in blankets though, since-if technically possible-lithium is placed wherever such high fluxes appear, to maximize production of tritium. The blanket will also contain neutrons that have not yet been well moderated, i.e., that have energies a little too high to effectively split lithium atoms. And there will be 14-MeV neutrons that have been elastically reflected from the nuclei of the blanket; at the first wall their number is about equal to the number entering from the side of the plasma. All those additional neutrons should be considered, too, in precise activation calculations (3-5) but will in this paper be neglected, for three reasons (6): (a) they do not lead to different radioisotopes than the ones considered; (b) their contribution depends on the design of the reactor, and the optimum design is still in process of development; (c) only the order of magnitude of the activity of the high-level radioactive waste coming from fusion reactors is relevant, today. For the upper limits a and b of the neutron fluxes discussed above, the amount of waste having a half-life longer than 10 years generally increases proportionally to the number of incident neutrons-the exceptions will be pointed out below. Therefore the flux of 14-MeV neutrons can be assumed arbitrarily as 4.5 X 1013cm-2 s-l, corresponding to 1 MW/m2. The thermal neutron flux is assumed to be equal to the 14-MeV neutron flux.
Neutron Actiuation Reactions In principle in an individual activation reaction the number An of waste nuclei generated during an operation time At of the reactor is proportional to both the number $At of neutrons and the number n of exposed nuclei, the mother nuclei. The ratio An/(n$At) is called the neutron activation cross section u. It has been determined directly by experiment or calculated with the help of a nuclear model that correlates the basic properties of the nucleus, such as the number of protons and neutrons and their individual sizes, with the size of the nucleus and the chance of a nuclear activation reaction. The activation cross sections for most thermal neutron reactions used here are from the German chart of the nuclides (7). The rest of the cross sections have been taken from a compilation of Alley and Lessler (8). They were used in the Plowshare project for calculating the activity induced in soil by peaceful nuclear explosions. As the compilation gives u values for most of the nuclides, even radioactive ones, it is a useful data base for
Table I. Cross Sections and Effective Half-Lives of Mothers Involved in Nonlinear Activation Calculations cross mother(n,x)daughter section, barns 150Sm(n,7)151Sm 152Sm(n,2n)151Sm 151Sm(n,7)1 Vrn 5 1 E ~ ( n , 75)21 E ~ 51Eu( n, 2n)' SoEu '5 3 E ~ ( 2nn, ) l 5 2 E ~ '5 3 E ~ ( n , 7 ) 1 5 4 E ~ '"Hf( n,7)17'Hf* 1'7Hf(n,7)''8Hf 1911r(~,~)1921r* 1921r*(n,7)1931r lg31r(n,2n)'921r*
931r(n,7)1 941r
1.0 x 2.1
lo2
1.5 x 5.9 x 6.1 X 7.6 x
104
3.9 x 1.0 x
103 10-l
10-l 102
lo-'
3.7 x 102 9.2 X 10' 1.1 x 103 5.9 x lo-' 1.1x l o 2
Table 11. Elements Present in Larger Than Trace Amounts within the Blanket' A1 As B
T,,years 4.9 2.3 X 10' 3.3 x 8.3 X 8.1 X 10' a 1.3
co
Cr cu Fe Mn Mo Nb Ni
a 1.3 5.4 4.5 a 4.5
x 10-1 x lo-'
Effective half-life determined by reaction in the next line.
activation calculations of the kind presented in this article. On the other hand many of its cross sections have been determined by using nuclear structure theory, and therefore a discussion of possible variations of the activities is necessary. It is greatly eased by a linearization of the activation calculations: Except in the reactions given in Table I, the number n of mother nuclei will be assumed to change only insignificantly, Le., An/n = a4At is assumed to be small. As long as the cross section a is of normal size, i.e., below cm2 = 10 barns, this is guaranteed by the technical constraints to the number 4At of 14-MeV neutrons entering each square meter of the first wall. They keep 4At below several MW year/m2 and An/n below several percent. The neutrons penetrating the initially nonradioactive blanket react with the stable nuclei, thus yielding a first generation of new nuclei, called daughters. Those, in turn, are exposed to the neutrons, Le., they act as mothers for a second generation of daughters, and so on. The sequence of generations is called an activation chain. Because of the exponential decay of radionuclides with time it is not possible to accumulate radioactive nuclei in any generation during a period At longer than roughly their half-life T . Thus the number of radioactive daughter nuclei cannot become larger than An N na4T, which is less than 1 X lo+ of the number n of mothers if the reaction cross section is normal and the half-life is in the order of hours. Such radionuclides terminate the activation chain, i.e, in the next generation of the chain the number of particles will be negligible compared with the number of first-generation daughters even of an impurity of the material considered. An investigation of the chart of the nuclides shows that n one can expect a very large fraction of all k ~ o w long-lived radionuclides ( T > 10 year) to appear in the blankets in the first generation of activation chains if one assumes at least traces of each of the 83 elements to be present in the blanket materials. In a few cases the long-lived radionuclide does not appear among the first generation of daughters, but instead a short-lived radionuclide takes that place. But after ita formation it decays entirely, yielding the long-lived waste nuclide. Therefore, in the calculation these cases can be handled without loss of accuracy by neglecting the intermediate short-lived nuclide and assuming that the activation of the mother leads directly to the long-lived waste nuclide. An example is the reaction 'wMo(n,2n)ggMo,which has a half-life of 2.8 days. All of the decays end in Tc99, which has a half-life of 2.1 X lo5 year.
N a
in Steel (316SS) P
5 x 102 3 x 102 10 5 x 102 1.8 x 105
103 6.3 x 105 2 x 104 2 x 104 5 x lo2 1.4 x 105 3
S
2 x 102
lo2
Si Ta Ti V
7 x 103
C N 0
6X
in Graphite 0
x 10'
2 x 102 102 2 x 103
lo2
lo2
2 x 102
3 x 102
Concentrations ( w )in grams per ton.
Results In Table I1 the elements are specified that appear in amounts higher than traces in the blanket. All other elements are arbitrarily taken to be present as traces of l mg/ kg of structural, multiplier, or moderator material. Furthermore, beryllium is assumed to be used as oxide (beryllia) and lead, bismuth, and graphite without additional metals or binders. The calculation specifies the activities q generated in 1 cm3 of blanket material (steel, BeO, Bi, C, Pb; unit, Ci/cm3): q = na4AtAF
(1)
where n = number of mother nuclei in 1 cm3 of blanket material (particles/cms), and
n = P(w/M)fNA
(2)
where w = weight of mother element in 1 g of blanket material (g/g), M = weight of 1 mol of mother element (g/mol), f = fraction of mother isotope present in mother element; i.e., natural abundance of mother isotope, N A= number of particles in 1 mol (particles/mol), p = density (g/cm3) of blanket material (steel, 7.7;beryllia, 2.5; Bi, 9.8; C,2.3; Pb, 11.4), a = activation cross section (cm2),4 = neutron flux, only 14-MeVneutrons and thermal neutrons [ (Maxwellian distribution of energies for a moderator temperature of 25 OC): 4.5 X 1013 cm-2 s-l for 14-MeV neutrons as well as for thermal neutrons], At = operation time of a blanket module (2 years), X = In 2 / T (year-'), except in footnote g to Table 111, where it is in seconds-', T = half-life of radionuclide (year), and F = factor necessary to get the results of the activities in Curie (Ci); numerically F is the reciprocal of the number of decays per second that is equivalent to 1Ci (F = 2.7 X 10-l'). The cross sections influencing the generation of 15'Sm, 152Sm, 150Eu, 1 5 2 E ~1 ,5 4 E ~178Hf*, , lg21r*,lg31r,and lg41r are very large (Table I): mothers or daughters have a large tendency to capture the thermal neutrons. Their number decreases during reactor operation, accordingly. One can attribute them an "effective" half-life T , = In 2 / ( a 4 ) ,which is defined analogously to the radioactive decay time. It is the time during which the number decreases by a factor of 2. As mentioned above, the number of daughers increases only during an operation time roughly equal to the half-life. Similarly one can estimate the effect of a decreasing number of mother nuclei: the activation reactions come to an end when the mother has disappeared. This roughly happens after about 1half-life of the mother. The activities for the radionuclides mentioned have been calculated by taking into account these effects. Envlron. Sci. Technol., Vol. 17, No. 7, 1983 427
Table 111. Data and Results of Activation Calculations mother(n,x)daughter
T, years 1.6E+ 6'" 5.7E+ 3
T l ( n ,y)36C1 40Ca(n,y)41Ca 58Ni(n,y )59Ni 62Ni(n,y)63Ni 78Se(n,y)79Se gzZr(n,y)g3Zr 93Nb(n,7 )94Nb 92Mo(n,y)g3Mo 98Mo(n,7)99T~ "Ru( n,y ) 9 7 T ~ Io6Pd(n,y)Io7Pd 07Ag(n,y ) l 08Ag* 112Cd(n,y)13Cd* IZoSn( n,y )lZ1Sn* 128~e(n,~)1291 34Xe(n,y )' 35Cs 32Ba(n, ) l 3Ba 144Sm(n,y)145Pm 150Sm(n,y)151Sm 51E~(n,y)152E~ 62Er(n,7 )lS3Ho 16sHo(n,7)166Ho* 77Hf(n,y)178Hf* Is5Re(n,y)186Re* Ig1Ir(n,y)'9zIr* lgzPt( n,y)lg3Pt n,y ) 2 0 5Pb 204Pb( zosBi(n,y)210Bi*
3.OE+ 5 1.3E+ 5 7.5E+ 4 l.OE+ 2 6.5E+ 4 1.5E+6 2.OE+ 4 3.5E+ 3 2.1E+ 5 2.6E+ 6 6.5E+ 6 1.3E+ 2 1.5E+ 1 5.OE+ 1 1.6E+ 7 2.OE+ 6 1.1E+ 1 1.8E+ 1 9.3E+ 1 1.2E+1 3.3E+ 1 1.2E+ 3 3.1E+ 1 2.OE+ 5 2.4E+ 2 5.OE+ 1 1.4E+ 7 3.5E+6
1.6E+ 6
4N(n,p)I4,
5.7E+ 3 5.7E+ 3 7.23+6 3.OE+ 5 1.3E+ 5 3.7E+ 6 7.53+4 l.OE+ 2 1.5E+ 6 1.5E+6 1.OE+4 l.OE+ 8 l.OE+ 8 1.4E+ 1 3.5E+ 3 2.1E+ 5 2.6E+ 6 2.1E+ 5 1.5E+ 1 5.OE+ 1 5.OE+ 1 1.6E+ 7 3.OE+ 1 1.1E+ 1 6.OE+ 4 9.3E+ 1 3.5E+ 1 1.2E+ 1 1.5E+ 2 3.3E+ 1 2.4E+ 2 5.OEt 1 1.4E+7
428
u,
barns
ALI,MPC, Ci/a,Ci/m3
q, Ci/cm3
BHP, m3/cm3
(A) Thermal Neutron Activation Reactions 9.2E- 3 3E- 5b 2E-6' 5E-12 9.OE-4 8E- 3b 3E-7d 6E-10 4.3E+ 1 2E- 3 2E-8 4.OE- 1 2E- 3 6E-10 4.6E+ 0 3E- 1 8E-4 1.4E+ 1 1E- 2 9E-2 5.3E-1 2E- 3 2E-10 2.6E- 1 2E- 3 3E-12 1.2E+ 0 8E- 4 3E-6 3E- 3 4E-6 5.OE- 2 1.3E- 1 3E- 3 2E-7 2.5E- 1 3E- 2 4E-13 2.9E- 1 3E- 2 9E-13 3.OE+ 0 8E- 4 9E-7 6.OE- 2 2E- 5 7E-8 1.OE- 3 2E- 3 4E-10 5E- 6 3E-13 2.OE- 1 8E- 4 8E-13 2.5E-1 2E- 3 4E-8 8.5E+ 0 7.OE- 1 3E- 5b 7E-8 l.OE+ 2 1E- 2 8E- 8 1E- 3 4E-4 5.9E+ 3 3E- 5b 4E- 8 1.9E+ 1 3.5E+ 0 3E- 5b 1E-7 1.OE- 7 3E- 5b 2E-14 1.1E+ 2 2E- 3 9E-9 1E- 3 3E- 6 9.2Et 2 1.4E+ 1 9E- 3b 9E-8 3E- 3 4E- 8' 6.6E- 1 1.4E- 2 8E- 4 2E-7' 2E-13
7E- '1 2E- 6 4E-4d 8E- 7 8E- 5 2E- 6 2E- 2 7E+ 1 8E- 7 1E-8 3E- 2 1E- 2 5E- 4 1E- 10 2E- 10 9E- 3 3E- 2 2E- 6 5E- 7 8E- 9 2E- 4 2E- 2 6E- 5 3E+ 0 1E- 2 3E- 2 7E- 9 4E- 5 2E- 2 1E-4 1E-4' 2E- 3' 2E- 9
(B) Fast-Neutron Activation Reactions ' 1.3E-1 3E- 5b 2E-7f 2E-13 3E-10 3.6E- 2 8E- 3b 3E- 8f 1E-9 3E-7 2E-10f 8E- 3b 5.3E- 2 6E-13 3E-10 3E- 5b 1E-10 5.9E- 3 9E-11 1.5E-1 2E- 3 8E-13 2E- 3 9.3E- 2 5E-7 3.6E- 1 5E- 2 2E- 5 3.8E- 1 3E- 1 2E-3 1.1E+O 1E-2 1E-11 1.5E+O 2E- 3 1E-9 2E- 3 4.5E- 2 4E-7 1.5E- 2 3E- 5b 2E-10 3E- 5b 4.1E-1 2E-10 3E- 5b 6.OE- 2 1E- 3 8E- 3 3.3E- 1 2E-5 3E- 3 5.6E- 1 3E-6 3E- 3 4.4E+ 0 5E-13 3E- 2 9.4E- 1 7E-13 3E- 3 1.5E-2 1E-6 2E- 5 8.6E- 1 5E-8 2E- 3 9.OE- 1 2E-10 2E- 3 3.2E- 3 2E-12 5E- 6 1.2E+0 1E- 8 1E-4 6.OE- 2 2E- 3 1E-7 8.5E- 1 2E-12 3E- 5b 1.9E+ 0 4E-9 1E- 2 2.1E+ 0 2E-8 3E- 5b 6.1E- 1 1E-6 1E- 3 7.6E- 1 9E-8 3E- 5b 2.6E- 1 5E-8 3E- 5b 2.1E+O 2E-8 1E- 3 5.9E- 1 3E-7 9E- 3b 1.2E+ 0 2E-6" 3E- 3 2.4E+ 0
7E- 2f 7E- 8 1E-4 4E- 5f 1E- 6 4E- 4 2E- 7f 7E- 11 4E- 7 3E- 5 4E- 7 3E-9 8E- 5 5E-4 2E+ 0 4E- 8 4E- 6 1E- 1 7E- 5 7E- 5 1E+ 0 5E- 2 8E- 3 1E- 10 2E- 9 4E- 1 2E- 4 8E- 7 3E-6 8E- 4 4E- 4 7E- 7 3E- 6 7E- 3 8E- 3 3E- 2 2E- 2 2E- 4 3E-4 5E- 3'
Environ. Scl. Technol., Vol. 17, No. 7, 1983
c p , g/L
ce,
5E-3
9E-7
g/L
5E + Oe 8E- 5 3E- 3 5E- 2 1E- 5 1E- 2 8E- 1 1E-4 2E- 2 2E- 5' 3E- 3g 3E+ 1 9E- 7 3E- 7 5E- 3 2E- 2 2E+O 5E- 5 2E-llg 1E-4 2E-9 6E-7 2E- 7 1E+0 2E-4 3E- 7 8E-5 1E+ 2 5E+ 0
4E- 8f 4E+ 4 7E-5 6E+lf 2E- 3 1E- 2 1E+4f 3E+ 0 1E+ 1 1E- 1 2E- 2 2E+O 2E+O 2E+O 6E- 4 1E- 1 2E- 3 3E- 5 7E-2 8E- 2 3E- 6 2E- 3 2E- 5g 3E- 3g 2E- 5 2E-8 3E- 5 1E- 2 3E- 3 7E- 6 2E- 5 1E- 2 2E-3 1E-6 9E- 7 3E- 7 5E- 7 6E- 5 2E- 5 2E+O
5E-4 5E-4
2E- 3
4E- 5 1E-4 4E- 5 3E-4
1E-4
9E- 7
3E-3 3E- 3 5E-4 5E-4
2E-3
1E-4 4E- 5 4E- 5 3E-4
1E-4
Table I11 (Continued) mother(n,x)daughter 209Bi( n,2n)208Bi
T,years 3.7Et 5
barns 2.5E-t 0
o,
ALI,MPC, Ci/a,Ci/m3 3E- 5 b
q , Ci/cm3
3E-4C 3E-10
BHP, m3/cm3 1E+ 2c 1E-4
cp, g/L 9E- 5
Ce,
g/L
Value valid only for neutron multiplier. Value valid only for neutron Value is MPC. a 1.6E+6 represents 1.6 X lo6. moderator. e No maximum concentration ce given in ref 12. f Uppermost value valid for beryllium/carbon from moderator, middle value valid for beryllium/carbon from multiplier, and lowest value valid for beryllium/carbon from steel; M ' = M w ~ / w , where wd = weight (grams) of daughter element in 1 g of blanket material. 10-4(MPC)/(N~~F/M because '), activity = 99 (145)g. per gram is N A a F / M ' , where M ' = weight of 1mol of Tc (Pm) as generated :i %:blanket
Since the radiotoxicity of the various radionuclides does not show in the activity, a quantity called the biological hazard potential (BHP; unit, m3/cm3) is calculated. The metabolism of radionuclides in the so-called reference man has been formulated in models on the basis of which the maximum permissible concentrations of radionuclides in drinking water (MPC) have been calculated (9-11). The biological hazard potential (BHP) is the upper limit to the volume of drinking water (having a concentration of O.l(MPC)) that could be contaminated by the inventory q. According to the definition of MPC (9)the consumption of 0.8 m3 of such water will cause a radioactive burden ("dose") of 0.5 rem, which is supposed to be independent from the distribution of the water: if one person drinks the whole 0.8m3, he receives the whole dose; if he shares some of the water with another person, his radiation burden is less and the other person gets the rest, the sum of the burdens again being 0.5 rem. So the BHP is also a measure of the maximum dose the inventory q of a radionuclide can cause: BHP = q/(O.l(MPC))
(3)
where MPC = ALI/0.8 if ALI is given in ref 10 or 11for radionuclides specified, if not, MPC value is taken from ref 9,023= annual amount of drinking water of man (m3), and ALI = annual limit for intakes of workers (Ci/year). Both quantities, q and BHP, give a measure of the amount of activity generated in 1cm3of blanket material such as the first wall. The total amount of first-wall material is fixed, again, by the technological limit to the 14MeV neutron flux (several MW/m2 (see above)) and the optimum power output of a reactor from the point of view of economics (500 MW or more), the thickness of the wall being bound by stability conditions to vary only little about the value of 5 mm. So the first wall of a fusion reactor will have a volume of several cubic meters. The thickness of the neutron multiplier, and thus also its volume, is about a factor of 10 higher (see Figure 3). Accordingly the total activity of long-lived waste and total potential amount of contaminated drinking water (level O.l(MPC)) is about 6 7 orders of magnitude higher than the specific values q and the BHP given in Table 111. As an example the volume of drinking water that can potentially be contaminated by the inventory of 63Nia t shutdown of the reactor is about lo8 m3. The potential amount of contaminated drinking water BHP decreases exponentially with time after shutdown of the reactor according to the radioactive decay law: BHP(t) = BHP exp(-At) (4) where BHP is the value a t shutdown of the reactor given in Table I11 and t is the time after shutdown. To find out which radionuclide dominates the hazard at a given time, the straightforward procedure is to plot all BHP(t) curves in a double-logarithmic coordinate system. The disadvantage of such plots is that they are
Figure 4. Blologlcal hazard potential of blanket materials.
optically overloaded by the large number of curves. Those curves all have the same shape and differ only by the lengths of their horizontal branches. Therefore in Figure 4 the curves are left out, except one for @Ni,which is drawn to give an example. Instead of the curves the shutdown values BHP are plotted vs. their half-lives T: the position of the chemical symbol for the radionuclide (only the element is specified) has the coordinates x = T, y = BHP. It is easy to reconstruct the plot with the decay curves, if necessary. But obviously one can already find the major radionuclides without explicitly drawing all curves, because they stick out from the mass of the others. The sum curve BHP(t) looks very much like the canvas of a tent being fixed at the prominent radionuclides as is the canvas at the tentpoles. The chemical symbols in Figure 4 have been underlined whenever the corresponding activity has been generated by 14-MeV neutrons, to distinguish the effect of the two components of the neutron flux. A degree sign attached to some symbols indicates that the-less reliable-old MPC values (9)were used to calculate the potential hazard. Furthermore, the value 3 X Ci/m3 assumed for the long-lived bismuth and beryllium is not based on their specific behavior in the human metabolism. It had to be assigned to those radionuclides according to the general prescription given by the ICRP, stating that for nuclides for which no MPC values have yet been specified in ref 9, the value 3 X 10" Ci/m3-given for an "unidentified" radionuclide-should be used. It can be expected that the ALI values that will be published in the future will result in higher MPC values for those nuclides. The asterisk indicates that the corresponding hazard potential rests with either the neutron multiplier (Be, Bi, Pb) or the moderator (C). The nuclides in fusion reactor waste having the potential to contaminate the largest amounts of drinking water are the long-lived isotopes of nickel, niobium, technetium, lead, Environ. Scl. Technol., Vol. 17, No. 7, 1983 429
T (second column of Table 111). To simplify the diagram, the distinction between the activation reactions (thermal or fast) was given up and only the smaller of the c values was used for the plot. The highest concentrations oLserved in the environment (12) are surrounded by a square in Figure 5. As one can see in Figure 5, the highest observed concentrations for the metals range between about and lo* g/L, whereas the permissible levels lie below lo4 g/L for the short-lived waste metals and increase with increasing half-life of the radioisotope because of the decreasing radiotoxicity of the long-lived radionuclides.
~\wpiNp1fJ
10-8
Eu
10-9
(Am1
10-10
10-11
io
lo2 10’
IO‘ lo5 lo6 lo7 to8 lo9 Half Life IYr)
Flgure 5. Permissible concentrations of activated elements in drinking water compared with maximum observed values.
and the ones (beryllium and bismuth) for which newly calculated radiotoxicities (ALI or MPC) are still missing, as one can see in Figure 4. For comparison Figure 4 includes the potential hazards due to the actinides americium, plutonium, and neptunium, which are generated in today’s fission-power reactor fuel. The volumes of highlevel fission and fusion reactor waste that have to be disposed of after the generation of the same amount of thermal energy in both types of reactors are about the same, roughly 1 cm3 of solid material/(kW year) if the comparison is limited to the volume of the fuel elements on the one hand and that of the first wall on the other. This is one of the reasons for relating the activities q and potential hazards BHP to 1 cm3. Having in mind the volume of the neutron multiplier in a fusion reactor (about 10 m3 in INTOR), one can take from Figure 4 that the total inventory of lead has the potential to contaminate about lo6 m3 of drinking water for periods on the order of lo7 years. The activated lead of the multiplier is a mixture of radioactive and stable lead isotopes, the ratio at shutdown being radioactive stable
-- Nda4Atfl M’
Ci/g of P b
(5)
where M‘is the weight of 1mol of multiplier lead (207.2 g/mol). The highest observed concentrations c, of lead in groundwater is roughly 100 Fg/L (12) (see last column in Table 111). Therefore the highest amount of activity that might be found in a liter of groundwater could be times that ratio (eq 5 ) . This value should be compared with the upper limit recommended by the ICRP for the contamination of drinking water, i.e., O.l(MPC). To facilitate the comparison, that limit being given in the unit Curies per cubic meter of water is converted into a limit cp (unit, gram of activated material per liter of water): cp = 10-4(MPC)/ratio of eq 5
(6)
where the factor 10“ appears because the limit is O.l(MPC) and lob3m3 = 1 L. In Table I11 cp has been calculated for all waste elements (M’ = M unless otherwise stated in footnotes to the table). In Figure 5 the maximum permissible concentration cp of each of the activated elements is plotted vs. the half-life 430
Environ. Sci. Technol., Vol. 17, No. 7, 1983
Conclusions The usual classification of high-level fusion reactor waste inventories according to the potential volume BHP of drinking water it can contaminate or the potential radiation burden of a population into the drinking water supply of which the inventories have penetrated shows that the potential hazard is large. As soon as more reliable radiotoxicities have been determined for beryllium and bismuth, the BHP will probably not be practically constant for millions of years but might turn out to decrease by 4 orders of magnitude over 1 X lo7 years. The medium-term hazard (up to lo3 years) is due to activated nickel, and the largest long-term hazard (up to lo7 years) is due to the neutron multipliers, unless the designers employ the refractory metals niobium and zirconium for the first wall. While characterizing the amount of activity and its effect on man (once the entire inventory has entered a drinking water supply), the hazard potential completely disregards the behavior of the waste nuclides in the environment. One such environmental property is the highest observed concentration c, in water. In Figure 5 one can see that for some metals the potential concentrations c, are orders of magnitude higher than the permissible ones c . An example of such an unwanted activateJ metal is nickel containing the radioisotope 63Ni(T= 100 year). Its permissible concentration cp is a factor of 50 lower than the maximum concentration c, found in environmental water. If the repository keeps it from entering the groundwater for a period of at least 6 half-lives (600 years), the activity of 63Niwill have decayed sufficiently, because the permissible concentration cp(ti)at the time, ti = 600 years, is a factor of exp(600 A) higher as the ratio (eq 5) of radioactive to stable nickel has decreased exponentially: cp(ti)= cp exp(Ati) = c,
(7)
where cp is the value calculated according to eq 6 and ti will henceforth be called “necessary isolation time”. Analogously as in Figure 4 one could now replace the shutdown values cp by the exponential curves cp(t) in Figure 5, in order to find the dominating elements. But for the same reason as before only the curve for a nickel isotope is included. The activated heavy metals that might appear in too high concentrations in drinking water are cadmium, nickel, and silver, but the isolation time ti is determined by silver alone (ti N 700 years), as long as all activated elements remain together in the groundwater (Le., in this paper the possible selective immobilization of elements by soil particles is not considered). An investigation of Figure 5 shows that even major uncertainties (order of magnitude) of the activation cross sections or the highest observed concentrations will have little effect on the isolation time, if they refer to elements to the left of silver. On the other hand, a drastic prolongation of the necessary isolation period will result if 10 pg/L is not the upper concentration limit for the still little
Environ. Scl. Technol. 1903, 17, 43 1-435
investigated elements to the right of silver. For example, a minor increase of either the activation cross section or the highest observed concentration of molybdenum will extend the isolation time drastically, from 700 years up to more than the half-life of B3Mo(3.5 X 103 years). Among those elements molybdenum, niobium, and technetium are the more hazardous ones, as one can see in Figure 4. Using the data presented in Figures 4 and 5 in the way described, it is possible to determine a mimimum isolation time necessary (here 700 years) if one does not want to rely on possible dilution or immobilization of radionuclides in aquifers or if the release of the radioactive inventory to the groundwater might excessively affect a population, i.e., the potential biological hazard is too large. Registry No. lmSm, 14907-33-6; lS2Sm,14280-32-1; lslSm, 1571594-3; lSIEu,14378-48-4;lSEu, 13982-02-0; ll1Hf, 14093-09-5; lglIr, 13967-66-3; '%, 14694-69-0; '%, 13967-67-4;W1,13981-72-1; "Ca, 14092-94-5; 68Ni,13981-79-8; 62Ni,13981-81-2;lase, 1483316-0; 92Zr,14392-15-5; 93Nb,7440-03-1; 92Mo,14191-67-4; 9sM0, 14392-20-2;g 3 R ~15128-32-2; , 'OSPd, 14914-59-1;'07Ag, 14378-37-1; l W d , 14336-65-3; 120Sn,14119-17-6; ' q e , 14390-75-1; la4Xe, 15751-43-6; 13%a, 1506585-7; l%m, 14981-82-9; 16?Er,15840-05-8; l@Ho, 7440-60-0; lSRe, 14391-28-7; lg2Pt,14913-85-0; 204Pb, 13966-26-2;%i, 7440-69-9; 14762-74-4; 14N,17778-88-0; 'IO, 13968-48-4; nAl, 7429-90-5; %K, 14092-91-2;4%a, 14333-052;64Fe, 13982-24-6; @Ni, 13981-80-1; %Ni, 14378-31-5; 94Zr,14119-12-1; 94Mo, 14683-00-2; lWMo, 14392-21-3; 9sRu, 18393-13-0; 9 9 R ~ , 15411-62-8;l14Cd, 14041-58-8; '%n, 14119-18-7; ' q e , 14390-72-8; 13"Te, 14390-76-2; lalBa, 13981-97-0; l%Ba, 15193-77-8; 138La, 15816-87-2;15?lb ' , 51691-20-4; laEr, 14900-10-8;'%Pt,14998-96-0; 206Pb,13966-27-3; "'Eu, 14683-23-9; lWEu, 15840-16-1; "'Eu, 15585-10-1; l'ISHf,14265-77-1; lMIr,14158-35-1;36Cl,13981-43-6; 41Ca,14092-95-6; 59Ni,14336-70-0; 63Ni,13981-37-8;79Se,1575845-9; 'O7Pd, 17637-99-9;lO8Ag,14391-65-2; l13Cd, 14336-66-4;lZ1Sn, 14683-06-8; lBI, 15046-84-1;lacs, 15726-30-4; 133Ba,13981-41-4; 93Zr, 15751-77-6; 94Nb, 14681-63-1; 93Mo, 14119-13-2; g9Tc, 14133-76-7; %Tc,15759-35-0; '&Pm, 15706-44-2; laHo, 14391-21-0; lBBHo,13967-65-2; lsBRe,14998-63-1; Isapt, 15735-70-3; 20sPb, 14119-28-9; 210Bi,14331-79-4; 'OBe, 14390-89-7; 14C,14762-75-5; 26Al, 14682-66-7; 63Mn, 14999-33-8; 91Nb, 14682-97-4; 92Nb,
13982-37-1; l3ICs,10045-97-3; l3ILa, 14834-69-6;l q b , 15759-55-4; neutron, 12586-31-1.
Literature Cited (1) O'Banion, K. Enuiron. Sci. Technol. 1981,15, 1130-1136. (2) "INTOR-International Tokamak Reactor, Phase 1, Con-
(3) (4)
(5)
(6) (7)
lac,
(8)
ceptual Design", EUR FU BRU/XII 2/81/EDV-50, IAEA-VIENNA;Commission of the European Communities, Director General XII-Fusion Programme: Brussels, Belgium, 1981; Vol. 111. Gruber, J. "Evaluation of the Activity Levels in Fusion Reactor Blankets"; HMLB202; Hahn-Meitner-Institut fur Kernforschung: Berlin, Germany, 1977. Gruber, J.; Schneider, J. "Transmutation und Aktivierung von Stainless Steel 31655 in einem thermischen Fusionsreaktor-Blanket"; HMLB212; Hahn-Meitner-Institut fur Kernforschung: Berlin, Germany, 1977. Gruber, J.; Schneider, J.; Lehmann, B. "NUCCON-A Program for Calculating the Neutron Activation"; HMIB312; Hahn-Meitner-Institut fur Kernforschung: Berlin, Germany, 1979. Gruber, J.; Schneider, J. 8th Symposium on Engineering Problems of Fusion Research San Francisco, CA, Nov 1979; IEEE Publication 79 CH 1441-5 NPS. Seelmann-Eggebert, W.; et al. "Nuklidkarte" Verlag Gersbach und Sohn: Munchen, Germany, 1974. Alley, W. E.; Lessler, E. M. Nucl. Data Tables 1973, 11,
621-825. (9) ICRP-Publication 2, "Radiation Protection Recommenda-
tions of the International Commission on Radiological Protection"; Pergamon Press: Oxford, Great Britain, 1959. (10) Adams, N.; et al. "Annual Limits of Intake of Radionuclides for Workers"; NRPB-R82 Harwell; Didcot: Oxon, Great Britain, 1978. (11) ICRP-Publication 30, "Limits for Intakes of Radionuclides by Workers"; Pergamon Press: Oxford, Great Britain, 1978; parts 1 and 2. (12) "Drinking Water and Health"; National Academy of Sciences: Washington, D.C., 1977. Received for review April 23,1982. Accepted February 22,1983.
Chemical Analysis of Acid Precipitation: pH and Acidity Determinations Neil R. McQuaker," Paul D. Kluckner, and Douglas K. Sandberg
Environmental Laboratory, Ministry of Environment, Vancouver, British Columbia V6S 2L2, Canada The effects of the residual streaming potential and residual junction potential on the accuracy of pH and acidity measurements are discussed and quantified. Detailed analytical procedures minimizing these effects are described; a Gran's titration is used to assess total and strong acidity. The procedure for pH determination is shown to have a precision and accuracy of fO.O1 pH unit. The acidity procedure shows acceptable accuracy and mean precision values (expressed as relative standard deviations) of 1.4% and 3.4% are obtained for strong and total acidity on the intervals 24-97 and 34-110 pequiv of H+/L.
Introduction Acidification of precipitation due to anthropogenic input has been recognized as perhaps one of the most severe environmental problems facing mankind. It is a problem that is often transboundary in nature, and with the advent of network monitoring on a national and international scale the standardization of the chemical analyses of precipi0013-936X/83/0917-0431$01.50/0
tation has assumed increasing importance. The chemical analyses of precipitation for the principal ions (excluding hydrogen), which are usually considered to be Na, K, Ca, Mg, NH4, C1, NO3, and SO4 ( I ) , may be accomplished by using conventional techniques. That this is so is supported by the results of the latest World Meteorological Association (WMO) interlaboratory study (2). In this study the participating laboratories were requested to use the analytical methods normally employed. The majority of the laboratories achieved acceptable results, and the mean values obtained were very close to the expected values. The favorable situation that exists for the determination of the principle ions unfortunately does not exist for the two parameters of primary importance in assessing the environmental impact of acid precipitation, i.e., pH and acidity. It has recently been recognized that, for both of these parameters, there are severe measurement problems and a need for standardization exists (1, 3, 4). This is reflected by the results of WMO interlaboratory studies (2,5) appearing in Tables I and 11. The mean values obtained, particularly for acidity, show considerable de-
0 1983 American Chemical Society
Environ. Sci. Technol., Vol. 17, No. 7, 1983 431
