Atomic Energy of Canada Limited

RADIOACTIVE WASTE MANAGEMENT IN THE LONG TERM DM-123

by

W. BENNETT LEWIS

Chalk River Nuclear Laboratories Chalk River, Ontario October 1972 AECL-4268

DM-123

RADIOACTIVE WASTE MANAGEMENT IN THE LONG TERM

by W. Bennett Lewis

Chalk River, Ontario October, 1972

AECL-4268

DM-123

RADIOACTIVE WASTE MANAGEMENT IN THE LONG TERM

by W. Bennett Lewis

ABSTRACT

A future is envisaged in which a world of 15,000 raillion people is supplied with energy from nuclear fission at an average of 50 thermal kilowatts per capita. The resulting radioactive wastes are managed permanently within the boundaries of plants that recover fuel for recycle and fabricate the new nuclear fuel. It is foreseen that a single plant would manage the fuel and wastes for 250 to 300 million kilowatts electric generating capacity. By the year 2000 about four such plants may be needed in North America. In the long-term future about 1,000 such plants would meet the envisaged world demand. An outline is sketched of the operations in such a plant on a near-breeding thorium-uranium fuel cycle. The operations are characterized by multiple parallel cycles for all materials and retrievable storage of radioactive wastes.

Chalk River, Ontario October, 1972 AECL-4268

Gestion des déchets radioactifs à long terme

par W. Bennett Lewis

Résumé On envisage un avenir ou une population mondiale de 15 milliards d'êtres humains disposera d'une énergie d'origine nucléaire en moyenne à raison de 50 kilowatts thermiques per capita. Les déchets radioactifs provenant des centrales nucléaires seront gérés au sein des usines qui retraiteront le combustible irradié et qui fabriqueront les combustibles nucléaires neufs. On prévoit qu'une seule usine pourrait gérer le combustible et les déchets provenant de centrales nucléaires ayant une capacité totale de 250 a 300 millions de kilowatts électriques. En l'an 2000, il se pourrait que quatre usines de ce genre soient nécessaires en Amérique du Nord. Dans un avenir plus lointain, il en faudrait L000 pour répondre a la demande mondiale prévue. On donne un aperçu des activités qu'une telle usine aurait avec un cycle de combustible thorium-uranium quasi-surgénérateur. Les travaux seraient caractérisés par des cycles parallèles multiples pour tous les matériaux et pour le stockage récupérable des déchets radioactifs.

L'Energie Atomique du Canada, Limitée Laboratoires Nucléaires de Chalk River Chalk River, Ontario Septembre 1972 AECL 4268

AECL-4268 (DM-123)

RADIOACTIVE WASTE MANAGEMENT IN THE LONG TERM by W. Bennett Lewis

INTRODUCTION The second United Nations International Conference on the Peaceful Uses of Atomic Energy of 195S was not only the occasion of massive international exchange of technical atomic energy information published in the 33 volumes of its proceedings, it also stimulated the compilation and publication of the shared information in numerous specialized volumes. One most notable such work is "Atomic Energy Waste: Its Nature, Use and Disposal" edited by E. Glueckauf, Butterworths, 1961. The editor's introduction notes "In the first years of atomic energy, the problem of how to deal with the radioactive waste products used to be approached with a feeling of apprehension. This reaction is, of course, quite general with every new dangerous phenomenon and its intensity is usually inversely proportional to the knowledge and experience existing in the new field. However, as early as 1955 ... it was pointed out that the disposal of the fission products ... would require only a small fraction of the ingenuity that brought them into being, (later set at about 2% of the annual atomic power costs) . This has been borne out by developments ... The confidence with which engineers and scientists approach these developments is based on the vast experience which has been obtained in the field during the last years". That was written more than ten years ago before the nuclear power industry was distinguishable from the nuclear weapons operations. It may still be another ten years or more before the great simplifications of radioactive waste management for the nuclear power operations are commonly appreciated. ADVANCES IN THE LAST DECADE In the meantime information has greatly expanded. Most of it is favourable, such as the experience at Chalk River with high level wastes fused in glass blocks buried in the ground in a basin in which the ground water is monitored and available for treatment.'-^ i^' The outflow is controlled to drinking water levels of radioactivity. Some new information, however, is unfavourable, such as the recognition of 8 year half-life europium-154 produced by neutron capture from the stable fission product Eu-153, and the 13 year Eu-152 isomer

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from stable Eu-151 making europium one extra element needing longterm management. The effects of these are minor compared with what may be called the reversal of the viewing direction, explained below, on high activity waste management and the increasingly favourable economies of delayed processing. Moreover, the whole system becomes controllable and finite by the introduction of recycle in multiple parallel cycles. It becomes economic by assigning appropriate levels of decontamination to each recycle stream. Some streams pass back through nuclear reactors, others are confined within the fuel processing and waste management site and for these the decontamination factors may be low. The impact of the reversed viewing direction is reached by noting that in an expanding or steady system the wastes of previous years have a lower activity than this year's wastes because of radioactive decay. Storing or otherwise managing earlier wastes amounts to only a fractional increase on the accumulated wastes of the last ten years, except in the total mass. It becomes in fact economic to plan to limit the total mass by the multiple recycling. Only one major cycle need be as long as 1000 to 2000 years and this does not seem impractical. In brief summary, it is proposed that: - Spent fuel is received regularly at a combined fuel processing, fuel fabrication and waste management engineered site, with all effluents controlled, and stored materials retrievable. - Fabricated fuel and radioactive shipments from the site are used in controlled cycles. - The effluents to be controlled may include solutions and suspensions in water, tritiated water, gases, aerosols, windblown dust and insect carried materials. - The received spent fuel is retained in its cladding and merely cooled adequately for an initial period of perhaps a year, or more if its inventory value is low. Then it is processed to recover fissile material for further use. - At this stage or some years later fertile material is adequately purified for later recycle as nuclear fuel and stored until the short-lived radioactive components have decayed, when it may again be incorporated in fresh fuel. - Fission products and other radioactive residues from fuel are kept in solution in specially cooled tanks until ten years out from the power reactor.

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- At this ten year time only the 15 to 20% of fission products (typically less than 0.7% of the spent fuel mass) which need continued special cooling are processed into solid (e.g. glass) blocks. These fission products may form 5% of the glass mass. - Other fission products, notably zirconium isotopes, may be recycled or stored in fully concentrated solid forms. - There is no need to decontaminate any materials to a high degree unless that is required for some special purpose. - Those materials such as fertile components may retain higher isotopes and other heavy elements when recycled. By taking care in advance any excess radiations from the fertile materials kept for recycle may be held to acceptable levels. - The total mass not recycled simply remains stored indefinitely on the site of the operating plant in the form of dense oxides of low solubility in the controlled water to which they may become exposed. LONG-TERM NEEDS The operations may be assessed quantitatively from considering the accumulated wastes from 2000 years of operation of enough nuclear power to satisfy all the world's needs. For human comfort it may be supposed that the world population has limited itself to about 15,000 million supplied with energy at 50 thermal kW per capita that is applied to the production of food, fresh water, clean air warmed or cooled as desired, fuel for mobile services and locomotion, ore reduction and chemical processing, maintaining waterways and certain roads free of ice, local climatic control, etc. This amounts to about four times the present world population utilizing 150 times the current energy flow not taken directly from the sun as water power or otherwise. All this extra energy is envisaged as derived from nuclear fission and amounts to 0.5% of that received by the earth from the sun, not enough to induce major new jlimatic problems. The suggested total, fission power of 750 terawatts (or 750,000 MkW) produces 7-5 x 108g of fission products/day (since 1 MWd-+lg F.P.). If divided equally among 1000 processing plants each receives 750 kg/ day. (Note: each plant serves 250 to 300 million kilowatts of electric generating capacity.) Assuming the fission products are contained in fuel at an average burn-up of 34 MWd/kg of heavy element (H.E. = U, Th, Pu, etc.) the spent fuel delivery to each plant would be 22 tonnes H.E./day or perhaps 26 tonnes total nuclear fuel per day. This is judged sufficient to satisfy the economy of scale for a primary fuel reprocessing plant. By the year 2000 A.D. current

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estimates forecast 800-1000 MkWe of nuclear power in North America, requiring 3 or 4 such plants and for the whole world 8 or 9 plants. Further expansion would gradually shift to be relatively great in Asia, South America and other populous but now less developed areas. Prom what follows it will be seen that only about 17% of the fission products remaining active after 10 years of cooling need continued special cooling and long-term (2,000 year) storage. If these are fused in glass and form 5% of the total mass, there would be about 3.4 kg glass product per kg of total fission products. Since the total F.P. in the full-scale plant is 750 kg/day, the glass product would be about 2-6 tonne/day or 1,900,000 tonnes in 2,000 years, or about 10 6 m 3 or 2 metres depth ever 0*5 km2 which would seem easily manageable. This material is regarded as recoverable and would be processed for use again after 1,000 to 2,000 years when the residual activity is conveniently low. The object of reprocessing would be to diminish the amount of new mineral required for absorbing the continuing feed of highly active fission products. If the plant is operating on the thorium fuel cycle after 1,000 years the residual activity in the glass blocks would be mainly from Sm-151, Cs-135 and 1-129. This may be seen from Table I in conjunction with Table II. TABLE I HalfLife

36 Sr-90

39 Y-90

q at 10^ g at lOOOy kg total F .P.

28-9 1 • 010 (28-9) ' 20

initial g

g at lOlOy kg total F.P.

3•835xlO~ 11 7 •67xlO~10

W/g

0.919

W at lOlOy kg total F.P.

7 •05xl0- 10

531-129

l-6xlO7

7 .899

0.999957

7 .8987

O.lOxlO"6

7 .90xl0~ 7

ssCs-135

2xlO6

4 .325

0.999653

4 .3235

0.62xl0~G

2 .68xlO~6

55 Cs-137

30.2

31 .563

62 Sm-151

93

0 .0808

1 .077xl0- 10 3•40xl0~9 5.796X10-1*

4 .683xlO~5

0.421

1.43xlO~3

0.00392

1 .84xlO~7

Moreover, it is of interest to evaluate the heat to be dissipated from the glass when it is first produced. As derived in Table II it is about 0.2 watts/g of the F.P. residue incorporated or 10 watts/kg of glass product and a total for a day's output

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of 2.6 tonnes of 26 kW. The heat from any given glass block decays with an effective half-life of about 28.9 years, which is that of the main contributor Sr-90 + Y-90. However, since fresh blocks are constantly added, the total heat output rises to an eventual equilibrium where the daily decay = daily addition. The equilibrium total from the bed derived from 750 thermal Gigawatts is thus 26/X kW where A = decay constant = 6.57 x 10~ 5 day"1 making the equilibrium heat 396 MW. To put this on scale it may be noted that the heat flux from the sun that is balanced by radiation, evaporation and convective cooling is -300 MW/km2 at latitude 45°. For a 0.5 km2 bed of glass blocks giving 400 MW the heat flux is 0.8 kW/m2. Convective cooling is given approximately by AT = /^

51

I

n

x 101* d e g . C ,

where AT is the excess temperature

of a horizontal flat surface exposed to air and H is the heat flux in kW/m2. Such convective cooling is too low to be directly useful. Evaporative cooling from a horizontal surface varies with roughness, wind, etc., but for a typical crop growing in a temperate climate is about 2.5 mm depth of water per day. So that at 2420 J/g for evaporation the rate of heat removal is about 0.07 kW/m2 which again is too small without stimulated flow. In addition, however, to the 400 MW from the glass blocks, the plant has also to dissipate the heat from all the fission products stored or in process during the first ten years from receipt, which at 520 kW/day for 3,652 days amounts to 1,900 MW assuming receipt about three months after discharge. Such amounts of waste heat are commonly dissipated to the atmosphere via a river, cooling-pond or lagoon or evaporative cooling towers. Any appropriate method may be adopted. Economics As indicated above a fuel reprocessing and waste management site would be expected to accept the spent fuel from 300 MkWe generating capacity. Then at 8,000 hr/y a charge of 0.05 mill/kWh contributes to the operating budget $300 * 10 6 x 8000 x 5 x 10" 5 = $120 million per year. Probably this suggests an excessively large operation but it may not be unreasonable to set the charge at 0.05 mill/kWh because the site may first be established for 6 MkWe generating capacity when the contribution would be $2.4 million/year. For initial operation at 6 MkWe on the CANDU natural uranium cycle with 8 MWd/kg Nat U burn-up valuing the recovered plutonium at $9/g fissile Pu would yield on the same duty cycle at 2.7g fissile Pu/kg Nat U.$20.25 million/year and the recovery cost at $15/kg Nat U would be $12.5 million/year.

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For the same 6 MkWe scale of operation on a CANDU-OC + thorium c y c l e ^ (and using approximate numbers to keep the arithmetic simple) at 35 MWd/kg H.E. (H.E. = heavy elements Th + U etc.) containing 16g U-233/kg H.E. valuing recovered U-233 at $15/g U-233 yields on the same duty cycle $34.3 million/year. If U-233 is recycled a net supply of U-235 of about 0.16g U-235/MWd is required. (See Fuel Sequence #12 Table 1 in refce (3)). At the old price of $ll/g U-235 its supply would cost $8.8 million/y or 0.185 mill/kWh in the total fuel cycle cost of 0.54 mill/kWh. Plant Operations The plan of operation at each plant would be to manage all wastes with significant activity for as long as necessary. Reviewing what is necessary, all fission products can be placed in four groups of elements: Group 1.

Those which after one year of decay may, if desired, be released because they have only stable nuclides or of such long life that they occur in nature Ge, A s , Br, Rb, Mo, Rh, In, Ba, La, Pr, Nd, Gd, Tb, Dy, and Xe when freed from Kr.

Group 2.

Those which may be released after a further 9y decay: Y, Ag, Te, Ce and the Mo, Nd and Gd which has grown in by decay of Zr, Nb, Ce and Eu.

Group 3.

Those which could be released after 2,000 years storage fused in glass blocks or the equivalent, Sr(+Y-90), Nb, Ru, Cd, Sb, Cs, Pm, Sm, Eu and Kr but note a possible restriction on Cs due to Cs-135.

Group 4.

Requiring indefinite retention or special management:

(a) (b) (c) (d)

Se (Se-79 - 6 x 10"y), Pd (Pd-107 - 7 x 1 0 6 y ) , Sn (Sn-126 - 2 x 10 5 y) Zr-93 - 1-1 x 10 6 y Tc-99 - 2-12 x 10 5 y 1-129 - 1-6 x 10 7 y

Group 3 is the only group needing special cooling after the 10 year period. It is accordingly envisaged that about one year after receipt of the spent fuel, when the intense activities of Xe-133; Ba-140, La-140; Zr-95, Nb-95 and Sr-89 have abated the spent fuel would be dissolved, processed for the recovery of its fissile component, the segregation of other heavy elements Th, U-238 etc. for later recycle, the trapping of Xe and Kr and control of tritium. i- Canada Deuterium Uranium Organic-Cooled Reactor

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The Xe and Kr would be separated, the Xe being shipped for use as stable and the Kr stored. The other wastes, less the fraction of those elements of Group 1, that may have been easily separated, would be stored in solution for nine years. After this nine years the solution would te chemically separated into as many groups as desired. It is suggested that all in Groups 1 and 2 would be set aside for storage or decontaminated and released. All in Group 3 with possibly Kr and 1-129 would be fused in glass blocks or equivalent and stored on the site for 2,000 years. This significant operation is discussed further below. For Group 4 the yield of elements in sub-group (a) is relatively small and they may have no special value, their heat output is low and they may be solidified and put in permanent storage. The yield of the element zirconium is relatively high and the activity low. It is suggested that it could be distinguished as fission product zirconium but reused as metal in nuclear reactors if kept segregated from natural zirconium. Its neutron absorption, though higher than for natural zirconium, is not large so it could be acceptable for some purposes. Until required it could be calcined and stored as oxide. The yield of technetium, Tc-99, is quite high, -6% per fission, but it is a unique nuclide not occurring in nature, it is expected that it would be chemically separated, and decontaminated as a useful material under circumstances where its radioactivity is acceptable. For iodine-129 there are several options, its half-life is so long, 1.6 x 10 7 years that it could be considered releasable. As the parent of a single stable isotope of xenon it could be segregated as a cow to be milked from time to time for its product. A third option is that it could be included with the elements to be stored for 2,000 years and would form only about 0.004% of the glass mass. A basic principle of operation throughout the plant and storage areas is that gaseous and liquid discharges are monitored and where necessary routed through trapping or recovery units for active nuclides. Material recovered from these traps or recovery units is put in a form suitable for injection at an appropriate point into a normal process stream or store. Treatment and 2,000 year storage of Group 3 Elements The glass block storage system so far tested at Chalk River ^ ^ (^requires development at several points. Basically some compromise was found necessary between a high melting point glass of

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good water resistance and a low melting point glass that minimized recycling of Cs and Ru needed because of their volatility. Having shown that in principle such a method can serve, it and alternatives should be re-explored to select a process of the greatest convenience. Essentially what is required is a water-resistant solid composition. The units may be small like marbles or as large as cannonballs, and may be homogeneous or layered in their internal structure. The storage bed would be divided into plots to permit monitoring. It would be possible, if desired, to use separate areas for say 500 year periods. This would facilitate quarrying for recycle when any area has become low enough in activity. One day's input into store may be spread over the whole of one such plot, or concentrated, depending on the method of cooling that is chosen. The heat output quoted in the discussion of the longterm needs is derived in Table II. The particular case evaluated is for U-233 fission considered constant in amount while irradiated in a Westcott neutron flux of 5 x 1 0 1 3 n/cm2/sec. to 4n/kb, i.e. for 80 Ms - 2.5 y. The calculation was made by the FISSPROD code as of September, 1972. Calculations of heat output were made by hand, using decay energies and half-lives from tabulations other than the FISSPROD library. It may be noted that iodine is included but krypton excluded from Table II. It seems quite possible that a means can be found for trapping krypton in the glass, taking into account the observed reabsorption of fission product gases at moderate temperatures in ceramic fuel under irradiation. The method for handling the krypton is, however, not yet selected. Recycle of Thorium Natural thorium is slightly radioactive and the radiation of most significance is the 2.6 MeV gamma ray from thallium-208 (ThC") the last active daughter in its decay chain. The half-life of thorium-232, 1.4 x 1 0 1 0 years, is so long that the level of its radioactive daughters is quite low. However, as a reactor fuel it is liable to be associated with uranium-232 of only 74 y half-life which is another direct parent of radiothorium, thorium-228 of 1.9 y half-life. Being an isotope of thorium, radiothorium is not chemically separable and the amount arising from uranium-232 may be many thousands of times that occurring in natural thorium. Consequently thorium chemically recovered from spent fuel can be very highly radioactive which is inconvenient for recycle. The radioactivity may, however, be kept low in a cycle of ten to twenty years. It is necessary to decontaminate the thorium from uranium-232. Chemically this may be achieved by repeated separations using uranium-238 (or natural uranium) as a carrier. Then, if stored, the

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TABLE II Group 3 - 2000 Year Storage Amount afteic 10 years Nuclide

38-Sr-86 -88 -90 (39-V-90) 44-Rv-100 -101 -102 -104 -106 (45-Rh-106) 48-Cd-110 -111 -112 -113m -113 -114 -116 51-Sb-121 -123 -125 (52-Te-125m) 53-1-127 -129 55-CS-133 -134 -135 -137 (56-Ba-137m) 61-Pm-147 62-Sm-147 -148 -149 -150 -151 -152 -154 63-EU-151 -152 -153 -154 -155 Total

Halflife years

28.9 (28.9) CD 00

1.008 (1.008]

13.6 00

OD

2.75 (2.75)

Atoms/fiss. x 10 3 0.023 54. 515 51. 470

0. 009 20. 722 20. 010

2. 752 31. 560 24. B21 10. 506 0.00125

1.189 13. 769 10. 980 4.720 0.00056

0.123 0.193 0.208 V. small 0.0018 0. 398 0. 178

0.058 0.093 0.101 0.000009 0.001 0. 196 0. 089

0.195 0 492 0 118

0.102 0 261 0 0638

55 203 2.06 0 113 2 « 10 6 7 416 30.2 53 334 (30.2) oa

CO

00

93

13.2 OQ

8.0 4.9

W/g

0.20 \x 1 13 0.93 I "

0.919

0.01 \ , 63 1. X

62 } -

5 854 16 » 10 6 14 1'5

2.62

gAg Total F.Ps

Decay* Energy MeV/Decay

3 211 7 8»9 31 0 4 31

714 0656 325 563

0 646

0 410

11 636 3 252 0 080 8 024 0 .124 4 .126 0 .3825

7 389 2 079 0 051 5 .199 0 0808 2 .709 0 .254

0 .0095 0 .00062 1 .887 0 .207 0 .041

0 .0062 0 .00041 1 .248 0 .1375 0 .0277 170 .73

Heat Output At 10 years W/g W/kg ResiF.Ps due

32.6

18.38

0.0183

0 186

0.258

0.00000242

0 533 \o 564 0 031 / °-

3.48

0.222

0 097

0.10 x 10-6

1 72 0 079 0 185 \o 823 0 638 )°0 072

13.22 0.62 x 10 -6 0.421

0.00000079 0.867 0.0000027 13.29

0.396

0.162

0 .0260

0.00392

0.000317

1 .238

1.310

0.000537

1 .505 0 .125

2.58 0.348

0.356 0.00966 33.305

0.1951

EpYEY + EfgPgEog where p ' s are emission p r o b a b i l i t i e s and !„ the average and EOg the maximum B-energy and f.(=i./Ei) e ) depends on the type of B-transition (allowed, 1st forbidden, e t c . ) . Decay data are taken from "Tables of Isotopes" by Lederer, Hollander & Perlman (6th edn.Wiley, 1967) or "Nuclear Data Tables" (ed. K. Way) where the data are available.

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radio-thorium will decay to a low level in 15 years. There is, however, one further possible complication that radium-228 (mesothorium-1) 6.7 y half-life that occurs in the thorium chain ahead of radiothorium should be kept down to not many times the level of its natural occurrence which in terms of mass ratio is extremely small, 6.7/(1.4 x 1 0 1 0 ) . From the radiation protection aspect the level of uranium-2 32 that can be left in thorium intended for recycle would have to be only 5 x 10" 9 of the thorium for its contribution of radio-thorium to be only equal to that occurring in natural thorium. It is possible to handle 100 kilograms of natural thorium without experiencing fields greater than 50 mR/h. In several geometries at normal working distances the field is about 0.4 yR/h per g of thorium or 40 mR/h per 105g. Recycle of Uranium There are several factors that seem likely to lead to the storage of uranium rather than recycle for a very long time, perhaps centuries. In fuel fabrication preference would go to natural uranium or the depleted uranium from isotope separation plants because of freedom from radiation and because of the special value of uranium-235, and low cost of depleted uranium. The freedom from radiation arises because of the hold-up in the uranium radioactive chain caused by the relatively long life of radium-226 in the main chain and of protactinium-231 in the odd mass or actinium chain. These nuclides are eliminated to a significant degree in the refinement of uranium. Uranium may be stored readily in massive form as UO 2 or U3Oe or mixtures. The time before recycling becomes economically competitive would be shortened in the event that the plutonium breeder reactors do not in fact check the rise in cost of natural uranium. Plutonium itself presents complex problems because of its numerous isotopes and the ingrowth of transuranic elements of higher mass and atomic numbers. No detailed study has been attempted along the lines of this report. Management of Auxiliary Wastes In the operation of normal chemical plants over a long term there is both renewal and replacement of equipment. Discarded equipment over the years amounts to a large total volume. The cost of decontaminating equipment to the degree desired to allow it to be moved from the site would be high so the alternative of re-using most of the material is likely to be adopted. The same applies to chemicals that cannot be reduced to non-radioactive effluents principally

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air, water and C0 2 . The plant will accordingly incorporate a growing section for recycling of equipment and chemicals, including water and liquid chemicals containing tritium. It is not possible to predict with any certainty the form these features will take, but from the beginning it is to be expected that considerations of ultimate disposal will influence the choice of equipment and chemicals to be used. Consequently the design of the plant will be unlike current conventional plants, but the design will pioneer features likely to become widely applied in other plants, especially as concern grows in the world to minimize the adverse effects of industrial operations on the environment.

Acknowledgements The novel features provoked many comments on drafts of this report. I hope that the presentation is now clearer and my thcnks are extended to all those who commented. Also I am especially indebted to W.H. Walker who updated the FISSPROD program to take account of revisions in the nuclear data from recent experiments, and to G. Cowper who reviewed the radiation exposures to be expected from handling thorium.

References (1)

L.C. Watson, R.W. Durham, W.E. Erlebach and H.K. Ra2 "The Disposal of Fission Products in Glass" P/195 Proceedings 2nd U.N. International Conference on the Peaceful Uses of Atomic Energy, Vol. 18, p. 19, 1958.

(2)

W.F. Merritt "Permanent Disposal by Burial of Highly Radioactive Wastes Incorporated into Glass" SM-93/29, International Atomic Energy Agency Symposium on Disposal of Radioactive Wastes into the Ground, pages 403-408, 1967.

(.3) W. Bennett Lewis, M.F. Duret, D.S. Craig, J.I. Veeder, A.S. Bain "Large-Scale Nuclear Energy from the Thorium Cycle" AECL-3980, Paper A/Conf.49/P/157 Proceedings of the Fourth International Conference on the Peaceful Uses of Atomic Energy, Geneva, Sept. 1971. Vol.9, pp. 239-253.

WBL/g

A d d i t i o n a l copies of this document may be o b t a i n e d from Scientific Document D i s t r i b u t i o n Office Atomic Energy of Canada L i m i t e d Chalk River, O n t a r i o , C a n a d a KOJ 1J0 Price - 50< per copy

2639-72