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
DM-123
- 2 -
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.
- 3 -
DM-123
- 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
- 4 -
DM-123
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
- 5 -
DM-123
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.
DM-123
- 6 -
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
- 7 -
DM-123
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
DM-123
- 8 -
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
- 9-
DM-123
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.
DM-123
- 10 -
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
- 11 -
DM-123
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.
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