ORNL-TM-3548.txt

 

 

 

NOV 23 1971 __ {3 o g 2

_ ..-..dfaf M ik A < ey Fom

OAK RIDGE NATIONAL LABORATORY
operated by _
UNION CARBIDE CORPORATION * NUCLEAR DIVISION
for the

U.S. ATOMIC ENERGY COMMISSION \
Ve
ORNL- TM- 358"~ }

THE LONG-TERM HAZARD OF RADIOACTIVE WASTES PRODUCED BY
THE ENRICHED URANIUM, Pu=238U, AND 233y—_Th FUEL CYCLES

M. J. Bell and R, S, Dillon

NOTICE This document contains information of a preliminary nature
and was prepared primorily for internal use at the Ock Ridge Notional
Loboratery. It is subject to revision or correction and therefore does
not represent a final report.
 

T S — m———— - . —— R e — .. e e — el - e —

 

 

This report was prepared as an account of work sponsored by the United
States Gowvernment. Neither the United States nor the United States Atomic
Energy Commission, nor any of their employees, nor any of their contractors,
subcontractors, or their employees, makes any warranty, express or implied, or
assumes any legal liability or responsibility for the accuracy, completeness or
usefulness of any information, apparatus, product or process disclosed, or
represents that its use would not infringe privately owned rights.

 

 

 
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A & R
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OAK RIDGE ‘NATIONAL I.ABORATORY
operated by
UNION CARBIDE CORPORATION  NUCLEAR DIVISION
for the

U.S. ATOMIC ENERGY COMMISSION
ORNL- TM-

 

' THE LONG-TERM HAZARD OF RADIOACTIVE WASTES PRODUCED BY
THE ENRICHED URANIUM, Pu=238U, AND 233y-Th FUEL CYCLES

M. J. Bell and R, S, Dillon

NOTICE This document contains information of a preliminary nature
and was prepared primarily for internal use ot the Oak Ridge National
Laboratory. It is subject to revision or correction and therefore does

not represent a final report.

 

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ORNL-TM-3548
Contract No. W-T4OS5-eng-26

CHEMICAL TECHNOLOGY DIVISION

THE LONG-TERM HAZARD OF RADIOACTIVE WASTES PRODUCED RBY
THE ENRICHED URANIUM, Pu-238(, AND 233U-Th FUEL CYCLES

M. J. Bell and R. S. Dillon™

NOVEMBER 1971

 

*Present address: AECOP, Oak Ridge, Tennessee.

OAK RIDGE NATIONAL .LABORATORY
Oak Ridge, Tennessee 37830
operated by
UNION CARBIDE CORPORATION f
for the
U.S. ATOMIC ENERGY COMMISSION
iii

CONTENTS

Abstract . . . ¢ f e e e e e e e e e e e e e e
1. Introduction . . . . . . « « + 4 . .

2. Reactor Operating Conditions and High-Level-
Waste Radionuclide Inventories . . . . . . . .

3. Hazard as a PFunction of Age of High-Ievel Waste

4. Long-Term Relative Hazards of High-Level Waste .

5. Isotopic Dilution of High-Level Wastes . . . .
6. Relative Hazard of Transuranium Wastes . . . .
7. References . . . .

ApPendiX « v . v w e e e e e e e e e e e e e e e

11
1k
16
18

19
 

v THE LONG-TERM HAZARD OF RADIOACTIVE WASTES PRODUCED BY
’ THE ENRICHED URANIUM, Pu-<3%U, AND <=33U-Th FUEL CYCLES

 

 

M. J. Bell and R. S. Dillon

ABSTRACT

An evaluation has been made of the long-term hazards of
~ "high-level"” and "alpha" radioactive wastes generated by irra-
diation of three fuels representative of those which will be
used to generate electrical powér in the next several decades.
In this evaluation the composition and radioactivity of the
wastes generated by typical LWR, IMFBR, and molten-salt breeder
“reactors utilizing the enriched uranium, Pu-228U, and 233U-Th
fuel cycles,respectively, have been computed for times up to
30 million years after discharge. |

. The volumesof water necessary to dilute the various types of
high-level waste to the radiation concentration guides (RCG) for

- ingestion in unrestricted areas have been computed as a function
of age of the waste. It was found that the volumes required to
dilute the three types of waste were similar when evaluated at
the same age and exposure. The volume of water necessary to
dilute any of the three wastes aged 1000 years and the associated
salt in the proposed Federal Repository to the RCG for unrestricted
uses is less than that required to dilute the ‘same amount of uranium
ore tailings to the RCG. The high-level waste deposited in the
Federal Repository will result in an alpha activity of less than
10 uCi/kg-at 10,000 years after burial when averaged over the total
mass of the Repository.

. The hazard of a number of types of waste contaminated to
10 uCi/kg (the upper limit assumed for surface burial) of initial
parent alpha activity was also calculated as a function of time, -
and the time of maximum relative hazard was determined, Trans-
uranium or alpha wastes contaminated to -this level will present a
maximum ingestion hazard similar to naturally occurring uranium
ores.

 

1. INTRODUCTION

‘High-level and alpha wastes generated in nuclear fuel cycles for .
production of electrical power will contain.isotopes which remain radio-
active for miiiions,of years. The "high-level" wastes are principally
the fission-pré@uét congentrates that arise from the recovery of fissile

and fertile materials from.spent fuel. Typically, however, these wastes
contain a variety of actinides that are made from transmutation of fuel
material and, in addition, quantities of uranium, thorium, and plutonium
that are not economically recoverable for recycle., "Alpha'" wastes —
materials contaminated with substantial concentrations of long-lived
alpha emitters — are produced primarily in plants for preparation of

nuclear fuel materials.

Sighificant variations will occur in the compositions of the wastes
generated by various reactor concepts because of differences between types
of fuel, neutron energy spectra and flux level, and length of irradiation,
as well as'efficiéncy of utilization of fuel. Calculations have been made,
as a function of decay time, of the compositions and radiocactivities of
the wastes generated by typical light water reactors {(LWRs), liquid
metal cooled fast breeder reactors (IMFBRs), and molten-salt breeder reactors
(MSBRs), which are representative of enriched 237U, 229py, and 233U fuels,
These radicactivities have been used,along with the radiation concentration
guides for ingestion in unrestricted areas given in the Code of Federal
Regulations (10 CFR 20), to evaluate the relative hazard of the wastes.
 The compositions of 460 fission products and 80 actinide elements and
their decay products were included in the calculations, which were per-

formed with the ORIGEN isotope generation and decay code.l’2

The authors wish to acknowledge the assistance of H. C. Claiborne,
J. 0. Blomeke, and J, P. Nichols in reviewing this report and for their

helpful suggestions in the course of the work,

2. REACTOR OPERATING CONDITIONS AND HIGH-LEVEL-WASTE
- 'RADIONUCLIDE INVENTORIES
The quantities present at time of processing in wastes resulting

from 33,000 MWA of exposure in typical LWRs, LMFBRs,and MSBRs are given
in Table 1 for those nuclides found to be of importance in the evaluation
of the long-term safe disposal of radioactive wastes. The LWR considered
has been described in ref., 3. It is fueled with 3.3% enriched uranium
and is operated at an average specific power of 30 Mwymetric ton of heavy
metal charged to the reactor to a burnup of 33,000 MWd/metric ton. The
fuel is assumed to be processed at 150 days after discharge, with removal

of 99.5% of the uranium and plutonium for recycle,

 
Table. 1. Quantities (Kilograms) of- Long-Lived Hazardous Nuclides Present
at Time of Processing for 33,000 MWd of Exposure in Reactors Utilizing
the Enrlched Uranlum, Pu-2387, and 233U-Th Fuel Cycles

 

 

Isotope Half-Life Enriched Uranium® P 2337.Th®
% Sr 27.7 v - 0.5k - 0.302 0,918
1297 1.7 x 107 y - 0.231 o o.2l1 0.35
232Th 1.41 x 1010 y - . - 24L
233pg T .27.0 4 o o 0.0042
233y 1.62'x 105 y 0.0167
2347 2.47 x 105 y | 1.15 x 1075: 1.36 x 10-4 0.0041
235y 7.1 x 108 y 0.0k ‘ 0.00745 0.0011
2367 2.39 x 107 y 10,0226 - 5,52 x 1074 0.0011
2387 4.51 x10° y - 4,72 4.38
237Np 2.14 x 10F 'y 0.483 0.128 -~ 0.0317
238py . 864y 8.4 x 1074 0.00534 ' 0.303
. 239py - 24,390 y "~ 0.0265 0.288 © 0.0037
240 py 6580 y 0.0107 0.0997 1.1 x 10-%
241py 13.2 y 0.0050 0.026 - - 5.67 x 10~
242py 3.79 x 105 0.0017 0.016 " |
2alpm - L33y 0. 0446 0.472
2438y 7950 v 0.0925 0.2h4g
2420 2163 4 - 'bf00582 0.0188
244Cn - 18.1y 0.0278 0.0145

 

3 3% enriched uranium irradiated at an average specific power of
30 MW/metric ton to a burnup of 33,000 MWd/metric ton in a typlcal
PWR. Proce331ng at 150 days after discharge.

bLWR discharge Pu and depleted U irradiated in core and blanket of AT
Reference Oxide LMFBR at an.average specific power of 58 MW/metrlc ton
to a burnup of 33, OOO MWd/metrlc ton. Processing at 30 days after discharge.

Reference MSBR equilibrium fuel cycle with continuous chemical processing by
fluorination-reductive extraction and the metal transfer process.
The LMFBR considered was the Atomics International Reference Oxide
Design.:‘}'5 The mixed fuel discharged from the core and blankets of this
reactor has been irradiated to an average burnup of 33,000 MWd/metric ton
of heavy metal charged to the reactor at an average specific power of 58,2
MW/metric ton. The fuel is assumed to be processed at 30 days after dis-

charge with a 99.5% recovery of uranium and plutonium,

The MSBR is a fluid fuel reactor that operates on the Th-233U fuel
cycle.6 The present concept employs fluorination-reductive extracfion
of the fuel salt to isolate 233Pa outside the reactor core with a 10-day
removal time, This chemical processing step is also responsible for
removing plutonium and a number of fission products from the fuel salt
on & 10-day cycle. Strontium, barium, and the rare-earth fission products
are removed from the fuel salt by an extraction process called the metal
transfer process with removal times ranging from 16 to 51 days. In addi-
tion, salt containing thorium is discarded to waéte on a hZOO—day cycle.
This mode of operation, which results in thorium utilization of only -
13.7%, makes fairly inefficient use of fertile material relative to the
LWR and IMFBR concepts. The reference MSER has a yield of 3.2% of the -
reactor fissile inventory per year, and it was assumed that 1/2% of the
uraniufi removed from the reactor for sale was lost to waste. Also, high-
level wastes are removed from the system in batches every 220 days follow-
ing fluorination to recover uranium which might be present in the waste
streams. It was assumed that 1/2% of the uranium in the waste streams
was not recovered by the fluorination, and that all the 233Pg which
remained at the end of the 220 days was lost with the fission product

waste,

It should be noted that the LMFBR and MSBR are advanced concepts
with thermal efficiencies of 40O to L5%, while typical PWRs achieve thermal
efficiencies of about 32%. 'Consequently, an LWR generating the same
electrical energy as an LMFBR or an MSER will produce 25 to 40% more
waste than that given in Table 1 or in subsequent tables since they are
all based on 33,000 MWd of heat production.
3. HAZARD AS A FUNCTION OF AGE OF HIGH-LEVEL WASTE

The composition, radioactivity, and hazard of the three types of
waste were computéd.as a function of age for tifies up to 30 million
years. Tables 2-L4 show the radioactivity of a number of isotopes of
special interest as a function of age. In these tables, the actinideé
are grouped according to their decay chains, so that it is easier to
observe daughters building up and gradually reaching equilibrium as

~ their precursors decay.

Table 5 presents the values used for the RCG for the isotopes found

to be most important in determining the long-term hazard.

A measure of the ingestion hazard associated with a radionuclide
is the quantity of water required to dilute the nuclide to the RCG for
unrestricted use of the water; the larger the amount of Wéter required,
the greater the hazard. - The volumes of water required for the three
types of wastes are shown in Table 6,and the isotope'which is the
principal hazard at a given time is shown in parentheses. The fission
product 2°Sr is the principal ingestion hazard for the first few hundred
years and the hazards are about the same for the three types of wastes.
In the first 30 to 300 years after disposal, the measure of hazard
associated with the wastes decreased from around 10! cubic meters of
water to around 3 x 108 cubiC'meters,due primarily to thé decay of
P05y, For the period 300 to 3000 years after disposal,the 222U fuel
waste is somewhat less hazardous than the others due to the absence of
transplutonium isotopes. The hazard associated with all the Wasfes
rises slightly at about a quarter of a million years after disposal,
which is due to peaking of the 22€Ra activity. The hazard associated
with the 233U fuel wastes diminiéhes less quickly than the other wastes
due to the presence of the relatively large amount of *32Th, the parent
of the isotope 228Ra{which is the predéminant hazard in this waste after

10 years).
Table 2. Typical Radiocactivity (Curies) of Long-Term Hazardous Nuclides in Waste from a Thermal

Reactor Fueled with Enriched Uranium as a Function of Age for 33,000 MWd Exposure

A

- Age of Waste (years)

 

 

Nuclide - 107 108 : 104 105 108 107

90 Sr 6480 1.5 x 1078 |

1297 0.038 " 0.038 0.038 0.038 0.036 0.025
241 A 145 o 3k | 1.9 x 10°%

243 A - 17.6 16.3 T2 . 2.09 x 1072

239py 1.68 | 2.06 - 4.06 0.57

234y 0.022 0.0h2 _ 0.041 , 0.032 L.0 x 1073 1.6 x 1073
226 Ry, 1.7 x 1077 © 5.3 x 1078 2.65 x 1078 0.021 | 5.2 x 1078 2.3 x 10-8
Total B~ 33,900 37 27.1 15.5 k.o 0.13
curies ' . _ .
Total Q . 276 61 14.9 | 2.2 - 0.15

curies

 
Table 3. Typical Radioactivity (Curies) of Long-Term Hazardous Nuclides in Waste from a Fast

Reactor Fueled with Pu and 238U as a Function of Age for 33,000 MWd Exposure

 

Age of Waste (years)

 

 

Nuclide ' 10° 10° - 10t ' 108 | 108 107
90 5y 3.62 x 10® 8.27 x 1077
1297 - 0,035 0.035 ~ 0.035 0.03k4 0.033 0.023
341y 1.39 x 102 356 4.33 x 1074
243 ppy W74 L3.7 19.3 5.62 x 1072
239py 17.8 18.5 21.0 2.31
2347 0. 0847 ~0.187 0.183 0.1k2 0.0127 0.0015
226Rg 6.29 x 1077 - 2.22 x 107 1.17 x 1072 -0.0931 0.0181 0.0021
Total B~ 31,400 68.5 b2, 2 a7 4.2 0.29
curies _ '
Total 1760 Lu6 50.2 " 5.1 3.2 0.17

curies

 
Table L.

Typical Radioactivity (Curies) of Long-Term Hazardous Nuclides in Waste from a Thermal
Reactor Fueled with 233U and Th as a Function of Age for 33,000 MWd of Operation

 

Age of Waste (years)

 

Nuclide 102 10° 10% - 108 10° 107
90 gy 11,000 2.5 x 106
1297 0.057 0.057 0.057 0.057 0.055
231pg 0.661 0.649 0.535 0.078 1.01 x 1076 1.00 x 105
223Rp, 0.635 0.649 0.535 0.078 1.01 x 1078 1.00 x 1076
2347 1.02 1.87 1.82 1.4 0.113 3 x 10711
226Rg .06 x 105 .50 x 1073 0.118 0.926 0.167 3 x 10712
232Th 0.0266 0.0266 0.0266 0.0266 0.0266 0.0266
228Rg, 0.0266 0.0266 0.0266 0.0266 0.0266 0.0266
237Np 0.022 0.022 0.022 0.022 0.016 8.8 x 107¢
2337 0.199 0.198 0.190 0.129 0.0028 8.8 x 1074
229Th - 0.039 0.050 0.132 .. 0.136 0.0029 8.9 x 10-4
228 Ra 0.037 0.050 0.132 0.136 0.0029 8.9 x 107
Total B L4 400 17.7 16.7 13.2 L.h 0.2k
curies . , _ .
Total o 2360 11 9.6 12.9 2.1 0.22

curies

 
9

- Table 5.HfiThe;RCGs-Usedgin‘Evaluating.the Ingestioanazards
from Radionuclides (from 10 CFR:20, Table II, Column:2)

 

RCG

 

Nuclide Half-Life - (ci/m?)
90 gy : 27.7 v 3 x 10-7
1297 1.7 x 107 y 6 x 1078
223Rg, | 11.4 g 7 x 1077
226Rg - 14.8 a 6 x 10~7
226 Ry, - 1620 @ 3 x 1078
22°Ra | 6.7y 3 x 1078
229 7340 ¥ 2 x 107®
230 Th | | 8 x 10% y 2 x 10-®
2321 1.41 x 101° y 2 x 10-®
231pg - 3.25 x 10% vy 9 x 1077
233pg - 27.0 d 1 x 107%
233y 1.62 x 105 y 3 x 107®
2347 2.47 x 108 y '3 x 10°°
2357 7.1x 108y 3 x 107
R367y 2.39 x 107 y 3 x 107°
238y 4,51 x 10° y 4L x 10°°
237 Iip 2.14 x 10° y 3 x 107°
238 py 86.4 v 5 x 1078
239y - | 24,390 y 5 x 1078
240 pyy - 6580 y 5 x 107
241py - 13.2 y .2 x 1074
242py  3.79 x 108 y 5 x 1078
24lpm 433 y 4 x 107¢
2437y 163 d h x 107®

 
10

Table 6. Volume of Hz0 (Cubic Meters) Required to Dilute Wastes
Resulting from 33,000 MWd of Exposure of Enriched 235U, 23%py,
and 223U Fuels to Levels Permitted for Unrestricted Use
(RCG from 10 CFR 20, Table II, Column 2)

 

Age of Waste

 

(years) 2362 239y P 2337C

30 1.26 x 10™ (®8r)  7.22 x 10%° (%§r)  2.11 x 101 (%0%5r)
100 2.24 x 1019 (®°gr) 1.32 x 101° (90gr) 3.75 x 10%° (®03r)

300 © T 2.00 x 108 (°°g8r) 3.93 x 108 (241an) 3.76 x 108 (%°3r)

1,000 1.55 x 107 (241Am) 1.09 x 108 (241pm) L.72 x 108 (223Ra
| and 228Ra)

3,000 6.53 x 10° (242Am) 2.2h x 107 (243pm) 5.06 x 10 (223Ra
and 228Ra)
10,000 h.26 x 108 (239Ppy) 1.26 x 107 (243Am) 9.34 x 10F (226Ra)
30,000 2.4 x 10° (239py) 6.88 x 10f (239py) 2.2 x 107 (226Ra)
100,000 2.14 x 108 (2°%Ra) 5}7h'x 10° (226Ra) 4,45 x 107 (228Ra)
300,000 2.38 x 10° (226Ra) 5.81 x 10° (226Rg) .68 x 107 (225Ra)
1,000,000  1.58 x 10f (1291) 2.03 x 10° (1297) 9.42 x 1CF (226Ra)
3,000,000 9.93 x 10° (2°1) 9.53 x 105 (12°71) 1.80 x 10° (228Ra)
10,000,000  5.28 x 105 (12°1) 4.88 x 105 (129T)  1.56 x 1CF (228Ra)
30,000,000 2.57 x 105 (297) 2 ¢ ] 1.20 x 10F (22°Ra)

.38 x 105 (1291)

 

2.0 x 108 = volume of water (cubic meters) required to reduce ingestion hazard
of the corresponding amount of uranium ore (7900 tons of ore contain-
ing 0.17% U).

volume of water (cubic meters) that results from dissolving the salt
required to store waste equivalent to 33,000 MWd exposure to a termi-
nal concentration of 500 ppmn.

1.1 x 107

1l

1.07 x lO6 = gpproximate volume of water (cubic meters) required to reduce
ingestion hazard potential of the corresponding amount of earth
containing naturally occurring U + Th in equilibrium with their
daughters at the average concentration in the earth's crust.

aReference PWR fueled with 3.3% enriched U, operated at a specific power of
30 MW]metrlc ton of heavy metal charged to reactor. Processing losses of
1/2% of U and Pu to waste are assumed.

bAI Reference Oxide IMFBR mixed core and blankets fueled with LWR discharge
Pu and diffusion plant tails. Average specific power of blend is 58.2 MW/metric
ton of heavy metal charged to reactor. Processing losses of 1/2% of U and Pu
to waste are assumed. -

CReference MSBR with continuous protactinium isclation on a 10-day cycle and rare-
earth removal by the metal transfer process. Thorium is discarded on a L4200-day
cycle, and 1/2% of excess uranium is assumed to be lost to waste (see text).
11

Lk, LONG-TERM RELATIVE HAZARD OF HIGH-LEVEL WASTE

The isotopes ®2°Ra and 228Ra are the predominant alpha-emitting
radionuclides for ages greater than 100,000 years., These isotopes occur
in nature as daughters of 238U and 222Th and,. if they are sufficiently
dilute in the waste, will present a hazard similar to those of naturally
occurring uranium. and thorium deposits. In the proposed Federal Reposi-
tory in bedded_salt,.the waste equivalent to 33,000 MWA of exposure may
be assumed to be associated with about 5500 metric tons of salt and 2400
tons of .shale, This is based on the current design of the high-level
facility of the Repository, assuming that the accumulated waste from
3.2 x 10° MWd(t)* of exposure is dispersed in the 900-acre by 300-ft-
thick section of the bedded salt. ©Since the LWR or IMFBR waste from
33,000 MWd(t) of exposure will have a volume and a mass of about 3.3 ft®
and 140 kg, resPectlvely, the salt bed serves to dilute the wastes by
approximately a factor of 37,000 in volume and 56,000 in weight. (The
dilution would not be asAgreat for the present MSBR concept because of
the larger qfiantity of waste generated per 33,000 MWd of operation. )
This dilution results in an average R387 concentration in the bedded
layer of salt ahd‘shale of less than 1 ppm by weight for the 2357 and
239py wastes, and an average 232Th concentration of 30 ppm by weight for
the 2337 fuel wastes The amount of 934U which is present in the waste
1n1t1ally, plus that Wthh is produced by decay of 242Cm and 238Ppy, is
well above that whlch would be in equlllbrlum with 1 ppm of naturally
occurring 238U, The quantltles of 2247 which ultlmately occur in the
wastes correspond to those which would be in equilibrium with 15, 5k,
and 680 wt ppm 238U in the bedded layer of salt and shale for the

wastes from 235U, 23°%Py, and 2"3"3U fuels, respectively.

The average concentratlons of uranium and thorium in the earth's
crust are 6 and 12 ppm, respectlvely, and the volume of water requlred
to reduce the ingestion hazard of the corre3pond1ng mass of earth (7900

metric tons) containing these concentrations of uranium and thorium in

 

* _
The most recent forecast7 estimates that the high-level waste to be

accumulated in the Federal Repository, up to the time of closing in
year 2000, is 319,000 ft?,Whieh is equivalent to a total exposure of
3,19 x 10° MWd(t). Most of this waste will be from LWRs since most of
the IMFBR waste will not yet have been delivered to the Repository.
12

equilibrium with their daughters is 1.07 x 10° m® (see Appendix). The
" hazard associated with the LWR type waste (the predominant type in the
‘Repository) is only a factor of four higher 10,000 years after disposal,
and the hazards from the other wastes are about an order of magnitude
higher. However, the hazard associated with the 233U fuel waste rises
again as the 226Ra concentration increases. The presence of 226Ra and
228Rg in high concentrations in the waste from the 232U-Th system is
‘the result of the absence of a step for recycling plutonium and the
assumed inefficient use of thorium in the present reprocessing schemes.
If the processing flowsheets are improved to eliminate these two
problems, the 233U fuel wastes would remain very similar to the 235U

and 22°9Pu wastes at ages greater than 10,000 years.

Other thermal reactor concepts employing the Th-233U fuel cycle are
the ngh Temperature Gas Cooled Reactor (HTGR) and the Light Water Breeder
Reactor (LWBR). The wastes produced by these reactors would be similar
to MSBR wastes since the higher 9°Sr’concentra.tions Would'be present
initially, since 238Pu will not be separated from the waste and recycled,
and because efficient recycle of the relatively inexpensive thorium is
not envisioned.at present, Wastes from 228U fuels aged 30 years have a
thermal'power 64% higher than 23°Pu wastes of the same burnup, and wastes
aged 100 years have a thermal power h6% greater Since the peak tempera-
tures in the mine are reached at about 50 years after burlal, the hlgher
heat generation rate from ®33U fuel wastes will require a greater spacing
between waste cans, with a subsequent increase in cost. The burial cost

of the waste is a small fraction of the fuel cycle cost, however.

An alternative method of assessing the relative hazard of the radio-
active wastes is to compare them with naturally occurring radiocactive
surface deposits. The southwestern United States has uranium ore reserves
containing 0.2% U305 or greater, totaling almost 50 million tons. In the
year 1968, domestic production of U,04 totaled over 12,000 tons, indicating
that some 6 million tons of such ore were processed and & corresponding
amount of uranium ore'tailings ‘was disposed of.‘ The volume of water
required to dilute the radionuclides contained in 7900 metric tons of

uranium ore tailings to the RCG for unrestricted use of the water is
13

2.0 x 108 m®, Consequently, the ingestion hazard associated with a
quantity of uranium ore and uranium ore tailings equal to the amount

of salt and shale asseciaied With 33,000 MW4 of Waste‘is greater than
that associated with any of the three high-level wastes aged 1000 years.

Another method of assessing tne relative hazard of a radioaetive waste
is to compare it:with the hazard of naturally radioactive monazite sand.
Thorlum and uranium are present 1n the mlneral monazite which occurs in
beach sand in some areas of India, Brazil, Malaysia, and the southeastern
United States.8 Beach sands in Malagasy Republic average 2 to 2, 5%
monaz1te, the monazite containing 8.8% ThOz and 0.41% Us0g. Monazite .
sands found on the southeast coast of the United States average 3.1%
Th024and O.H?%,Uéoa. To reduce the ingestion hazard from radioactivity
ifi these two sands for a mass‘equivalent to the waste and associated
salt in the prOposed"Federal.Repository will require about 6 x 107 and
2 x 107 m?,of wafer for the Malagasy and United States sands, respectively.
Henee, dissolving the buried 235U or 233U fuel wastes aged in the range
of 1000 to 3000 years, along with the associated salt, would present a
hazard similar te.dissolving the radium from the same mass of monazite-
contalnlng beach sand (noting, of course, that monazite sand is among

the least soluble of naturally occurring materials).

Another consideration with respect to the disposal_of.high-ievel
radioactive wastes in bedded salt is that to dissolve the naste for a
33,000 MWd exposure and associated salt to a potable concentration of
500 ppm NaCl by weight would require 1.1 x 107 m® of water. Thus, 235U
and 233U fuel wastes aged 3000 years, if diluted to drinking water con-
centration in NaCl, would be below the radiation concentration guide for
ingestion by the general population. . Hence, if the salt mine is dissolved
some thousands of years in the future with sufficient water to dilute the
radionuclides to the RCG and the dissolved materials find their way to:
drlnklng Water supplles, the water would be unacceptable as drinking _

water because of 1ts sodlum chlorlde content,
1k

5. .ISOTOPIC DILUTION OF HIGH-LEVEL WASTES

It'has been suggested that a radioactive isotope may be dispoeed'of
without further dilution provided that it is diluted to an acceptable
specific activity with an adequate quantity of the stable element in the
same chemical form and if the diluted material is also acceptable on the
basis of external hazard and chemical toxicity.9 Consideration has been
given to this dilution method for'reducihg the ingestion hazard associated
with the high-level wastes. Table 7 lists the isotopes which present the
greatest hazard at a given age of the #aste; their radicactivities, and
the quantities of stable elements which would be required to dilute the
wastes below the maximum permissible specific activities given‘in ref. 9.
In ref., 9 it is assumed that the isotope 9°Sr must be diluted with stable
strontium, rather than another bone-seeking element (sueh as Ca), but
that the elements americium and plutonium may be diluted with rarée earths.
It is also assumed that an "acceptably low" level of specific activity
is the vaiue such that, if a person were to assimilate the element (or
chemically similar species) only from the source of interest, the body
burden of the radioisotope would not exceed the maximum permissible body
burden for occupational exposure. USing these assumptions it can be
calculated that 432 metric tons of stable strontium are required to dilute
the 2°8r in 100-year-old 235U waste resulting from 33,000 MWd of exposure
to an acceptably low specific activity. Since each 33,000 MWd of waste
is associated with 8000 metric tons of salt and shale, a concentration
of 5% by weight stable strontium would be required throughout the salt
mine, which would be impractical. The concept may prove feasible for .
diluting waste aged 300 years, since only 3 metric tons of stable

strontium would be required for each 33,000 MWd exposure.

If it is assumed that the behavior of americium and plutonium in
the body is similar to that of rare earths, then 49 and 5.7 metric tons,
respectively, of rare-earth elements are required to be added to the
wastes to eliminate the ingestion hazards from 241Am and 239Py, Also,
if 250 g of stable iodine is added to each 33,000 MWA of waste, the

ingestion hazard from 12°I becomes acceptable for occupational exposure.
Table 7. Mass of Stable Carrier Required to Dilute Radioisotopes in 238U Fuel High-
Level Waste to Maximum Permissible Specific Activity in the Environment

 

Mass of Carrier

 

 

_ , - Maximuma Required to Dilute
Age Isotope Radioactivity Carrier Critical Permissible to Specific Activity
(years) (Ci/33,000 MWd) . Organ Specific Activity (metric tons)
(Ci/metric ton) 33,000 MWd Exposure
100 90 gy 6480 Sr Bone 15 432
300 ogr W67 Sr " Bone | 15 o 3.1
1,000 241pm o 34.h rare earth  Bone 0.7 L9
10,000 239py k406 rare earth Bone - 0.71 5.7
100,000 22§Ra 0.021 . b Bone b | ' -
1,000,000 1291 0.036 T  Thyroid 80° ' 4.5 x 1074

 

*National Academy of Sciences-National Research Council, Publication 985, Disposal of Low-Level
Radioactive Waste into Pacific Coastal Waters, Washington, D. C., 1967.

 

bInadequate biological data to epply the specific activity concept to 226Ra. -

“Calculated from data given in International Commission on Radiation Protection, Report of Committee II
on Permissible Dose for Internal Radiation (1959), Pergamon Press, Oxford, 1960.

 

ST
16

6. RELATIVE HAZARD OF TRANSURANIUM WASTES

For disposal in the Federal Repository, Godbee afid,Niéholle have
proposed a lower limit for transuranium or "alpha":waste of 10 uCi/kg of
initial parent alpha activity. Alpha wastes (which mostly arise from
fuel fabrication) of lower activity are to be disposed of by surface
burial.‘ To evaluate the hazards associated with surface burial of alpha
wastes,an investigation was made of the relative ingestion and inhalation
hazards of a nufimer of isotopes and mixtures of isotopes expected to be
present. Table. 8 compares the radioactive properties of uranium-and
thorium in the average earth's crust, uranium ore, and‘firanium ore
tailings with those of a number of transuranifim wastes contaminated to
an initial parent alpha activity of 10 uCi/kg. Shown in the table -are
the properties at the time of isolation of the paren£s~and at the time
of maximum ingestion hazard;' At the time of maximum hazard, the wastes
present an ingestion hazard similar to uranium ore tailings althofigh
their total activities’, thermal powers, and inhalation hazards are about
an order of magnitude greater than uranium ore tailings. The maximum
ingestion hazgrd resulting from an initial parent alpha activity of
10 uCi/kg of either natural uranium or natural thorium is larger by an
order of magnitude than any of the other wastes considered. For some
transuranium materials, 238Pu, 242Cm, and *44Cm, the maximum ingestion
hazard is associated with the initial 10 uCi/kg of parent alpha activity.
The ingestion hazard associated with low-level plutonium wastes contami-
nated to a specific activity of 10 uCi/kg is appreciably less than that
associated with a kilogram of uranium ore tailings, and the ingestion
hazard associated with plutonium wastes is comparable to the maximum
hazard associated with other alpha wastes contaminated to the same
initial specific activity. The ingestion hazards for various wastes
are sufficiently similar that a criterion based on total alpha activity
alone, regardless of isotopic composition, may be sufficient for classi-

fying transuranium or alpha wastes.
-
.

Table 8. Comparison of Radioactive Properties of Naturally Occurring Thorium and Uranium Deposits
with the Assumed Lower Limit for Classification as Transuranium Waste (10 pCi/kg)

 

Total Activity Cubic Meters of Water or Air at RCG®

 

 

Alpha Activity Time Since /kg) Thermal . :
Source of Parents Isolation of (1Ci/ke Power ~ E=C Kilogram Dissolved or Suspended
(uCi/kg) Parents (y) Beta Alpha (nW/ke) Water® AirP
Average Earth Crust® .
U 0.0021 4.5 x 10° 0.012 0.016 0.00058 0.091 (22€Ra) 31,300 (23°Th)

Th 0.0013 | k.5 x 10° 0.0052 0. 0078 0.00030 0.045 (228Ra) 9,250 (228Th)
Total 0.0034 L5 x 10° 0.017 0.02k4 0.00088 0.136 (22%Ra) 41,000 (23°Th)
Uranium Ored 0.59 L.5 x 10° 3.5 b7 0.16 25.6 (226Ra) 8.8 x 108 (230Th)
Uranium Tailings® 0.029 4.5 x 10° 3.5 4.1 0.15- 25.6 (22%Ra) = 8.6 x 108 (=2307y)
Natural U - 10 0 0 10.0 0.27 0.292 (234y) 2.90 x 10° (238y)
Natural U 10 1P 57.1 76.2 2.76 433 (226Ra) 1.5 x 108 (239Th)

Natural Th 10 o . 0 . 10.0 0.24 5.0 (232Th) 1 x 107 (232Th)
Natural Th 10 25 Lo.o" 60.0 2.31 °© 346 (228Ra) 7.1 x 107 (228Th)
TWR pul - 10 0 470 10 0.34 | L4 (241py) 3.1 x 108 (241py)
LWR Pu 8.6 308 114 20.4 0.68 - 5.23 {(241Anm) 2.3 x 10° =238py)
238Ppy 10 0 0 10 - 0.33 . 2.0 (®38py) 1.4 x 10® (238epy)
236 py 10 0 0 10 0.35 h h

236 py 0.0068 308 0.63 1.89 . 0.075 0.23 (22%Ra) 1.9 x 10 (228Tnh)
2420y 10 0 0 10 0.37: 0.498 (242Cm) 2.5 x 10F (242Cnm)
2440 10 o - 0 10 0.35 0.143 (244Cm) 3.3 x 107 (244Cm)
2327 10 0 0 10 0.32 0.33 (232u) 1.1 x 107 (232y)
282y 9.1 108 18.1 54, 3 2.1k 6.6 (224Ra) 5.6 x 107 (228Th)
233y 10 0 0 10 0.29 0.333 (=22U) 2.5 x 10F (233y)
233y 8.8 3 x 1048 25.7 51.8 1.96. - 18.9 (225Ra) . 1.1 x 108 (229Th)
93% 238y 10 0 0 10 0.27 0.330 (238y) 2.5 x 10F (235y)
93% 236y 10 - 1088 uo.p 69.8 2.6 36 (223Ra) - 3.8 x 108 (231pa)
237Np 10 0 o - .10 0.29 © 3.33 (227Np) 1 x 108 (237Np)
237Np 9.1 3 x 1058 29.8 50. 3 1.86 18.3 (225Ra) 1.8 x 10% (237Np)

 

%Based on radiation concentration guides for continuous exposure in unrestricted areas (10CFR20, Table II).
bThe principal contributor to the hazard is shown in parentheses. |

“Assumes U and Th concentrations of 6 and 12 ppm, respectively.

dassumes ore containing 0.17% U. . _

“Uranium ore tailings are assumed to have 5% of the original uranium but all of the radioactive daughters.
Tpu is assumed to have isotopic composition of 1% 238Pu, 59% 239pu, 249, 240py, 129 241Py, and 4% 242py,
gTime since isolation of parents at which maximum ingestion hazard occurs. |

hRCG values are not defined for 238py,

LT
10.

18
7. REFERENCES
D. E. Ferguson et al., Chem, Technol. Div. Ann. Progr. Rept. May 31,

1969, ORNL-4422, p. 89.

M. J. Bell, ORIGEN - The ORNL Isotope Generatlon and Decay Code,
ORNL.-4628 (1n preparation).

M. J. Bell, Heavy Element Composition of Spent Power Reactor Fuels,
ORNL-~TM- 2897 (May 1970).

K. Buttrey, O. R. Hillig, P. M. Magee, and E. H. Ottewitte, Liquid
Metal Fast Breeder Reactor - Task Force Fuel Cycle Study, NAA A-SR-
MEMO-12604 (January 1968 ).

 

:Staff of the Chemical Technology Division of ORWNL, Agueous Processing

 

of IMFBR Fuels - Technical Assessment and Experimental Program Defi-
nition, ORNL-4436 (June 1970).

 

M. W. Rosenthal et al., MSBR Program Semiann. Progr. Rept., February
28, 1971, ORNL-U4BT6.

 

F. L. Culler, J. O. Blomeke, and W. G. Belter, "Current Developments

in Long-Term Radioactive Waste Management," paper A/CONS/L9/P839,

presented at the Fourth United Nations Conference on the Peaceful
Uses of Atomic Energy in Geneva, Switzerland, Sept. 6-16, 1971.

U.S. Bureau of Mines, Bulletin 630, Mineral Facts and Problems,
1965 ed., "Thorium," pp. 947-959.

 

National Academy of Sciences-National Research Council, Publication
985, Disposal of Low Level Radioactive Waste into Pa01f1c Coastal
Waters, Washington, D. C., 1962,

 

 

H. W. Godbee and J. P. Nichols, Sources of Transuranium Solid Waste
and Their Influence on the Proposed National Radioactive Waste
Repository, ORNL-TM-3277 (January 1971) (Official Use Only).

 
APPENDIX

RADIOCACTIVITY IN EQUILIBRIUM WITH NATURALLY OCCURRING URANIUM AND
THORIUM AND THE ASSOCIATED INGESTION AND INHALATION HAZARD
(Cubic Meters) Potentially Contaminated to RCG by Radiocactive

20

Table A.l1. Radiocactivity and Volumes of Air and Water

Daughters in Equilibrium with 1.0 g of 232Th

 

 

Radiocactivity Ingestion Hazard Inhalation Hazard

Nuclide (uCi) (cubic meters of water) (cubic meters of air)
L 0.109 ' 0.0547 1.09 x 108
228Rg, 0.109 - 3.65 1.09 x 108

228 pc 0.109 1.22 x 1073 1.82 x 102
228Th 0.109 0.0156 5.47 x 10°
224Rg, 0.109 0.0547 5.47 x 10°
220Rn 0.109 a 1.09 x 10t

216 pg ; 0;109 a. - “’ a

212pp 0.109 5.47 x 10-28 1.82 x 102
212Bj 0.109 2.73 x 10-% 3.65 x 10!
212pg 0.070 a a

=208m 0.039 a a

Total 1.09 3.78 7.71 x 108

 

-aRCG not available; isotope is ignored in calculation of hazard.

 

’,
21

-Table A.Z, Radioactivity and Volumes of Air and Water
(Cubic Meters) Potentially Contaminated to RCG by Radioactive
Daughters in Equilibrium with 1.0 g of Natural Uranium

 

 

Redioactivity Ingestion Hazard Inhalation Hazard
Nuclide (rCi) (cubic meters of water) (cubic meters of air)
2387 0.331 " 8.27 x 1073 1.1 x 10°
234Th - 0.331 ' 1.65 x 1072 o 331
234 py, 3.31 x 10-3 a ' a
234Mpg 0.331 a a
2347 0.331 1.10 x 10~2 8.27 x 10%
230 Th 0.33L 0.165 .13 x 10°
226 Rg, 0.331 11.0 1.65 x 10°
222Rp 0.331 a 110
21l8pg 0.331 a a
214pp 0.331 a )
2l4Bj 0.331 a, : a
2l4pg 0.331 a a
210 py 0.331 3.31 8.27 x 10%
210 B 0.331 8.27 x 1078 1.65 x 103
210 pg 0.331 0.473 4,73 x 104
2385y 0.0154 5.15 x 10-% 3.86 x 103
231Th 0.0154 7.72 x 10-% 0.39
231pg 0.0154 0.0172 3.86 x 108
227 pAc 0.0154 7.72 x 1073 1.93 x 10°
227Th 0.0152 7.61 x 1074 2.54 x 108
2237y 2.2 x 1074 a a
223Rg, 0.0L54 0.0221 1.93 x 103
219Rn 0.0154 a a
218pg 0.0L54 a | a
211pp 0.015k4 a a
211Bj 0.0154 a a
21llpg h. 6 x 10-5 a a
2077 0.0L54 a _ a
Total 4.8 15.1 ‘ 5.21 x 108

 

®RCG is not specified; isotope is ignored in calculation of hazard.
-3

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