ORNL-~TM- 3064

Contract No. W-Ti05-eng-26.

CHEMICAL TrCHMCLOGY DIVISION

NEUTRON-INDUCED TRANSMUTATION OF HIGH-LEVEL RADIQACTIVE WASTE

H. C, Claiborne

DECEMBER 1972

 

 

NOTICE
Thlt report was prepared az sn account of work
tha linbed Secrex Gummnlnt__ Nekhwr

 

 

CAX RIDGE NATIONAL LABORATORY
Oak Ridge, Tennessee 37830
aperated by
UNION CARBIDE CCRPORATICON
for tae
U.S. ATOMIC ENERCY COMMISSION
iii

CONTENTS

Abs tract - - - " - - » - ” » - » + - - + - - . - - - - - -

AN

Cn

3

9.
10.

Appendix I: # Comparison of RCGs Calcuiated ty LaVerme (Ref
ersl

IneroduClion & o o o v o & o o o o o « = =« o = o s o .
SUHMEYY o s - s s o s = s+ e s s e s a s 4 o s e . . o
Method for Determining the Hazard of Radicactive Waste
Nuclesr Calculational Methed . . . . . . . . . . . . .
Reactor Type and Standard Conditians . . . . . . . . .

Contribution of Eack Component to the hHazardi of the
Waste from a PWR Spent-Fuel Processing Plant ., . . . .

Transmitaticn of Fission Preduct Waste . . . . . . . .

7.1 Maximm Buraout-to-Troduction Ratios for Fission
Pz‘wucts - a & & & & @ » L - - - - - - > - - -» -

7.2 Reactor Residence Times Reauired for Fission
PI‘OdnCt mo“t * - . + -, 4 - » - - . - - - - - -

7.3 Application of Transmuetation Schemes . . . . . .
Actinide Recycling Ina PWR . . . . « ¢+ o ¢ o - . .
801 HMheet . ’ L J » . - - - A - - - - - - - - » - -

8.2 Chemical Processing for Waste Marsgement
Simplification . & v ¢ ¢ 4 4 b e e e b e s . e

8.3 Effect of Recycle on Resctivity and Flux . . . .
8.4 Effect of -Recy<cling on Fazard Messure . . . . . .

8.5 Effect of Fecyeling on the Hazards of Chemical
Processing angd ruel rFabrication . . . . . . . . .

Conclusions and Recoammendations . . + « + o » o « » «

Reference s - - * - - L - - - - * - - - - & . - - L - -

Regulations (Ref. &) . .

- - - -

- - - .

)
 
iv

Aopendix II: Radioactivity and Hazard Messure of Each Actinide

Nuclide a2s 2 Funetion of Time After Discharge for the Standerd
Case and After the 50th Recycle .

- - - . - . . - . - . -

Appendix III: ZXHazard Reduction Achieveble by Erhanced Removal
of Actinide Zlements (Ref. 19) !

s v = * . - » T s e *r - v .

 
NEUTROR~INDUCED TRARSMUTATION OF HIGH-LEVEL RADIOACTIVE WASTE

H, C. Claiborne

ABSTRACT

The possibility of reducing the potential hazard of high-
level radiocactive waste by neutron-induced transmutation has
received little study. In this report the mvailagble information
on fission product transmatation is reviewed and discussed, the
contribution of individusl actinides Tto the potential hazard of
the waste is calculated, and expected hazard reduction factors
that would result from recycle through a PWR are calculated for
the actinide waste from chemical processing of spent fuel.

It is not practical to burn fission product wastes in power
reactors because the neuiron fluxes are tco low, Developing
special turner reactors with the reguired neutrca flux cf the
order of 1017 n/ eml-sec or burning in the blankets of thermo-
nuclear reactors is beyond the limits of current technology.

It seems that ultimate storage in deep geclogical formations,
such as bedded salt, remains the best method for {ission product
disposel.

When plutonium and uranium extraction efficiencies exceed
0%, a significant reduction in the long-term hazard potential
of the waste can be obtained by similar removal of neptuniwm,
americium, and curium (the other ectinides beirg very small
contributors). Consegquently, it seems reasonable to concentrate
on developing economical chemicel processes to extrect these
three actinides for seperste storage or for recycling through
the reactors that produce them.

The results of cuch recyceiing calculaztions show that the
long-term hazard potential. of the waste from light water reactors
may te reduced by factors up to 200 if no more than 0.1% of the
actinides are discarded to the waste in ealh pass through the
reprocessing plent. lLarger reductions of the hazard potential
of' the waste will become practical if methods are developed to
Froduce sharper separations between tne actinides and fission
products as the spent fuel is processed.

 
1. INTRODUCTION

The management of high-level, long-lived radicactive wastes associ-
ated with a2 highly developed nuclear nower economy based on fission
regctors will present & formidable problem to present and Twiture gener-
ations, Schemes for management of these wastes that Lave been under
serious consideration involve conversion of the squeous wastes to solid
forms with subsequent storsge in man-made vaults or in deep geological

formations such as bedded salf.

The possibility of ultimate disposal into deep space or the sun
(the only method for complete and permanent removal from the earth) has
begun tc raceive more considerafion because of the recent and prejected
advances in space technology. The only otker known method of ultimate
disposal (in contrast to permanent storage) is to transmute or burn out
(fission in the case of some of the actinides) long-lived radiocactive
nuclides to stabie or short-lived nuclides by expesure to & neutron
flux.

Studies have bheen madel’z cn the posaipility of using special high-

flux "burner reactors” to reduce the stockpile of the "problem fission

products" 85Kr, =0 L3es,

Sr, and The excess neutrons from controllsed thermo-

3,k

these fission products and the waste actinides,

nuclear reactors have also been sugsested for use in transmutation cf

Aside from the problems associated with burning fission preducts
(which are discussed later in this report), 908r and ~31cs decay to com-
pletely innccucus ievels in less than 1000 years, a time for which ceon-
tairment in appropriate geclosicel formetions can be provided with good
assurance, The nuclides 85Kr and 3H with shorter half-lives are even
more suitable for long-term siorage in geologicel formations. The
igotope 1297 (half-life,16 million years) is one of the sxcepticnal
fission products that has an extremely long life but is produced in
such lew o-nZsuatraticns that its hazard may possibly be reduced tc
wopropristely low levels by isotopic dilution (i.e., by wmixing with

stable isotoves of the same chemical element).
In contrast, many of the actinides that are produced by transmutation
of uranium and thorium in reactors have half-lives in the thousands of
years, occur in large quantities, and are ncot suitable for isctopic dilu-
tion becauce stable forms of these slements do not exist. Consequently,
an aven stronger motive exists for completely destroying or restricting
the accumlation of these slpha-emitters since predictions of the tectonics
of geologizal formations for 105 to 10" years have a lower confidence
level compared to those for the order of 1000 years. In present concepts
of power reactors, it is planned that only 99.5 to 99.9% of the uranium,
plutoniuvm, and thorium will be recycled., Consequently, it is customarily
assumed that all other heavy elements (Cf, Bs, Cm, Am, Np, Pa, Ac, Ra,
ete. ) will be rejected as waste along with the 0.1 to 0.5% of the U, Pu,
and Th that goes to the waste in the present generation of spent fuel

reprocessing plants,

The hazard potential of this actinide waste can be reduced by recy-
cling the actinides through the power reactors prcducing them; eliminaticn
oceyrs by Tfission at points in the reaction path. The primary objective
of this work was to determine the extent of the reduction of the radio-
logical hazard of the waste streams from chemical processing plants and
the effect on the neutror eccnomy of a pressurized water reactor (PWR)
caused by recycling of the actinides (except for the smell smounts lost
in the waste streams) back through the reactors producing them. In
addition. the individusl contribution of each actinide %o tThe waste
hazard was determined as a funciion of decay time and compared with the
hazerd from all the wacste, which inciudesthe fission products, nuclides
produced from structural materials, actinides, and all decay products.

In the following sections the bases for calculations are given and
pertinent results are gresented and discussed. A modified version of
ORIGEN,5 an isotope generation and depletion code, and its associated
nuclear library was used in all the calculations,

The author wishes to agknowledge the many helpful suggestions and
criticisms by J. P. Nichols and the careful review of this work by him;
J. 0. Blomeke, and M. J. Beli.
2. CSUMMARY

It is generally impractical to appreciably change the haéard potential
of fission product wastes by transmuting these wastes with neutrons in
nuclear reactors. Developing special burner reactors with the reguired
neutron fiux of the order of 10t7 n/cm?'sec or burning in the;hlankets of
thernconuclear reactors is beyond the limits of current techno%ogy. It
appears thet ultimale storage in deep geological formations is the best
method for fission product disposal since less thar 1000 years are re-
guired to reduce their radiocactivity to an innocuous level, aitime span
for which tectonic stability can be essentially assured in fo%mations

such as bedded salt.

In contrast tc the fission products, meny of the actinid%s in the
waste from spent-fuel processing have half-lives of thousandsiof years
and are not suitable for isotopic dilution. Consequently, afstronger
motive exists to find an alternative method of restricting the accumi-
lation of these alpha emitters since the tectonics of geolog;cal forma-
tions cannot be predicted with as high a confidence level for the longer

periods that are required for their decay to innocuous levels.

The determinetion of the extent of the reduction of the radiclogical
hazard of the waste streams from chemical processing plants and the effect

on the neutron economy of a PWR caused by recycling of the qbtinides was

I

the primary cbjective of this study.

The relative importance of the contribution that the various com-
ponents make to the hazard messure (the total water required to dilute
each nuclide of a mixture to its RCG') of the waste from a2 PWR spenti-fuel
processing plant is sanown in Table 1. Beyond about 4 years, the acti-

nides ard their daughters dominzte from & hazard viewpoint. When RCGs

5,9

are used that are liess conservative than the recommended default

values of the Code of Federal Fegulations, the importan:% of the acti-

nides diminish scmewhat for decay times greater than 107 years.

The actinide waste hazard is controlled by the americium and curium

vp to lOL years, At longer decay times the long-lived ?37Np and its

 

+* .
Radiation Concentration Guide value, which was formerly called MFC,
\A

Table 1. Relative Contribution of Actinides anéd Their laughters to the
Hazard Measure of the Weste and of Each Actinide and Its Daughters
to Actinide Waste with 99.5% of U + Pu Extracted

 

Weter Required for Dilution to the RCGE (% of total

Muelides to waber required for the mixture) for Decay Times (yr) of:

 

 

 

Waste 10° 5 x 10° 10+ 10° 1?
411 Components of Waste:®
Actinides 0.3 ol ok 98 99
Fission Products C 99+ 5 6 2 1
Structuial Q.04 i 0.2 0.03 b ox 10'LF
Actinide Waste:b
Americium 51 56 2h 8 8
Curium 41 37 5S e
Neptunium 0.2 0.3 12 80 89
0.5% U + 0.5% Pu 8 7. 5 3 1
Other 5x 102 1x 100 5x10°2 6 x 1075 ail
8

a'Using CFR RCGs and recommended default velues for the unlisted muelides.

bRound-off may cause ccivrmn not to total 100,
daughters 2egin 1o dominate. Another important point is that the
remaining actinides, namely, Ac, Th, Za, Bk, Cf, and Es, mzke 2 negli-
givle contribution tc the hazard of the waste, The import of these
results is that in any wasie management system in which at least 99.5%
oL the uranium and plutonium is extracted, a significant further reduc-
tion in the actinide waste hazard can be obtained by removal of most
of the smericium,. curium, and neptuniuvm from the weste. If 99.5%
removal of these three actinides is 2lso effected, the uranium and
plutonium become controlling and it would then be profitable (from a
wzste hazard viewnoint) to increase the extraction efficiency of these

latter elements, particularly the plutonium.

Tne effect of recycling of 99. 5% 2md 99.9% of the actinides other
than U or 2u on the bazard measure is shown in Table 2 in terms of a
hazzrd reducticn factor as a funchion of postirradiation decay time.
The hazard reduction factor used here is defined as the ratio of the
water required for diiution of the waste to the RCE for the standard
case (nc removal of the actinides other than Pu + U at the indicated
extracticn efficiency) to that required to dilute the waste afier each

successive reactor irradiation cycle.

These resuits show that wren recycling is practiced, the hazard

. measure of the waste is apprdximately proportional to the neptunium,
americium, =nd curium sent to the waste sinces the hazard reduction
factor is about five times greater when 0.1% of the actinides is sent
to the waste alter each c¢ycle than that for tne 0.5% case. This obtzins
logically because the reactor discharge composition is little affected
Ty a change of only 0.4% of recycled actinides in the feed stream. In
addition the standard case iz alss little affected by whether 0.1% or
0.5% of U + Pu is present since the americium and curium predominate at
snaller decay times and neptunium after 105 years. It follows that if
99.99% remcval of all actinides is effected, the hazard reduction
factior for the actinide waste will increase by about a factor of

10 up to around 2000 at 106 vears, The table also shows that

the hazard reduction factors decrezse asymptotically with the

number of reeycles, which is a result of the bulidup
Table 2.

Effect of Recycle of Actinides Other Than U and Pu on the
Hazardé Measure of Waste from PWR Spent Fuel Processing

 

Recycle

Water Recuired for Dilution to RCG,a Ratio of Standard

to Recycleb Case (Hazard Reduction Factor) for

Decay Times (yr) of:

 

 

No. 10° 10° 0¥ 107 10°
Actinide Extrection Efficiency, 9%.5%:
0 12 15 18 28 52
1 9.3 1z 13 20 L6
2 8.2 10 11 18 Lk
3 7.6 8.h S.2 17 L3
L 7.2 7.4 8.3 17 L2
5 6.8 6.5 7.5 17 L2
10 5.8 h.7 5.8 17 42
20 5.1 3.8 4.9 17 Le
30 5.0 3.6 k.6 17 42
Actinide Extraction Efficiency, §9.9%
o 58 73 8o 137 256
1 Lk 59 6k 96 22k
z 38 L8 52 7 213
3 36 Lo Lh 8k 210
L 33 35 39 83 209
5 32 31 16/ 83 208
10 27 22 27 83 207
20 - 18 22 82 206
30 - 17 21 82 206

 

EI'Using CTR RCGs and recommended defszult values for the uniisted nuciides.

°Chemical processing assumed al 150 days alter reactor discharge: one
cy¢le represents 3 years of reactor operation,

8
&

of the higher transuranics, and that effective equilibrium is attained

in 20 cycles mcre or less,dspending con the decay time.

when the RCGs tvaed dy Bell6 ard those calcvlated by LaVerne9 are
used in place of the recommended defanlt valves for the unlisted ruclides
in the Code cf Federsl Regulatiorns, tha hazard reduction fa¢tors become
5.5 and 10 rfspectively. The corrasponding values ©ar 99 9% extraction
of the actinmides ere 25 and 49. Although the RCGs calculated byhlaﬂerneg
are more realistic than the more conservative recommended default values,
the Code of Federal Reguleticns must be follovad in nuclear reactor

design and operation.

Reeyeling of the actinides and achicvirg a 99.9% extraction effi-
ciency reduce the hazard measure of the actinides at eguilibrium to the
same order as that of the long-lived fission products (1291, 932:, 93mNb,
99Tc, angd 35"5) for the longer decay times. the hazard measure of the
actinides being abcutl twice fhat of the long-liived fissicn precducts at
1000 years and dropping to aboul one-nalf of the fission produet veiue
at 106 years. However, if 1291 is eliminated as a hazzrl} by isctopic
dilution {or separate storage), the actinides would continue to control
the total waste hazard potential. An actinide extraction efficiency of
9G,99%-% along with the recycling is recuired before the hazaré measure
of the total waste hazard potential is controlled by the long-lived

. : iz
fission preoducts cther than 91.

At some point, however, Turther
extracticon of actinides from the waste will become senseless because

- e d T Y e 1 - P - 3 - 2 - - + -
tve Wwill then have a Long-torm nazard noitentizl that 1c less than

that of naturally occurring fcrmations of uranium ard thorium. (See,

\

O\

Tor example, the arguments presented in ref,

The decrease in the average materizl nsutron moltiplicaticon for a
typical DPWR containing recycled actinidec was only 0.8%. This loss of
regeTivity can be cormpensated by increasing the fissile enrichment of
the ra2acter by only about 2% {e.g., from 3.3 to 3.! L% erricament in 2

typical DWR).

L

Recycling of reacter actinide waste will increase the radiation

oroblem zssociated with chemical procassing and fuel fabrication because
of the increased radiocaltivity of tae reactor feed and discharge
sireams. After a few recycles,zﬁgcf builds ur to be the grestest

source of neutrons and reaches 1042 reutrons/sec per metric “on of

spent fuel at 150 days ai'ter discharge. A reduction ol 3 FJactor ol

300 is possible if the eplifernium is removed. This caz be accomplished
by not recycling Cf even though thares is an increase in the Cf proguc-
tion with curium buildup. Significant °7°CT buildup occurs from

' 240 2 \
successive neutron captures starting with © “Cr and SOCf, whoze

Precursors are Ehgﬁk engd 2503&.

Recycling of ectinides through a resctor adds to the inventory
of hazardous materiels but will probably have nc messursbie affect on
the potential severity of design pasis asccidents, Tﬁe hzzard measure
of the actinide wastc based on irgestion was incressed by only 1294
after 60 recycles. The total is sbout cme-ienth of that for the
fission products. If the hazerd measure is based on inhalsaticn,
recycling increases the potential hezaxd by 2 factor of 2 ot dis-
charge with the everage in the reasctor being significantly higher.
The actinides have an irhalaticn nazard measure of & facior of 7
higher ti:an the fission prodecus at discharge. The above stalemenis
assume that the reccncentration factors in the envirconment are sdproxi-
nately the same for actinides and fission producis. ?Present informsiiom,
however, indicates that certain tission products (e.z., S azd I) gre
reconcentrated %o a greater extent in the epvironment. This hHxx the
effect of causing the fission products to be the dexinant source of
both ingestion and inhalation hkazard during reactor operation. The
actinid= ccncentratioz in & rescicr, however, is not significant in
aralyzirg the "maximum credible sccident” {MCA) since the sctinide
compounds cannot be sigrnificently dispersed into the stnmeiphere by
any credidble resctor accident. Transmutation of lission products
in burner reactors would, of course, edd tc the notentisl hazard
of the MCA because the welatiie fission products are controlling in

an accident analysis.
Recycle of setinider in the LiIF3Rs should preduce even higher
nazard veductinn fattors since the sverage fission-to-capture ratio
af She artinides should be nigher in 2 Tast remcior than in 52 thermal
one. The author her foumd it difficplt o suantify this effect bhocmuse
of the current paucity of neutron cross-reciion date for the higher
actintdes in Tail soectre. FTast oross-section data for the higher

actinide; should de developed 32 that recycling siudies can 5e made

it aizs appears that recycling of the actinides is particulerly
suited Tor & Fluid Suel reactor such &z thz HSIR. & processing schexne
Za3 beern visuvalirzed thal recycler efsentinliy all the urasnium, neptunium.
thoariun, and most of the other actinides. Considersbly less americium
and curiyx sre produced cospared to & PR, Wwnich consideradly zisgplifie:z
the wesle cisposal prchien. In addition, being a fluid fuel resctor,
the prodlens srising Iroo Fabrication snd handiing of kesvy neutron.
exitting fuel elemernir are eliminated,
3. METHCD FOR DETERMINING THE HAZART OF RADIOACTIVE WASTE

In cgqpariﬁg the potential hezerd from different mixtures of radic-
sctive materials, a stendard method is required for determining a specific
value for each mixture that is s reflection of its biological hazerd. The
specific eetivity alone is insufficient since bicleogicszl factors are not
included.

The controlling consideration of hazard from the viewpoint of lonz-
term storage or disposal of radioactive materials is the danger of their
dissolution or dispersal irn unéerground waiter wilh subsequent ingestion
by humen beings. Cansecuently, a good measure of the ingesticn hazazd
associated with a xmixture of radionuclides of widely varying sctivities
is the quantity of water required to Gilute the radioactive mixture to a
concentration low encugh to permit unresicicied use of the water: the
larger the amount of water requirad, the greatsr the potentisl hazard.
The hazard measurs for the mixture is determimed by summing the amount
ol wnter recuired to dilute each individual muelide to its Radiation
Concentration Guide velue {or RCG, which was formerly calleq MPC) for
wrstrictes use of water., This methnd, which was used in & previous
unrks on the hezards of long-term storage of radicactive wastes, was
selected for use in the study. The method has the virtue ¢of =implicity
in soplicaticor and relates to the maximum veluve oo the hazerd since no
consideration i given to fractionation and paths of travel to human
beings. e most rxecert digiussion ofF cther methods of evaluafing the
hatard poteniinl of radicactive weste is given by Gera and Jacobs,7
who also propose & new hazard mesasure that invoives both the ingestion
hazaré uzod in this study, the inhsjation hazard, andg the prohability of
beling taxen up by humans. Determination of these probabilities is very
Sifficuit, however, 3ince statistical data regerding the prcbability of
gezidents ané other radicactivity releases, Inrluding their conseguences
in 21} phases of radiocactive waste manasgement, are not readily available
or orgily estimated,

The RUGz vsed In this study were taken frem the Code of Federal

Regalntian:,s which iy currently the guide for unrestricted use of
12

water in which these ruclides mey be dissclved. For nuclides with
unliéteg RCGs, the recommended defeult values were used, nemeiy,

3 x 107" uCi/ml for beta-decay ruclides with half-lives grester than

2 ar and 3 x 107 Cifm> for miclides that Gecay by elpha emission or
sponteneous fission, These default values represent a conservative
estimate of the RCGs, Some of the results in this report are alse
compared on %the basis of the KCGs used by Bell and Bﬁ.lcn:a6 and those
recently calculated by IaVernme’ feor unlisted muclides. Bell and Dillen
used € x 1077 and 2 x 10'6 Ci/m3 for 2°7Re and 229m,reSpectively,an&
unity for all other unlisted nuclides. LaVerne calculated RCGs for all
the uniisted nuclides and 5 x 1077 and b x 10™! Ci/m> for >°?Ra gnd
2291"‘:3, respectively, the two muclider thet contributed to most of the
differences that occurred dque 4o the particuliar RCGs that were uced.
13

i

b,

TUCLEAR CALCULATIONAL METHOD

The puclear calculations dwring reactor irradiation and after
discharge were made with 2 modified version of the nuclide generation
and devleticon code SRIGZN.’ The calculation daring irradiation is based
on three neutron energy groudps, namely, thermal, a 1/t energy distribu-
tior in the resconance region, and a fast group. The crozs sections in
the libzrary had been predetermined from basic data vy weighiing with a
typical P¥R neutron energy spectrum. Xore details of the original code
and cross-section library, wiich included data for actinides only up to

254
Cni, are given in refs, 10 mnd 11.

For calenlaticns involviag recyeling of the actinides, it was
necessary to erxpsnd the iibrary te include some higher transuranics and
inerease the calcnlaticual sccpe of the ORIGEN code. Cross-section and
decay data for the following muciides were added to the PWR sctinide
Library: ahou’ 2hamﬂp,‘ahoﬁp: ahk?u’ 2hs?u: 2h5am, ah5Cm, 2h6cE’ 2&7Cm¢
2hecﬂ’ ehgcm, 35°cm, 2‘;Bk, 2503&; 2&9Cfr esonJ ESICf: 252cr, 253cr,

25h6£, 35333. actinides higher then einsteinium were not expected to

heve a significant effect because they all decay (X-deczy, along with 2
little spontaneous fission) with short nelf-lives, thus preventing buildup
of the nuclides beyond 2”7Zs. The calculations confirmed this expectation.
The decay michod and neutor irterastion provabilities are such that no
significant amounts af the actinides can be remcived from the reaction-
ﬁécay chain except by fission. Cross sections end decsy constants for

the transuranic elements thzt vere adled to the library were taken fronm
ref, 12, '

.
A calculﬁ?icn af the meteriel maltiplicaticn constant or k_ was
added to the code since it was necessary to know the effect of actinide
recycle on the veactiviiy. Althouvgh the kﬂ aalculation igrores core
i=akege and control rods or other control poisons, the results, which
wouid not be adeguate for ithe core physics, seer adeguate for reiative
comparisons. Neutron yields per {ission as a function ol energy wers
saken from the SNDF/B-IT dets file’> for most of the Tissile muclides.
for those not incliuded ir that fle, the neutron yielids were taken or

infervred Irom tne pudblications by Gordeeve and :‘:‘ﬁirenkin_.l4 Hopikins
14

‘ me 1 . 16 Lo s _ . o o

and' Diver, 2 and Clark. The effective neutron yield from fission of
each ?uclide hy resonance energy neultrons was obiained by weighting the
energy dependent yieids with a 1/E neutron flux., For fast fissions,

the figsion spectrum was used &8s the weighting ITunrction.

Otker code chenges include a recycle option for any number of
actinides, en ebility to specify removal of any number of actinides
afier an arbitrary decay time subseguent to reactor discharge for
recycling or further decay of the remaining materials, ard ar ability
to account for the fissions of all the fissionable materials.
15

5. REACTOR TYPE AND STANDARD CCNDITIONS

The reactor selected for this study was the Diablo Canyon. which is
typical of & PWR design. When operating at equilibrium, the fuel is 3.3%
enriched uranium with a burnup of 33,000 MWd/metric ton of uranium. It
was assumed that this burnup was obtained by continuous coperation at e
specific power of 30 MW/metric ton cver a three-year pericd. Tor the
usually assumed plent factor of 0.8, intermittent operation at a specific
power of 38 MW/metric ton for 80% of the time would produce the same
burrup. Sinee {for the time periods involved), the waste hazard measure
resulting from a particular burnmup is not a sensitive funetion of any
-reascnable cperaticn schedule, it was deemed unnecessary to complicate
the cealculeations and analysis by considering a perticular operation
zchedule.

The fuel regicn is divided intc thres zones with each ons conteining
about an equal weight of fuel (approximately 28.3 metric tons of uranium).
The central zone is discherged yesrly ané the remaining fuel shuifled

inward with the cuter zone being recharged with fresh fuel.

In the caleulations it was necessary to ignore control rods and to
assume that the neutron flux wes woifcrm throughout a region, ana that
the regions were neutronically unccupled. A calculstion cycie comprised
three years of irradiation time between charge and discharge of a zone.
This procedure gives the correct values (within the accuracy of the
assumptions) for the discharge compositicn after the irradiation cycle.
The average coxposition, neuntron fiux, and km for the entire reactor
loading =Ty over one-vear cycles hecause of the yesriy charge and
éischarge and are noct expiicitly given in the output of the CRIGEN code.
However, these sveresge values cer be constructed easily from the ouiput
of & calculaticn cycie.

The “standard” Tor compsring the effect of actinide recycie on the
actinide weste hazard measure was the waste obitined by removing a stiru-
lated percentege of uranium amd plutonium at 150 days after discharge
i sending the remaining quantities to waste along with sl11 the other
ectinides, and all actinicde daughters generated since discharge from

the reactor.
16

6. CONTRIBUTION OF EACH COMPONENT TO THE HAZARD OF THE
WASTE FROM A PWR SPENTFUEL PROCESSING PLANT

The results of the calculations presented in this section show the
relative importance of the contribution that the varicus corponents make
to the hazard of the waste from z PWR spent-fuel processing plant for
the previcusly described standard conditions and 99.5% recovery of
uranium apd plutonium; i.e., 0.5% of U and Pu and 100% of ali other
components are discharged ac waste and stored some place after suitable
processing.

Table 3 shows the percentage contribution of the actinides (inc¢lnding
their decay pro&acts) to the total hazard measure (water regquired for
dilution of the content of one metric ton to the RCC for the mixture) of
the waste as determined with three different sets of RCGs and the effect

of removing 1291

from the waste. Beyond sbout 00 years, the actinides
and their deughters dominate from a hazard measure viewpoint and show no
significant effect up to about 1()ll years due tc the different sets of
RCGs. At greater times, the relative importence of the actinides dimin-
ishes somewhat when the RCGs of rcf. 6 or ref. 9 (see Appendix I) are
used for the unlisted nuclides in place of the recommended derault values
in the Federal Code of Regulstions. Most of this difference can be
atfributed to the difference in RCGs for the nuclides of the 233U decay
chain (4n+l series), particularly those for 2297 and “%pa.

Toe remaining contribution to the hazard measure is almost all
(strustural clements are not important) from fission products with 1221
supplying 88% of this total at 103 years and rising to 98.4% at 106 vears.

Essentialiy all of the remaining fission product hazard for the longer
By 9 Py s 135
17 2 < WS e

times is contributed by the ~No and -
The yelative contributions of each actinide and itz daughter o

the total hazard memsure resulting from the mixture of the actinides
and their davghters are given in Table L, which shows that wp to J.O1¥
years the actinide waste hazard is mostly controlled by the americium
and curiuvm with no significant differences resulting from the different
RCGs. At mmch greater decasy times,the long-lived 2371\?;3 (2.1 x 106 year
17

Tzhle 3. Relative Contribution of Actinides apd Thelr Daughters
to the Total Waste from PWR Spent-Fuel Processing

 

Contribution of Actinides and Their Davghters (%)
at Decay Times (years) of:

iy =
102 & x10% 103 10" 10° 5 x 10° 10°

 

Using CFR RCGs and Recommended Default Values for Unlisted Nueclides:

1297 tpesent 0.3 9.3 97.5  93.8  97.8  90.2 99.1
129

I removed 0. 34 96.7 99.6 99.1 99.8 99. 9+ 99. G+
Using CFR RCGs and Values from Ref. & for Unlisted Nuclides:

129, present 0.3k gk, 3 97.5 92.5 59.3 70.8 61.6

1290 removed 0.3 96.7 . 99.6 58.8 55.7  9B.4 99.0

Using CFR RCGs and Vaelues from Ref. § for Unlisted Nuclides:
129I

present C.34 ok, 3 97.5 92.6 72.9 78.L 73.8
129

T removed 0.3k 96.7 99.6 g8.9 96.4 98.9 09, 4

 
 

 

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half-life) and its daughters begin to dominate. Another important point
is that the remaining actinides along with their deughters, namely, Ac,
Th, Pa, Bk, Cf, and Bs, meke a negligible coniribution to the waste
hazard. The contribution of uranium to the hazard of the U + Fu mixture
alone varied from negligible to a maximum of 25% at 106 years. The import
of these results is that in any waste menagement system in which at least
939.5% of the uranium and plutonium is extracted, a significant reduction
ir the actinide waste hazard can only be obtained by removel of most of
the americium, curium, and neptunium from the waste. If 99.5% removal of
these three actinides is a2lso effected, the uranium and plutcnium become
controlling and it would then pay (from 2 waste hazard viewpoint) to
increase the extraction efficiency of these latter elements, particularly

the plutonium.

The absolute values of the contribution of each cocmponent to the
nazard measuvre in cubic meters of water per metric ton of fuel are shown
in Tabie 5. To put these values in perspective, consider the required
2.3 x 1039 m3/metric ton for dilution of all the muclides to the RCG
after decaying 100 years. This volume of water is approximately equal
to the yearly flow of the Mississippi River into the Guif of Mexico.

Note that the last two rows in Table 5 are based cn the RCGs given in
refs. 6 and 9, respectively, for nuclides unlisted in the Code of Federal
Regulations, which (for beyond 10k years) results in an increasingly
smaller hazard measure that is about a factor of 67 and 37 lower, respec-

tively, at 106 years.

The apparent large quantity of water regquired for dilution to the
RCS for just one ton of fuel tends to megnify the potential hazard. Wo
reasconable scengrio can be constructed that visuelizes repid mixing or
dissolution of waste that has been processed into a very slightly soluble
form. The ingestiﬁn hazard measure refers to potential long-term solution-
ing. However, consideration of such quantities of water dces present ore
argument for decreasing thé”quantity of actinides for ultimate disposal
by recycling the actinides back %hrough the power reactcrs prodncing'
ther. On the other hand, Bell and Dillon6 (using their RCGs) point out

that, after aging 1000 years, the actinide hazard measure of waste stored
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- -
At &

&I
5':

guantity of uraniwm cre and tailings ogual to the smcunt of salt and

stizle associated with the waste from one meiric ton of fuel from a ?%R:
rurthermore, they show thet il the salt ded iz dissolved some thousands

¥ years in the future with sufficient water o dilute the radicnuclider to
their P05s, the water womlé be unaccepisdle 25 potable water decause of

,

its sodium chloride content relher than i%s radisactiviiy

A1l of this discussion indicates a reed for siandardizing the values
TR
for the RCGs that are not listed, particularly those in the ~-“U decay
c¢hain (bn+1 series), so that 3 better evalustion of the hazards of very

long-term storsge cen be made.

Table 6 gives the activity in curies per metric ton of fuel in the
wacte siream for each actinide and fts daughters.

Table 7 skows the eoffect of neglecting the actinides higher than
2hka on the hazard measure of the waste. When the higher actiniiles are
ircluded, the hnzard measure of the waste increases slowly up o a mexi-
mim factor of near R at a little over loh y=ars compared to that osteinec
vhen they are .eglected. Note, aowever, there is very little difference
in the values for the activity messured in curies.
 

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As previousiy mentioned, the concept of burning the “problem fizsien
. 85 : I37~ . N : -
sroduets” JKr, ©~Sr, and ~2!Cs in nuciefr reactors has been studied by
1,2

Steirherg and co-workers.” ’” In this section, their work is discussesd

3,4

briefly aleng with the gasted use of controlled thermonuclesr
Y ‘g SUEK

reacLors.

™he prodiem fissien products cannot be elimineted by any system cf
Tission power resctors operating in either a stagnaut oy expanding nﬁclear
pawer eccnhomy since the production rate exceeds the elimination rate by
burnout and decay. Only at eguilibrium will the production and removal
retes be equal, # condition that is never attained irn power reactors.
Equilibrium can be obtained, however, for a system that includes the
stockpile of fission nroducts as part of the system inventory since the
stockpile will grow until its decay rate ecuals the net production rate
of the system. Far the projected muclesr power eccnomy, however, this
will reaquire a very large stockpile with its associafed potential for
release of large quantities of hezardous radioisotopes to the environ-
ment. It is this stockpile that must be greatly reduced or eliminated
from the biosphere. A method suggesfed by Steinberg et g}_ is transmi-
tation in "burner reactors,” which are designed to maximize neutron
absorption in separated fission products charged to a2 reactor. If
sufficient rumbers of these burners are used, the fission product
inventory of a nuclear power systiem can Lhen reach equilibrivm and be
maintained at an irreducible minimur, which is the guantity contzined
in the reactors, the chemical processing plants, the transportation
system, and in some industrisl plents.

Burning fission products in the blanket of a fusion reactor with
the excess neutrons that are produced is,in theory,an excellent rethod
since no fission products would be produced, Considerable tritium will
be produced, of course, butl this presents a much less severe dispcsal
problem.
25

{oviously the use cf burner reactoz's' ‘or fusion rescters in the
system will increase ihe cost of ouclear power and reduce potential
breeding capsnity dut transmifetion is certainly one of two known
metbods (the other Yeing dispossl in spece) of eliminating most of
these bazardous materiasls witkh po possibility of return to the dbio-

sphere,

7.1 Haddmeoww Buememtoto-Production Ra_tios for Fission Products

I$ the asswmpiicn is made thatl burner reacteors are g desirable
adjunct to a nucleer economy, what are the design requiremenis and
limitations? It is obvicus that they must maximize (with due regard
to ecoromics) the ratic of burncut of a particular fission prodgwet to
its producticn rate in fission reactorg,end the neutron flux must be
high enough to cause a significsnt decrease in ile effective half-life.
Of the fission types, the breeder reector has the most efficient neutron
eccnomy and in principle would make the most efficient burner if a1l or
part of the fertile material can be replaced by & Sr-Cs mixture without
causing chemical processing problems or too large 2 perturbation in the
flux spectrum beceuse of the different characteristics of these fission
products. The cost accounting in such a system would seb the value cf
nzutrons anscrbed in the fission product feed at an accounting cost equal
to the value of the fuel bred from those meutrons.

The maximm possible burnout of fission products would occur when
the excess neutrons per fission that would be absorbed in a fertile
materisl are absorbed instead in the fission product feed. The largest
possible burnout ratio would then be the breeding ratio {or conversica
ratio for non-breeders) divided by the fission product yield. The esti-
mated breeding ratiec for the Molten Salt Bresder Reactor (MSER), a thermal
breeder, is 1.05 and for the Liquid Metal Fueled Fast Breeder Reector
(IMFER), 1.38. The yield of 13/0s + PO%r is 0.12 atom/fission, but a
nurber of other isotopes of these elements are produced which would alsc
absorb neuntrons. Hewever, if the fission product waste is aged two years

tefore separation of the cesium and strontium, the mixbure will essentially
26

be composed of about 80% 1705 + 903‘: and 20% 135 {anich will capture
neutrcons to form *3 Cs that decays with e 13.-day half-life)}; conseguently
the maximm burnout ratio for To'Cs + Z0Sr will be decreased oy 204,

This leads to e maximum possidble burnout ratio of about 7 for the MSBR
and about 9 for the LMFBR. Unfortunetely, however, the neutron fluxes
in these designs are well belov 10°C nfcn’-sec. Any modifications of
these desigrs to create high neutron fluxes will increase the neutron
leakage and decrease the burnout ratics significantly.

7.2 Reactor Hesidence Times Requiresd for Fission Producet Burnmout

Table 5 was prepared to iliusirate the effect of neutron flux on
the residence times (which affect recycle costs) required for burnout
ané decay of 99.9% of the importent nuclides using s burner resctor with
the neutron spectrum similsr to thet of & typical light water power
reactor. 1% is apparent that the efficiency of turnout incresses with
increases in neutron flux, c¢ross sections,ard haif-life. With the excep-
tion of 1391, which is not nearly as large a probliem 8s tne others and
can probably be essentially disposed of by isotopic diluticn,a the times
shown in Table 8 inficate that neutron flux levels are reguired which
are much higher than those that have been attained :m present nuclear
reactors (~ 5 x 10 5} and that fluxes near 1017 n/ cm -Sec¢ are probadly
necessary before serious consideration could be given to burner .eactors.

In a conceptual design study by Steinberg et al. » it was concluded
that the quantities of 137(‘.'5 9081- , and 851(1- scheduled for permanent stor-
age in the projected nuclear economy could be reduced by a factor of 1000
by burner reactors cperating with neutron fluxes up to 1016 ny cmz- sec
for added costs of 0.63, 0.2k, and 0.021 mill/kWhr(e) respectively. The

]
905:9 end *3TCs in such a system, along

estimated costs for burning out
with the probable escalation ir an actusal design study that includes
directly the costs of transfer between plants, canning the fission

products, additional chemicel seperations, various temporery storage
facilities. and reactor residence times seem to preclude use of this

method. A cost of 0.021 mill/kWhr(e) for burning SKe seems sufficiently
7

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28

low for comsideraticn on an economic basis but the neutron abscrption
cross-section of 85Kr was takep as 15b, 2 value now known to be low by
around sn order of magnitude. Recent work by Bemis 33_3;.17 gives a
Maxwellian-sveraged thermal value of 1.56 b and & resonance integral of
1.8 b, A reevaluation of the 85Kr removal system using the lower cross

section would increase the cost to an uneconomic level.

7.3 Applicaticn of Transmutation Schemes

- 8 . :
Nichels and Blomekel have made estimates of the effect of wvsrious
schemes of uneutron-induced tra-smutation on.the potemtial inventory of radio-

S0

jisotopes and costs of electric power (Table 9). The isotope © Sr wes
used as an exemple because it is the prime contributor to the radio-
logicel hazard of spent fuel and does not reguire the use of isctopic
895r — half-life 50 days)

before recycle to e burner reactor. Their analysis of the use of each

separations (other than providing for decay of
reactor system shown in Table 9 are given in following secticns,

T.3.1 Pressurized Water Reactors

Rows 1 and 2 of Tepvle ¢ illustrate that the effect of recycling cf'
9OSr within a system of light water reactors is to cause essentially no
change in the total quantity of 9OSr thaet is asscociated with the system
gince the rate of neutron-induced transmitaticn is small as compared
with the rate of decay. Under current policies and plans,most of the
9OSr assoclated with the system would bes stored at a federal waste
repository. In the recycling system most of the inventory would be in
reactors while the remainder {~ 25% of the total) would be in canals for

postirradiation decay, reprocessing plants, and fuel fabrication plants.

This example illustrates a primary disadvantage of systems for re-
cycle and neutron transmutation of fission prcduct nuclides. These
schemes have the common characteristic that larger quantities of radio-
active nuclides are being actively handled arnd processed than if the
rpelides were stored. Conseguently, larger quantities of these nuclides
occur in a dispersible form and are associated with potentially large

sources of energy that could provide a mechanism for dispersal.
29

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30

The additional cost of recycling 9OSr to light water power reactors
was estimated roughly as 0,1 mill/kWhr(e). The primary source of this
cost is an approximately 25% increase in the unit cost of reprocessing
and fuel fabrication. This reflects the increased separations and product
handling operations that would be required st the reprocessing plants and
the regquirements for shielding and remcte cperation of the fuel fabrica-
tion plants. The estimated costs cf high-level waste disposal weuld be
decreased from about C.05 to abeout 0.04 mill/kWhr(e), however.

7.3.2 Liguid Metal Tast Breeder Reactor

The effscis of recycle of 998r in IMFBRs (rows 3 and 4) are essen-
tizlly the same as those for a system of light water reactors. Inven-
tories of 9OSr are Jlower, howeveQ; because of the lower yield of 905r

from fission of pluteonium.

7.3.3 High Fivx Isotome Reactor

The effect of recycle of 90Sr to a system of High Flux Isotope
Reactors is shown to indicate the relative change in inventory that
would result from the use of the meximum thermal neutron flux levels
that are available in present reactors («-fo 1015’h/cm?‘sec with
targets in place). ZFven with these high éélues of flux the effect of
recycling is to decresase the inventory assoéiated with the system by
only about 6C%.

This type of reactor would not be an economical source of electric

power, however, because of its small size, high refueling cost, and high

7.3.k Tusion Reactors

A proposal by Steiner3

involves using the excess neutroas from
fusicn reactors. which in theory will provide a chegp and sbundant
source of neutrons and has the advantage of not producing any long-
lived fission products. Consideratle tritium will be produced, of

course, but this presents z much less severe disposal prcblem. Steiner
estimates, on the basis of calculated tritium breeding ratios and antici-
pated tritium doubling times, a neutron excess of 20% end a thermal
neutron flux aveileble for burnout of 3 x 1016. On this basis, & recycle
system from which 984 of the power is generated in IMFBRs and 2% is gen-
erated in & fusion burner reactor would have an order-of-masgnitude lower

903r inventory than a system of IMFERs,

In & recent paper by Wblkenhauer,h some aspects of the problems of
burning fission products in controlled thermonuclear reactors were con-
sidered in more detail. He concluded that if a D-T reactor with a tritium
breeding ratio of 1.2 is used t¢ fTransmaite the totsl 13703 and 908r from
a nuclear power economy, 5% of the generating capacity would have come
from CTR plants. Only 1% of the generation capacity would be reguired
if D-D reactors were used. Using the worth of neutrons for the produc-
tion of fissile plutonivm as & basis, it was estimated that the cost of

transmiing 137Cs and 20

Sr would be at least 10 times as sxpensive as
the estimated cost of storage of all fission products in deep salt

formations.

Regardless of any potential merits ¢ using controlled thermonuclear
reactors to burn fission products, such systems cannot he seriouély con-
sidered at present since it is generally Telt that the practical fusion
reactor is stili 30 years in the future.

7.3.5 Spallation Reactor

In an effort to devise a system with both a high neutron flux and a
high burnout ratio, Gregory and Steinberg2 have suggested the use of a
sPéllation reactor. A typical spallstion burner reactor would use a
1000-MW(e) nuclear reactor Lo power a nigh-energy accelerator; the
accelerator, in turn, would produce a 500-MW beam of 10-BeV protons,

a neutron source of greater than 1020

neutrons/sec in a liguid uranium
target, and a thermal flux of about 2 x 10t n/er-seec in en array of
Dzo-moderated 9oSr targets. This apprcach would require extensive
development including, in particular, a method for copirz with the

potentially severe radiation damage and heat transfer problems.
32

In this system one spallation reactor of capacity 1000 MW(e) would
be associated with each 900C MW(e) of power produced by IMFBRs. The
cost penalty would be spproximately C.8 mill/kﬁhr(e), primarily associ-
ated with the capital and operating costs of the spsllation reactor that

does not produce electricity for sale.
33

8. ACTINIDE RECYCLING IN A PWR

In this secticn the actinide recycling calculations made wiik: the
medified ORIGEN code are discussed and the pertinent results given. In
sddition the chemical processing requirements are discussed in general
from & viewpoint of simplifying waste management and, more specifically,
as applied to actinide recycling.

8,1 TFlowsheet

The generszal flowsheet assumed for these actinide recycling calcu-
lations is shown in Fig. 1. TFor the calculations, it vas assumed that
chemical processing of the spent fuel occurred instantanecusly at 150
days after discharge from the reactor. For simplificetion, the actinides
recycled were taken as those present at that time., In any actual recycling
scheme, the recycle material would spend more time out of the reactor,
However, becanse of the relatively long half-lives of the actinides
anlthe very small buildup of daughters in any resscnable time between
processing znd recycle, no significant differences in the results would
occor for holdup times of a factor of 2 or so longer.

In addition, it is quite possible that any recycling scheme would
include transursnium wastes {(from sweepings, sludge, scray metal, filters,

ion exckange resins, etc.) produced in the nuclear industry. Although such
westes would be produced in large quantities, recycling them would not

cause significant difference in the results and conclusions of this
study.

Frar a caleculational standpoint, the method of including the recycle
material with the fuel is immaterial since the calculations mist assume
homogeneity. The recycle material for each recycle caleunlation was merely
considered as a uniform sddition to the normal loading of 3.3% enriched
uranium fuel.
ORNL DWG 72-1436

[—3.3% ENRICHED U FEED

 

 

 

FUEL | FEED + RECYCLED

 

 

 

 

 

 

 

 

 

 

 

 

 

ASSEMBLY & PWR
FABRICATION | ACTINIDES
? | REACTOR
' DISCHARGE
I 99.5 TO 99.9% ACTINIDES
| OTHER THAN U AND Py
| CHEMICAL
PROCESSING
|_FISSiON PRODUCTS -
¢.1 1O 0.5% ACTINIDES
| 1009% DAUGHTERS 99.5 TO 99.9%
U+ Pu
WASTE U+ Py
STORAGE STORAGE

 

 

 

 

Fig. 1. Flowsheet for Actinide Recyeling.
35

8.2 Chemical Processing for Waste Management Simplification

In any waste management systen, and.particﬁlarly if recyeling of
the actinides is practiced, chemical processing plays a key role. The
importance is such that the previously unpublished comments by Blomeke
19

and Leuze ” on separation of radiocactive materials into selected frac-

tions for improving waste management are given in this section.

One possiblé wey To simplify management of waste from nuclear fuel
reprocessing is to separate the radicactive materials into fractions
based upon the time they must be stored before release to the environ-
ment is allowed. An indicaticn of the magnitude of the problem when the
soluticn of the high-level waste management problem is gpproached by this
path can be gained from Tzble 10. The fission-prcduct and actinide ele-
ments of greatest concern are listed, and the reguired degree of sepzra-
tion of each from high-level waste is given for various times of decay.
Affer the bulk waste bas been stored for 10 years, 12 fission-product
elements and 11 actinides (constituting zbout 15 kg/ton of fuel charged
to the reactor) must be separated from the remsining fission products
(~ 20 kg/ton of fuel) and process reagent chemicals by factors ranging
from b (for ectinium) to 2 x 10+ (for strontium). The residuals would
then be of a nature that wonld permit their release under the present

Radiation Concentretion Guides (one-third of the values given in ref. 8).

The separated Tission products and actinides should ideally be
further separated from esch other, based on thelr retes of decay, into
at least three groups. The first group would contain Ru, Sb, Ce, Pm,
and possibly H, and would require containment for several decades
{¢ 100 years). It is conceivable that this group could be retained
on-site for this periocd of ftine.

The second group would be composed of those fission prodicts
requiring storage of the order of 1000 years, i.e., Sr, Bu, and
rossibly Sm. The remsining fission products with very long half-lives
are only feebly radioactive, and it m2y be reascnable 1o combine them
with The second greup for 1000-year storege, or zlse they could be

separated and stored or recyecled with the actinides.
36

Table 10. Decontamination Factors Required to Reduce Constituents
in Liquid West? to RCG Levels® at Various Times of Decay?

 

Time Following Reactor Discharsze (years)
Element 10 100 1000 10"

 

Pigaion Products

Ru % ¥ 107 - - -

Sb 6 x 10° - - _

Ce 1 x 107 - - -

Pn 4 x 107 - - -

H 1x 102 800 - -

Sr 2 x 1082 gz x 1089 _ -

Fu 2 % 108 y x 108 - -

3m 3x10§ 1 x 109 1x 105 .

Zr zx 10  2x 10} 2 x 103 2 x 103
Te 7 x 107 7 x 10 7 x 10 7 x 10
I 6x10°  6x 102 6 x 107 6 x 102
Cs 5x109  5x10° 3 x 10 3 x 100

Heavy Elfments

Ra 85 93 2 x 103 1 x 107
Ac L 12 13 20
™ 100 100 | 300 5 x 109
P 3 x 103 3 x 103 b x 109 b x 103
U 300 1 x 103 2 x 303 2 x 10
Np 3 x 107 3 x 102 3 x 102 2 x 107
Pu 2 % 107 1 x 100 2 % 100 1 x 100
Am b x 16!  bx 107 1 x 107 leog
Cm 3 x 10% 2 x 107 1x 10 6 x 10
Bl 120 - - -

cf 3 x 103 30 5 -

 

 

aThe Radiation Concentration Guide values are one-third of those given in
the GFRB (RCGs of ref. 6 were used for the unlisied nuclides), and should
result in s radiation dose to the general public of less than 170 mrem/year,
(These values, however, are based upcn ingestion of the ligquid effluents
and do not allow for reconcentration in the environment.)

bwaste is generated in reprccessing spent PWR fuel initially enriched to
3.3% 235U, and exposed to 33,000 MWd/metric ton at 30 MW/ton. The waste
consists of all the nongaseous fission products plus the actinides remain-
ing after removal of 99.5% cf the uranium and plutoniuvm following a post-
irradiation decay period of 150 days.
All of the actinides, except Bx and Cf, require containment for a
period greater than 10,000 years; hence, they would comprise a third
group. It is reasonable that Bk and Cf sbould be relegated to this
group since they would contribute insigrificantly to the bulk.

Practionation of waste inte such groups for waste monagement entails
a number of difficult chemical separations. A severe problem is caused
by hydrolysis of several materials tc form colloids esnd precipitates.
When these are present, it iz virtually impossible to !obta.in the neces-
sary separations. In most cases, the extremely high decontamination
factors reguired have never been demonstizted., Separation of the
trivaient actinides (americium and curium) must be mede from kilogram
quantities of the lanthanides, and the long-lived lanthanides (europium
and samarium) must be separated from the other lanthenides. Thede eie-
nents have very similar chemical behavior, and separations musﬂ be mede
by chromatographic ion exchange which requires close process control.
There is no practicel process avellable for removing tritium from large
volures of agueous waste, and iodine removal with decontaminastion factors
of 10° will be difficult. However, processes now 1mder develogpment for
IMFER fuel should make it possible to remove these meaterizls tefore and
during feed adjustment for reprocessing of the fuel.

Difficuwiiizs have been encountered duringz chemical separations with
the hydrolysis ¢f plutonium, thorium, protactinium, znd zirconium to form
colloidal material that does not behave well in s=parations processes.
Mozt of the plutonium in agueous waste from the first Purex cycle is in
#h inextractable form. Even exhaustive extraction will not remcve this
plutonium unless some treatment can be developed to convert it o a
soluble, icnic species. Experience has shown that when significant
emounts of zirconium are present, it often hydrolyzes to form colloids
or precipitates, or both, which carry polyvalent iuns such as americium
and curium, This greatly complicates the separations problems and makes
1t virtually impossible to remove quantitatively the zlrcopium and asso-

cizted ions from a waste stream.

Although separations processes have been developed for essentially
all of the heavy elements and fission products, these processes are not
38

directly applicable to the problem of quantitatively isolating tnese
materials into compact fracticns for waste menagement. The existing
processes were developed for the purpose of recovering significent
quantities of a particular element, and recoveries of 90 to 90% were
considered to be satisfactery. Furthermore, these processes usually
result in an ircrease in contaminated waste instead of a decrease, since
process chemicals required sre discharged into waste streams along with
significent amountis of contamination. Modification of these processes
to give decontamination factors of 108 to 10%0 without creating lasrger
volumes of waste from contaminated chemiggi*;;;é;ﬁéélﬁzil require a
major development program. Since the overall fission preduct decon-
tamination factors usuwaliy attained over a single Purex cycle are only
about lOu, it cannot be expected that fully developed processes for
waste fractionation, even for elements that are well behaved chemically,
will give larger decontamination factors. Thus two, three, or even four

8

cyzles will be reguired to give overall decontamination factors of 10

to 1010.

Unfortunately, the optimum grouping of radicactive elements for
waste mansgement does not correspond with natural groupings based upon
chemical behsvior. Processes for removing americium smd curium from
the waste stiream will also remove all of the lanthanides and yttrium
(~ 11 kg/ton of fuel) with compsrable decontamination factors. About
10 kg/ton of fuel of these are either nonradiocactive or have short
enough half-lives so they can be released after less than 100 years
storage if they are adequately decontaminated from Eu, Sm, Am, and Cm
(see Table 10); end the Eu and Sm mist be stored about 1000 years if
they are adecuately decontaminated from sz and Cm. Thus, the separatiorn
of Am, Cm, Sm, snd Zu from the other waste products and into groups for
ease in waste management entails a considerable number of process steps,
each reguiring close process control because of the chemical similarity

of these elements and veryv large decontaminetion factors required.

In summery, it can be concluded that the greatest contribution to
weste menagement through chemical separstions lies in separating the

actinide elements from zil of the fission products for either storage
39

or recycling, If not recycled, these elements require virtually permanent
containment and this could pfobably be acccmplished with greater ease in the
absence of the heat-generating fission products. A guantitative assessment
of the reduction in hazard achievable “rom actinide separations in excess

of those considered in the body of this report is presented in Appendix IIT.
There it is shown tnat, 1if separations processes can be developed to yield
an overall recovery of 99.999% of the uranium, 99.995% of the plutonium,
$9.95% of the neptunium, and 99.9% of the americium and curium, the residual
wastes would have zbout the same ingestion hazard es naturally occurring

radioactive minerals after only a few hundred years.
8.3 Effect of Recycle on Reactivity and Flux

Tne average material k {neutron multiplication constant) or k, of the
recycled actinides is lower than that for a normal reactor loading, but not
much lower, as shown in Tabhles 11 and 12. Table 11 shows the effest of
recyeling of 95.5% of the actinides up to 60 times [equivalent to 180 years
cf reactor operation). The meximur average reactivity decrease is about
0.8% @nd is attained in about five cycles. This decrease can be counter-
acted by only about a 2% incresse in fissile material, which is not prohib-
itive since this can be accomplished by incereasing the enrichment of the
fuel from 3.3 to 3.4%. Similar results aré shown in Table 12 for recycling
of $9.9% of the actinides which, as co be expectied, causes a slightly grester
reactivity decrease. Table 13 shows that the effect of recycling of the
sctinides (for either 99,.9% or 99.5% to three significant figures) on the
thermel flux is sufficiently smell to be of no significance to the reactor
operation,

8.4 Effect of Recycling on Hazard Measure

The effect of recycling of 99.5% of the actinides other than U or Pu
on the hazard measure of the waste from PWR spent-fuel processing at 150
dzys after reasctor discharge is shown in Table 14 as a function of post-
irradiation decay time. Similar results are shown in Table 15 for 99.9%

extraction and recycling of the actinides.

The ratio of water reguired for dilution of the waste to the RCG
for the standard case (no removal of actinides other than 99.5% of Pu

+ U, or 99.9% if the ratio is determined for the higher extraction
Table 11,

ko

Actinides Cther Than U and Pu

Effect on Reactivity from Recycle of 99.5%

 

Reactivity (ko)

 

Reactivity Change (%)

 

 

Recycle
No. Start? End Start End Average
0 1.20145 1.09391 0 0 0
1 1.19252 1. G890k ~0.743 -0.4khs -0.594
2 1.19029  1.08785 -0.929  ~0.555 0.7z
3 1.18964 1.08754 -0.983 -0.583 -0.783
4 1.18945 1.087L& -0.999 ~5.590C 0. 79l
5 1.18040 1.08745 -1.003 -0. 591 ~0.797
10 1.18938 1. 08747 -1.005  -0.589 ~0.797
15 1.18937 = 1.08748 -1.005  -0.588 -0.797
20 1.18937 1.08748 ~1.005 -0.588 ~0.797
Lo 1.18937 1.08749 -1.005 -0.588 -0.797
60 1.18937 1.08749 -1.005 -0.588 -0.797

P a—

®At start of an irradistion peried 1/3 of core loading has been
in reactor for 2 years, 1/3 for 1 year, and the remainder is

new fuel.
Table 12.

Ll

ffect on Reactivity from Recycle of 99.9%
Actinides Other Then U and Pu

 

Reactivity (k)

 

Reactivity Change (%)

 

 

Recyele

No. Start® End Start End Average

0 1.201k5 1.09391 0 0 Q

1 1.19261 1.08902 -0.736 -0. Ll ~0.592

2 1.19023 1.08782 -0.93kL -0.557 -0.746

3 1.189s8 1.08750 -0.988 -0.586 -0.787

L 1.18939 1.087hL3 ~1.00k -0.593 -0.759
5 1.18933 1,087k -1.009  -0.59%4 -0.801
10 1.1893% 1,087k -1.010 -0.59z2 ~0.801
15 1.18931 1,08745 -1.0i1 -0.5%1 -0.801
20 1.18031 1. 08745 ~1,011 -0.591 -0.801
30 1.18931 1.087L45 ~1.01L -0.591

-0.801

 

%At start of an irrediation period 1/3 of core loading has been
in reactor for 2 years, 1/3 for 1 year, and the remainder is

new fuel.
Lo

Table 13. FEffect of Recycling®on Thermel Neutron
Flux in a Typicsi PwR

 

Thermal Neutron Flux x 1073 (n/cm®-sec

at Irradiation Times (days) of:

-

 

Recycle HNo. 110 07 550 733 1100 Average
o 2.58 2.64 2.51 3.03 345 2.92
1 2.57 Z.64 2.81 3.02 3.hk2 2.91
2 2.57 2.83 2.80 3.C1 3.0y 2.91
3 2.57 2.63 2.80 3.01 3.40 2.90
L 2,57 2.63 2.80 3.00 3.40 2.90
5 2,57 2.83 2.30 3.0. 3.%0 2.90

10 2.57 2.63 2.80 3.0u 3.40 2,90
20 2.57 2.63 z.79 3.00 L0 2.90
Lo 2.57 2.63 z.72 3.00 3.40 2.90
60 2,57 2.%3 2.79 3.0C 3.he 2.30

 

a : ~ .
Cne cycle represents 3 years of reactor oderation.
Teble ik. Effect of Recycle of 99.5% of Actinides Other Than U and Pu
on Hazerd Measure of Waste® from PWR Spent-Fuel Processing

 

ter Required for Dilution to RCG,° Ratio of Standard
to Recycle Case for Decay Times (years) of:

 

Recycle

 

No. 10 10° 103 10 5x 100 10° 107
o 0.k 12.3 15.3 18.5 22.8 27.9 52.3
1 22.5 9.30 124 13.h 16.0 19.7 45.7
2 19.3 8.20 10.0 10.8 14,5 18.0  43.6
3 17.5 7.57 8.43 9,29 k.2 17.k 42,8
L 16.5 7.15 7.35 8.25 14.0 17.1 42,5
5 15.8 6.77 6.57 T.53 1L.0 i7.0 42.5

10 13.k4 5.76 4,72 5.75 13.9 17.0 42,5

15 12.1 5.32 k.16 5.53 1.8 17.0 h2.5

20 1.4 5.08  3.78 4.89 13.8 17.0 42,5

25 11.0 4.95 3.63 4,73 13.8 i7.0 k2.5

30 10.7 4.89 3.56 4,63 13.6 17.0 k2.5

L0 10,5 4.83 3.k9 L.55 13.6 16.9 k2.5

50 10.3 4.80 3.46 4.39 13.6 15.9 k2.5

60 10.3 4,80 3.46 L.39 13.6 16.8 2.5

Eff., 4 25.5  39.0 22.6 23.7 59.6 0.2  81.5

 

20.5% Pu and U sent to waste.

JﬁI‘n‘. recommended default RCGs in The Code of Federa.:. Regulations
(ref. B) were used for unlisted nuclides.
ly

Table 15. Effect of Recycle of 99.9% of Actirides Other then U snd Py
on Fazerd Measurs of Waste® from PWR Spent-Fuel Processing

 

Weter Required for Dilution to RCG,° Retic of Standard
to Reecycle Case for Decay Times (years) of:

 

Recycle

 

No. 10 10° 103 10 5x10° 100 . 10°
0 195 57.5 73.1 83.8 110 137 256
1 116 3.7 58.9 64.2 77.7 95.9 22k
2 s4.8  38.L L7, 7 51.6 7C.8 €7.3 213
3 85.8 35.5 Lo.1 I 68.4 8.k 210
L 80.7 33.k 34.8 39.3 67.5 83.L 209
5 77.2  31.7 31.1 3£.8 67.2 83.2 208

10 65.7 27.0  22.1 27.0 648.¢ 82.7 =207

15 58,7 24%.7  19.1 23.7 66.4 82.3 206

20 4.5 - 17.6 22.2 66.1 8z.1 206

25 52.0 - 16.8 21.4 66.0 82.1 206

30 50.6 - 16.5 20.9 65.8 82.1 205

e, , 4 25.4 - 22.6 23.5 59.8 9.9 80.5

 

20.19 of all actinides sent to waste.

bThe recommended defeault RCGs in the Code of Federal Regulations
(ref. 8) were used for unlisted nuclides.
k5

percentage) to that required for dilution of the waste after each
successive recycle is en indication of the efficacy of recycling

from a potential hazard viewﬁoint. This ra2tio is defined as the hazard
reduction factor (the higher the ratio, the greater the hazard reduction).
Teble 16 shows a different method based on the activities in curies for

evaluating the efficacy of recycling.

Table 1& shows that the hazard reduction factors of the waste with
99.5% of the actinides extracted equilibrates at k2.5 for a decey time
of 106 years. vhen the RCGs of refs. 6 and 9 are used for the unlisted
nuclides in place of the recommended default values in the Code of Federal
Regulations, the hazard redvetion factors become 5.5 and 10 respectively.
The ccrrasponding velues for 99,9% extraction of the actinides are 28 and
49. Although it can be argued that the RCGs calculated by L&Vérneg are
more realistic than the more counservative recommended defanlt values,
the Code of Federal Regulations must be followed in nuclear reactor

design and operation.

Note that the last row in both Tables 1i and 15 show recycle effi-
ciencies at each decay time. These efficiencies represent the percentage
of the maximum possible hazard reduction facter that is attainasble after
effective equilibrium 1s reached in the recycling prccess. The maximum
possible hazard reduction factor is the ratic of water required for
dilution of the standerd waste with only 99.5% of U + Pu removed {or
99.9%) to that required for the same waste when 99,5% (or 99.9%) of all
actinides are extracted at 150 days after discharge from the reactor.
This is precisely what is contained in Tzbles 1L and 15 for zero recycle
or cne pass through the reactor. These are obviously the largest hazard
reduction factors obtainable at a specified decay time since they are
based on the removal of 99.5% {or 99.9%) of &ll the actinides rather than
just the U + Pu. Each additional recycle increases the hazard measure
of the discharged material in asymptotic fashion. Trhe steady-state
reéycle efficiencies shown are simply obtained by dividing the values
for 60 recycles (or 30 in Teble 15) by the values in the first row at
corresponding decay times. In a similar fashion, the recycle efficiency
can be calculated for each cycle by dividing the velue for the particular
cycle by that for the zero recycle.
Table 16. Effect of Recyele of 99.5% of Actinides Other Than U and Pu
on Activity of Waste® from PWR Spent-Fuel Processing Based on

Total Curies as a Hazard Measure

 

Relative Radiocactivity, Ratio of Standard to Rezcycle

Recyele

Case for Decay Times (years) of:

 

 

Yo. 10 10° 10° 10 5z10° 107 10°
0 6.69  9.97  10.2 10,4 8.4k 13.9 40.9
1 6.12 7.9  10.0 10.0 7.90 11.6 35.8
2 5.88 7.4k 9.91 9.91 7.75 11.0 3. b
3 5.78  7.25 9.85 9.83  7.69 10.8 33.9
L 5.74  7.18 $.83 9.79 7.66 10.7 33.7
5 571 T.15 9,80 3.75 7.66 10.7 33.6
10 5.70 7.13 9.73 9.70 7.6k 10.7 33.5
15 5.70  T7.12 3,71 9.66  T.64 10.6 33.5
20 5.70 7.12 9.68 9,62 7.64 10.6 33.4
30 5.70  7.12 9.67 9.62 7.62 10.6 33.4
40 5.70  7.12 9.66 9.62 7.62 10.6 33.4
50 5.70 7T.12 9.66 9.62  T.62 10.6 33.4
60 5.70 T.12 9.66 g.62 7.62 10.6 33.4

 

®0.5% of all actinides sent to waste.
L7

The results in Tables 14 and 15 show that when recycling is prac-
ticed, the hazard measure of the waste is approximately proportional to
the nevtunium, americium, and curium sent to the waste since the hazard
reduction factor is about five times greater when 0.1% of the actinides
is sent tc the waste after each cycie than that for the 0.5% case. This
cbtains logically because the reactor discharce composition is little
affected by a change of only 0.4% of recycled actinides in the feed
stream. The standard case is also little affected by whether 0.1% or
0.5% of U + Pu is present since the americium and curium predominate
at smaller decay times and neptuniuvmw after 105 vyears. It is for similar
ressons that the cyele efficiencies are virtually independent of the
percentage of maberial that is recyecled. It fcllows that if 92.99%
removal is effected, the hazard reduciicn factors of Table 15 will
ircrease by about a factor of 10 to about 2000 at 106 years, All
three tables show that the hazard reduction factors decrease asymp-
totically with the nuuber of recycles, which is a result of the buildup
of the higher transuranics, and that effective equilibrium is attained

in 20 cycles more or less depending on the decay time.

Minime in the hazard reduction factors of Tebles 14 and 15 as a
function of dscay time occur at around 200 years in the first few
recycles with = sradual shift to between 105 and 10° years for larger
numbers of recycles., The reasons Tor this behavior are rather involved
ang include the relative change in toxicity as well as the change in

total activity of the varicus nuclides.

As an 3ic in uhderstanding this and other phenomensa, the relative
contribution {when > 0.C1%) of each zetinide by itself (regardless of
whether discharged from the reactor or generated by decay) and by each
of their daughters to the hazard measure of zeroc recycles and 60 recycleé'
are shown in Tables 17 and 18, respectively; the standard case for 99.5% ex-
traction of U + Pu is shown in Table 19. (See Appendix II for a listing
of activity in curies and the hazard measure of all actinide nuclides
as a function of time after discharge.) Note that the basis for these
tebles is @ifferent from that of Tables 4 to 6 where the contribution

from sach actinide includes all of its daugaters. Cbserve that in <he
Table 17.

Relative Contributicn of Bach Component of

the Azxtinide
Waste® to the Hazard Measure AFter One Pass Through the Reactor

 

Portion of Total Water Reguired|to Dilute Each Element
to Its RCG (%) for Decay Times (years) of:

 

 

Tlement 10 10° 10° 10 s5x10° 107 10°
Tb nil nil 0.01 1.0 I, 2 L2 1.7
Bi nil nil 0.03 k.o 17.9 19.5 16.6
Po 0.05 0.0% 0.05 7.5 32.2 3k.2 2.4
At nil nil nil 0.4 3.8 5.1 10.5
Rn nil nil nil 0.11  0.35 0.35 0.36
Fr nil nil nil 0.4y 3.8 5.1 10.5
R, nil nil 0.02 3.9 17.6 19.2 16.3
Ac nil nil nii o,y 3.8 5.1  10.5
Th nil nil 0.01 0.63 k.3 5.6 11.0
Mp G,02 0,03 0.13 0.38 0.2% 0.16 0.1¢
Pu 62.3 ™7 2.1 74.9 11.6 1.8 0.02
Am 21.7  63.4 54,0 1.6 0.03 nil nil
Crn 16.0 1.7 3.h h.7 0.12 nil nil

 

%0.5% of a1l actinides sent to waste.
49

 

 

 

Teble 18, Relative Contribution of Each Cowmponent of the Actinide
' Waste? to the Hazard Measure After 60 Recycles
Portion of Total Water Required to Dilute Each Element
to Its RCG (%) for Decay Times (yeers) of:

Element 10  10° 10 100 5x10% 100 100
Fb il nil nil 0.47 5.8 ho = 1.9
Bi nil nil nil 1.7 19.1 20.3 16.8
Po 0.05 0.05 nil 3.h 35.8 37.3 23.1
At nil pil nil 0.13 2.7 3.6  10.2
Rn nil pil nil 0.03 0.21 0.21  0.30
Fr nil nil nil 0.13 2.7 3.6  10.2
Re. nil nsl nil 1.7 18.9 20.1  16.5
Ac nil nil nil C.13 2.7 3.6 10.2
Th nil nil nil 0.19 3.1 b1 10.7
Iip nil 0.0L 0.0k 0.11 0.18 0.11 0.10
Pu 20.6  27.3 10.0 18.4 7.1 1.2 0.02
Am 5.6 24.9 2.4 0.55 0.03 nil nil
Cm 41.8  43.5 75.9 73.0 2.7 0.95  0.19
or 22.9 1.3 1.6 nil  nil pil  nil

 

a0.5% of all actinides sent to waste.
Table 19,

50

Relative Contribution of Each Component of the Actinide

Waste® with 99.5% of U and Pu Extracted

 

Portion of Total Water Required teo Dilute Fech FElement

to Tts RCG (%) for Decay Times (years) of:

 

 

Flement 10 10° 103 0% sx10t 100 1P
Pb nil nil nil 0.28 0.94 0.77 0.10
Bi ni nil 0.02 2.7 1k.8 15.2 bk
Po nil nil 0,03 3.6 17.8 17.5 4.4
At nil nil nil 1.7 11.7 12.6 1k,1
Rn nil nil nil 1.2 0.03 0.03 0.02
¥r nil nil nil 1.7 11,7 12.6 14,1
Ra nil nil nil 2.7 1k.8 15.1 4.k
Ac nil nil nil 1.7 11.7 12.6 14,1
Th nil nil 0.01 1.8 1.8 12.7  1k.2
Pe. nil nil 0.01 0.0 0,02 0.01  nil
U nil nil nil 0.02 0.02 0.02 0.0
p 0.09 0.38 1.0 1.8 0,76 0.39  0.13
Pu 7.1 16.7 7.7 13.8 2.8 0.37 0.13
Am 13.8 55.2 6.7 17.1 0,30 nil nil
Cm 79.2 27.8 L. 5 3.9 1.0 0.0L  nil

 

20.5% of U + Pu and 100% of other actinides sent to waste — standard

case,
21

staﬁdard case (Table 19) the curium is 79% of the total after 10 years

decay and varies from 16 tc L42% during recycling (Tables 17 and 18). In
}

Shon (18-year helf-life), but

its fraction of the totszl curium content diminishes with recyeling.

the standard case, the curium is mostly

Consequently, it is the relatively rapid decay.of the Zthm (the con-
trolling nuclide in the standard case) compared with the smaller effect

on the recycled waste that 1s responsible for the initial rapid drop in the
hazard reduction factors for the first 100 years cr so. The shift in the
location of the minima to larger decay time can be attributed to the
buildup of ZhBCm.and 2h60m

After the first few recycles, the hazard reduction factors rise to
gbout the same value, regardaless of the number of recycles for long decaj
times. This reflects the fact that Np discharged from the reactor
controls the waste hazard at long times and that the concentration of
this nuelide rapidly reaches equilibrium, ‘This is borne out by the
detailed date which show the 233 decay chain (4n + 1 series), of which
237

Np is a menber, mzkes the dominant contribution, Table 20 shows the

237

neptunium (aimost 21l Mp) discharged from the reactor attains a

constant value after five recycles.

Recycling of the actinides and achieving a 99.9% extraction effi-
ciency reduce the hazard measure of the actinides at equilibrium to the
same order as that of the long-lived fission products (1291, 93Zr, 93mNb,
99Tc, and 13503) for the longer decay times, the hazard measures of the
actinides being ahout twice that of the long-lived fission products at
1000 years and dropping to asbout one-half of the fission preoduct value
ot 10° vears. However, if 1291 is eliminated as a hazard by isotopic
dilution (or separate storage), the actinides would still control the
totel waste hazard. An actinide extraction efficiency of 99.999+% zlong
with the recycling is required before the hazard measure of the total

waste hazard is controiled by the long-lived fission products other than
129
I.

Figures 2 and 3 were prepared from the ORIGEN output to show the

rate of accumlation of the potential hazard of the actinide waste from
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12
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- -~
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= NO EECYCLE—QQ.S?.. CF U+ Py REMOVED
- —

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CUMULATIVE HAZARD, m3 OF WATER

 

 

 

1019—
_
i RECYCLE OF 925% OF ALL ACTINIDES
EXCEPT U AND ®u
| N
logr-: _—
- ..
|_ ~
- N
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108 ) ) | 1 i l
0 0 20 30 40 50 60 70

TIMZ AFTER DISCHARGE, yr

Tig. 2. Short-Term Cumulative Hazard of Actinide Weste from 60-Year
Cperation of =z Typical PWR.
54

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110l
55
one typical PWR operating at 1000 MWe with an 0.8 plant factor and an
annual average discharge of about 23 tons/year of spent fuel. Also shown
is the effect of recycling on the hazard reduction for ar actinide extrac-
tion efficiency of 99.5%. Trese plots represent the waste hazard accum-
lation per reactor for 60 years of operation when disposal or permanent
storage occurs aiter .10 years. This operation time span seems sufficient
to cover The lifetime of a FWR nuclear power industry. TFigure 2 shows
the cumnlative hazard from 1C te 70 years affer discharge. The hazard
reducticn Factor (ratio of the standard case to the recycle case at any
indicated decay time) achieved by recycling starts at about L0, rapidly
drcpslto 20 in 10 years, and slowly diops tc about 10 by 70 years after
discharge. Figure 3 shows the sanme resulis for decay times between 500
and 10~ years for which decreases are shown initially in the cumulative
hazard because the reactor was considered shut down at 60 years of
operation with no further additions to the waste. IAt later decay times
the cumulative hazard increases because of the buildup of “°7Th and its
daughters as the result of decay of 23TN§. This increase would be much
smaller if RCGs of ref., 8 were used and the curve would flatten ocut with
the RCGs of ref. 6 {see Teble 5). TFor the long-term decay pericd shown
in Fig. 3, the hazard reduction factor has dropped initially to a little
below 10, but after 10J+ years builds back up to a little over 40. At
these longer times the hazard reduction factor can be simply obtained by
averaging the values for the first 30 recyeles (60 years of operation)
shown in Table 1L because essentially all the hazard comes from actinides
with long half-lives, the shortest being 458 years for ZulAm.with the
‘others having half-lives measured in thousands of years. For such a
condition, it mekes 1little difference that the first discharge‘is 60
years older than the last one; each discharge contributes about the same
to the cumulative hazard for times greater than 500 years. This is, of
course, not true for the shorter times, which is the reason for the gap
between 60 and 500 years when the hand caleulations using the normal
ORIGEN outpul become toc long To be practical.
|
!
- - |
8.5 Effect of Recycling on the Hazards of
Chemical Processing and Fue?! Fabrication
|
Recycling of reactor actinide waste will increase the radiation
_ N b . .
problem associated with chemicel processing and fuel |Tebrication because
of the increased radicactivity of the reactor feed ané discharge streams,
In this section these problems are examined somewhsat {gnd the effect on

chemical processing 1s discussed.

Table 20 shows the recycle effect on the buildupof each actinide
in the reactor discharge stream after cooling 15C days and before any
chemical processing. The tabtle also indicates the attiinment of effec-
tive egquilibrium. True equilibrium cannot be atitained \in practical
irrediation times because of the small removal cross sedtions (decay +

2l z
capture + Fission) of “Cm and 220

Cm. 7The small changes irn the actinides
that are still occurring afier 60 recycles can be traced\primarily to -
2h8Cm; the amcunt of 25OCm.present is too small to hsave %:noticeable
effect. From a purely chemical separations viewpoint, thé changes in
compositions are not significant. Handling problems; howéwer, are
increased by the buildup of nuclides that undergo spontaneous fission.

The slight increase of sbout 3% in the gamme activity that occurred as

the result of recyecling iIs of no consequence.

The Purex separations process now in use removes only Pu and U from
the waste stream (see Sect. 8.2 for a2 more detailed discussion on chemical
proééssing). For. recycling, the other actinides must also be extracted from
this waste strzam centaining the fissicn products. It ic generally felt that
the process can be adjusted to permit $9.9% extraction of the U + Fu. By
small changes in the process, neptunium coulid also be extracted. Removal
of the Am and Cm Is not as =asy since som2 of the rere-earth fission
products have simiiar chemistry. The separation would not have to te
toc clean,btnt contaminaticn with rare carths with high neutron cross
zzctions shouid be large enough to degrade tne neutron economy of @he
resctor when recycied. Whether the Ac, Th, and Pa ere removed or sent
to waste is not important zince their effect on the hazard is negligible

see Sect. 5). The effect of Bk, Cf, and Esz produced in one pass through
P
o7

the reactor also has a neciigible effect on the waste hazard, but the
buildup of Cf and Cm seriously increases the neutron emission rate by
Ct,n reactions and by spontanesus fission, with the latter domin.ating.
Table 21 shows that the source of neutrons in the material vecycled is
wostly curium (primerily 2 icm and 2*2em) sfter two cyeles, but that the
Cf (primaxily 25 2Cf) rei:i_dly becomes controlling after a few recycles.

This incressed neviron activity due to recycling should cause no
real problem in chemical processing since the thick concrete walls
required for gamma shielding should also be adeguate for the neutrons,
A poteniial problem arises in fuel fabricaticn and handling, regardless
Iof' whether the racycle material iIs mixed with new fuel or wmade into
separéte elements. The same prcbleni alsc exisits fér fabricating fuel
elements from recycied plutonium. Bell and Nichols? estimate that the
neutren source for reeycled plutonium builds up to aboud 109 n/sec ver
metric ton of plutcnium. Table 21 shows that if the californium is
removed, the curium would control and the neutron source would only be
reduced by a factor of 300. The curium produced along with the sssoci-
ated neptunium snd americium would generate considerably more neutrons
Iper unit weight (3.6 x 10t

by spontaneous figsion ard &,n reactions than would plutonium., The

per metric ton of the mixture st egquilibrium)

quentities involved, however, are smaller than in the case of plutonium
recyele, It seems that actinide recycle materizl could be handled withe
out too mach change in the way of design or handling procedures developed
for piutonium recyrle fuel even if the neutron source strengtin is somewhat

_ — e
larger. Removal of the neutron source, 252

Cf, can be accomplished by not
recycling Cf even thouvgh there fs an incresse in the Cf production with
curium buildup. 3igrificant 22 ECf buildup occurs from successive aeutron

z25¢C

. cen 240 2hg
captures starting with Cf and Cf, vwkere pracursers are Bk and

£0%.  The difficulty imvolved in the efficient extraction of Am and.

Cm from fission products was pointed ouh in Seet. 7 2. Ip all processes for
removal of Am and Cm, Cf and Bk are alsc removed. Consequently, addi~
tional complicetions tc the fiowsheet wonuld be required o keep these |

elenents separate from the extracted Am and Cm.
58

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59

Recyeling of waste through a reactor does add to the inventory of
hazerdous materials in process which could increase the severity of an
accidental release of radicactive material. The hazard measure of the
ectinide waste based on ingestiorn was increased by only 12% after 60
recycles. The total is about one-tenth of that for the fission products.
If the hazard_measuie is based on inhalzblon, recycling inereases the
potential hazard by a tactor of 2 at discharge (see Table 21) with the
average in the reactor being significantly higher. Since the fission
products produce an inhalation hazard measure of only 3.7 ¥ 1016 memetric
ton of fuel compared to a value up to 2.5 x 1017 for the actinides, it
would seem that the potential hazard of an operating reactor would te
increased by recycling of the actinides. . However, actinide concentra-
tion ik a resctor is not significant in analyzing the "maximmm credible
sceident” (MCA) since the actinide compounds are not veletile and cannot
e significantly dispersed into the atmosphere by any credible resctor
accident, Transmutation of fission products in burner reactors would,
of course, add to the potential hazard of the MCA because the volatile”

fission products are controlling in an accident analysis,
€0

9. CONCLUSICNS AND RECOMMENDATICNS

9OSr, 13705, and 85Kr, by

neutron-induced transmutation as a result of recyelirg in existing or

Elimination of the fission prcducts,

projected designs of power reactors is not possible since the neutron
fluxes are not high enough to lower the effective half-lives of these
nuclides by a significant amount. Special burner reactors with neutron

17 n/cm?-sec are reguired for that purpose.

fluxes in the order of 10
Spallation reactors and fusion reactcrs are possibilities,but the latter
is certainly not feasible with current technclogy. The former, at best,
would require an extensive development program including, in particular,
a method for coping with the potentially severe radialion damage and heat
transfer problems. It seems that ultimate storage in deep geological
formations of Mmown cnaracteristics {such as salt mines) remains the
best method for fissicon product disposal since less than 100Q vears is
required to reducz the activity fto an innocuous level. Assurance of
tectonic stability for 1000 years with a very high degree of confidence
is quite possible in some geological formaticns. The actinides and their
danghters, of course, with half-lives measured In many thousands of years
should be =xcluded from the biosphere for a length of time for which
tectonic stability can be assured with a lesser confidence level. There
is, therefore, a stronger motivé for disposal or reduction in the sccummu-
lation of the actinides by scme other method such as by transmutation in

nuclear reactors.

Waen over 99% of the plutoniur and uranium has been extracled,
significant further reduction in the potential long-term hazard of the
waste from PWRs (and undoubtedly other types) can only be achieved by similar
removal of the neptunium, americium, and curium. Consequently, if the
actinides are to be disposed of separately from the fission products, it
pays from a waste hazards viewpoinit {o concentraie on developing economic
chemical processes for removal of the latter three actinides from the

fission products.

Recycling these actinides through the reactors producing them has
promise for réducing the long-term waste hazard, particularly if 99.9%
61

extraction of neptunium, americium, and curium is schieved, an extraction
efficiency that already appears feasible for piutonium and uranium. The
results of this study indicate that long-term hazard reduction factors

up to a@bcut 200 are possible with a $9.9% actinide extraction efficiency
with subsequent recycling of the neptunium, americium, and curium, L&rgér :
hazard redvction factors are possible with higher efficiencies, and the
hazard reducticn factors are approxiwately proporticnal to the quantities

of the zetinides sent to the waste.

Recyecle of actinides in the ILMFBRs shcoculd produce cven higher hazard
reduction fectors sincs the average fission-to-capture -atio of the acti-
nides should be higher in a fast reactor than in a thermal cne. Fast
cross—-section sets for the higher actinides shonld be developed so that

recycling'studies can be mads for the LMFERs.

It also appears that recycling of the actinides is particularly suited
for a fiuid foel reactor such as the MSBR.El A proecessing scheme has been
visuvalized that recycles essentielly all the uranium, neptunium, thorium,
and most of the other actinides. Considerably less americium end curium
are produced compared to a PWR, which considerebly simpiifies the weste
disposal problem., In addition, being a fluid fuel reactor, the problems
arising from fabricetlion and haniling of heavy neutron-emitting fuel
elements are eliminated. A study similar to this one should be made for
the MSER using chemical processing that minimizes the actinide content

of the waste streams.

Official or standard valuss for the RCGs for nuclides appearing in
the waste that are unlisted in the Code of Federal Regulations should be
established since using the recommended defeult wvalues seems too conserva-

tive for decey times beyend 10,000 years.

Some conslderatior should be given to otiher methods of evaluating
the potential hazard of the waste from chemical processing and possibly
a standard developed thet considers the probability of discharge into the
biosphere. In particular, scenariocs of possible interaction with the
environment and pétential pathways to the biosphere should be evaluated
as part of the conceptual design and site selection process for waste

repositories.
é2

10. REFERENCES

M. Steinberg, G. Wotzak, and B. Manowitz, Neutron Burning of Long-
Lived Fission Products for Waste Disposal, BNL-3558 (1958).

 

M. V. Gregory &and M. Steinberg, A Nuclear Transmutation System for
the Dispogal of Long-Lived Fission Product Waste in a» Expanding
Nuelesr Power Economy, BNL-11915 (1967,.

 

D, Steiner, Scme Preliminery Cbservations Concerning the Role of
Fusion Reactors as Radioactive-Waste Burners, FRT-MEMO-T1{2).
Intra-Laboratory Correspondence, Osk Ridge Nationeli Laboratory
(June 7, 1971). ‘

W, C. Wolkenhauer, 'The Controlled Thermonuclear Reactor as a Fission
Produet Burner,"” Trans. Am. Nucl. Soc. 15(1), p. 92 (1972).

M, J. Bell, ORIGEN, The ORNL Isotope CGeneration end Decay Code,
ORNL~4628 (in preparation).

M, J. Bell and R, 8., Dillon, The Long-Term Hazard of Radicactive
Wastes Produced by the Enriched Uranium, Pu-22°U, and ¢35U-Th Fuel
Cycles, ORNL-TM-3545 (November 197L1;.

 

F. Gera and D. J. Jacobs, "Hazard Potential of Rediocactive Waste,”
Paper 4, International Symposium on Rediocecology Applied to the
Preotection of Man snd His Environment, Rome, Sept. 7-10, 1871.
Code of Federal Regulations, Title 10, Fert 20, Col. 2.

M. E. IaVerme, unpublished results, Osk Ridge Naticnal Ieboratory.

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

M, J. Bell, Radiation Properties of Spent Pluftonium Fuels, ORNL-TM-
3641 (Januery 1972).

 

W, D. Burch, J. E. Bigelow, L. J. King, Transuranium Plant Semiannual
Report of Production, Status and FPlexns for Period Ending June 30, 1971,
CRNL-4T71S (June 1971).

 

Evaluaited Nuclear Data File B, Version II {ENDF/B-II), Tape available
from the Netlonal Neutron Cross-Section Center at Brockhaven National
Teboratory.

L., D. Gordeeva and G. N. Smirenkin, "an Pmpiriecal Formula for the
Average Number of Fission Neutrons," Sov. At. Energy 14, 6, 565 (1963).

J. C. Hopkins and B. C. Diven, "Prompt Neutrons Ffrom Fission,"
Nue. Phys. 48, 433 (1963).
16.

17.

18.
19,

20.

21,

63
H. K. Clark, "Critical Masses of Fissile Tramsplutonium Isotopes,"
Trans. Am. Nuc. Soc. 12, 886 (1969).
C. E. Bemis, Jr., R. E. Druschel, J. Hzlperin, S. R. Walton,
"Thermal-Neutron Capture Cross Section and Resonsnce Integrol for
10.7-Year 9%Kr," Muc. Sci. Emg. 47(3), 371 (March 1972).

J. P. Nicheols, private comﬁunication, Ozk Ridge Naticnal Lsabcratory.

J. P. Nichols, J. O. Blomeke, and R. E. Leuze, private communicztion,
Cak Ridge National Labcratory (October 1972).

M, J. Bell and J, P. Nichols, "Penetrating Radiation Dose Rates and
Shield Requirements in Fabrication of Fuels Cortaining 233U and
High Dxposure Plutenium,” p. T4, CONF-70C502 (1970).

L. E. MclWeese, private commnication, Oak Ridge National Laboratory.
65

APPENDIX I: A COMPARISON OF RCGs CALCULATED BY LaVERNE (REF. 9)
WITH THOSE IN THE CODE OF FEDERAL REGUIATICHS (REF, &)
 
 

 

 

Table Z2. Compariscn of RCGs from Ref. 8 and Ref. 9
ticn RCG
Inhalation RCG Critical _Iogestion RCG
iacal £ Ref. 8 Jrgan Ref. 9 Ref, 8
Nu~lide Orgen Rel. 9 .
-£
-1l GI (LLI) 5 x 10 - &
225 ¢ Laver 8 x ol gyl Bone 2 x 10 2 x 107G
ot Bone -9 -9 &I (ULI) 2 x10 @x1
2289.: Bone 2 x 10 3 x 10 ( P h .
-l3 '13 Bone l[» x 10_ X 10_
2hlpn Bene, k?.dney 2 x ig_l3 g ; ig_l3 Bone 5 10"25; 5 x 1078
ElﬁgmAm Rone, kidney 2 x 105" 2xlog oateLT, 101) xiot 1o
: oo miver 3 -13 2 x 10713 Bone b x 107 b x 10_3
2¢3am  Bome, kidney 2 x 107 :xic7  er(s) 2x 1073 510
ahbey Bone, kidney ix 10_7 L &1 (vi1) 1 x 10 _
e45xm  GI (ULI) 3 x 10 - i
- hyroid 6 x 10~ -
217a¢ Ovary, thyroid 9 x 1077 - Ovary, thyr h
-1 10-1t @I (iI) % x 10~ -
e oome :x 13-9 ; i 10~ GI (ULI) 3 x 107¢ -
250mKr Eone 5 x 1 i s
-10 -0 er (Lin) % x 10- x
€1l0g; Xidney 2 x 12_7 2 x 10 oI (8) 7 10'3 -
Bl ey 30 3r109 o1 (s) bx 109" b x 10
212 Kidney Ix 1 9 X o1 (55 5 l°:1+ =
2l3p; Kidney L x 30 2 - o) 2* 2 ;
2ibpy Kidnew 1 x 10 ) . 0-6
- 1
1k 1 a~-1k Rone L ox 10- b x 10
249 Bone ox a3 2k 13.13 Bone 1x1072 1x1072
poaCE  Boms sx10 Er0lh pge pEMe RIS
ggjz":f z x 1013 7 x 1013 61 (11I)” TP 2x 2%
253‘3f ?,‘;“e 3x 107t 3x10” 61 (ILI) é *® ig_?. i :,: W
Cf ne 213 &I (LLI) .
ESth Bone 2x 10 2x 10 .( s 2 10_5
-12 12 (i 2 x 10” x 1072
2ha Liver b ox 12.13 ; . lg-la Bone 5 x 10_2 5 x 10
zll:la;cm sone g i 1013 3x103  Bone 7 x 10 o Z X lg_
21‘ Bone 2x107+ 2 x10 Bone hox 107 b x i0'6
W o 2x1013 2x1023  Bome 3x1072 4 x 107
glrrcm pore 2x1013 2x1053  Bone 2 x 107 i x 58,7
O Bone -1k Bone ® L _
ehacm Bone Jx 10-7 i i ig-? ) 2 x 10'3 2 x 1073
249  Liver 2% 1071 S e E Xt X
2500 Bone 3x1ic - i -
- -11 ] 2 x 107 2 x
2335 Bome zx 10 3x10 GI (LLI) ;
-7 - Bo 8 x 10” -
22lye Body 1x 107, Bcggrr 3 % 10 b _
223n. Body b x 10 -
67

Table 22 {continued)

 

 

 

Critical Tnhalation RBCG Critical Ingestion RCG
Nuclide Organ Ref. 3 Ref. 8 Crgan Ref, 9 Ref. 8
gggpb GI (ULI) 2 x 10“{2 - 12 GI (8, TLI) 3x 10“,3; -7
UFb  Kidmey % x 20732 b x 10° Beme, kidney 2 x 107 1x 1€
Skt Kidney 1x 10_1 - o Kidney b x 107 -
e Xidney 6 x 10'80 6 x 10°1°  Xidney 2 x lO'E 2 x 1077

J*Pb Kidney 1l x1107 - Kidney 5 x 107 -
Eggn;; Bone 2 x 1033 - 33 61 (w3) 5 x 1072 - ¢
5381‘39 Bone 1x10 5 1x10 Bone 2 x 10° 3 x 107

p Bone 2 x 107 - GI [ILI) 5 x 1072 -
239 6I (LLI) 2 x 100 3 x10°° 6T (ITI) 8x105 1x10%
2home, g3 (s) 8 x 1077 . e1 (3) b x 10-3 -
e g1 (s) 5 x 1077 - 61 (S) 2 x 1073 -
23;33'.911 Bone 6 x 10‘15 -y L () 3 x 10772 -
23%py Bone 7 x 10'& 7 x 10'%1+ Sone 5 x 107 5 x 107
§§9Pu Bone 5 x 10t , 6x207,  Bone 5%10°2 5 x10°
2131’“ Bone 6 x 10714 & x 10 Sone 5x107 5 x 107
aLplt  Bone 3x20778 3x1071%  Bone 2 x 10 2 x 107
o Bone 6 x 10 & x 10~% Bone 5 x 107 5 x 1G
aT (UnT) 9x108 6x108 61 (ULI) hox10% 3 x 104
Eli-hm Bone 6 x 3_0'11" 6 x 10'11* Bone 5 x 10~ kx 10'6
2k5py GI (TLI) 1 x 107 - GI (LLI} 6 x 1077 -
210p,  Kidney, spleen 2 x 100 2 x 10011 Kidney 8x1077  Tx 10°7
21lpo Kidney, spleen 9 x 10;5 - 61 (5) 3 x 107 -

%0 Kidney, spleen 1 x 10 - ¢I (5) 4 x 10 -
21317: Kidney, spleen 1 x 1011 - GI (8) Ix 102 -
21550 Kidney, spleen 2 x 1072 - GI (8} 7 x 10 -
S12F0 Kidney, spleen 1 x 1077 - ¢I (S) 3x mg -

Po Xidney, spleen 2 x 107 - Kidney, spleen 7 x 10 -
218 -7 S1 tT) 3
Po Xidney, spleen 1 x 10 - GI (8} y x 30 -
Sglz?a Bene 4 x 10:?* L x 10'1h Bore 9 x 10'? 9 x 10'7
223E® Liver 7 x 1073 - B GI (LLI) 6 x 107 -
2%4 Kidney 2x107Z 2x10 61 {(LLI) 1207, 1x10
e 6T (S) b x 1075 - Gl (s 2 x 1077 -
Hpg 6I (ULT) 3 x 10 - cI (ULI) 1x 10 -
gf_,h Bone 6x20° 6x107  Bome 7 x 10"; 7 x 10%
s Rn Bone 2 x2010 2z x102%  Eone 2 x 107 2 x 10
< gRa. Bone 5% 1071t - 12 Bona 5 x 10‘_; -8
: 3381;2 Bone 1xX '}g 3x 1075  Bone l1x10 3x207
22 Bone 2 x10°% 2 x 10 Bone 3 x 10 3 ¥ 10
68

Tavle 22 {continued)

 

 

 

 

Criticel Inhalation RCG Critical Ingestion RCG
Nuclide Organ Ref. 9 Ref, 8 Organ Ref, 9§ Ref, B
|

20T oI (s} 2 x 1072 » GI (S) 8 x 2073

208 6T (S) 1x 10'2 ' 61 (8 5 x 1673

20979 GI (s} 2 x 10° 61 (8) 1 x 1072

aagm Bose 1 x 108 61 (iLl) 2z x 10'2 5

22 Bone 3x 1023 3 x 10713 Banie 7 % 107 7 x 10

229m, Bone 2 x 10'13 v Bone b x 1077 6

230ny, Bene T=x 10:%‘ 8 x 1077 Bone 2 % 107 2 x 107

231qy, GI (LLI) & x 205, ) 61 (IL1} 2z x 107% .

232q, Bone 7 x 10-1% 1 x 30712 Bone 2 x 10 2 x 1070

233q, eI (S) 5 x 1071 6I (8) 2 x 1003

234, Bone 2x10%  2x107 6I (L) 2x10°  2x10°

232y Eone bx1071% 3 x10°) GI (Lil) 3 x 10*5 3 x 1075

23% Bape 2 x10° 2 x107H 6I ([1I) = 3x 10\ 3 x1i072

234y Bone 2x1072 2x 10’5}1- 81 (ILI) 3 x 107 3 x 1072

gggu Bone 2x10l 2 107 Gl A1) 3x107; 3 x 1077

237 Fone 2 x 107 2 x 10 GI (LLI) 3 x 107} 3 x 1672

23Ty GI (LiI) 3 x 107 . 6I (LL7) 1 x 107

233211 Bone 2 x10° 3 x 1071 GI (LLI) 3 x 1077 4 x 1077
2y GI () 5 x 10°7 GI (S) 2 x 1073

240y GI (111) 9 x 1072 8 x 1079 GI (LLI) & x 1077 3 x 10~

 

ot ———
APTENDIX IT: RADTIOACTIVITY AND HAZARD MEASURE OF EACH ACTINIDE
NUCLIDE AS A FUNCTION OF TIME AFTER DISCHARGE FCR
THE STANDARD CASE AND AFTER TEE €0th XECYCLE
70

Table 23, Radicactivity of Each Wouelide as a Function of Time
After Discharge froam a Typical PWR

 

Radiosctivity’ {Ci/metric ton of fuel) after Decay Times (yvr) of:

 

 

Fuciide 0 12 107 10° 10° 10° 10°
223 1.93-7 L4.15-8 7.24-7 6.94-5  5.54.3  1.17-1  2.90-1
22 9.70-7 7.47-6 2.38-5 2.60-5  3.90-5 2.38-4  3.314
228, 2.73-12  2.1%11 3.78-11 1.08-10 1.22-9 1.86-8  1.95-T
22T, 9.0b-7  7.37-6 2.35-5 2.55.5  3.85-5 2.35.4  3.27.
228g 1.49-3  1.48-k 14.03-5 7.06-9 1.22-9 1.86-8  1.95.7
229, 3.31-8  4.15-8 T.24.7 6.64.5  5.55.3  1.17-1  2.90-1
#30, 1.78-5  2.09-5 3.45-5 3.71-%  3.79-3  2.32-2  5.59-3
231q, 7.03-1 8.56-5 8.58-5 8.74-5 1.16-4%  3.12-4  3.31-4
232q, 2.55-11  3.20-11 3.85-11 1.08-10 1.22-9  1.86-8  1.95-7
233m, 1.33-2 0 0 o o o a

23y, 3,181 1.57-3 1.57-3 1.57-3  1.57-2  1.57-3  1.57-3
2315, 2.44-5  2.47-5 2.48-5 2.60-5 3.90-5 2.38-%  3.31%
232pgy 3.57-1 0 o 0 ¢ 0 0

233pg 3.23-1  3.40-1  3.45-1  3.68-1 3.7M-l  3.64-1  2.72-1
23bm, 3.15-1  1.57-3 1.57-3 1.57-3 1.57-3  1.57-3  1.57-3
23bpy 1.25-2  1.57-6 1.57-6 1.57-6 1.57-6 1.57-6 1.57-6
232y, 6.07-3  B.69-5 3.92-5 6.77-9 0 o 0

233y b.55-5 1.52.5  1.46.4  1.53-3 1.56-2 1.29-1 2.99-1
23k 7.52-1  6.71-3 2.60-2 4.65-2  4.5%-2  3.56-2 L4, 3G-3
£3% 1.71-2  3.56-5 8.58-5 8.7h-5 116  3.12-k 3.3k
2350 2.88-1 1.44-3  1.46-3 1.69-3 3.10-3 4.93-3 3.92-3
237y 8.65+5 0 0 o 0 Q ¢

238y 3.04-1  1.57-3 1.57-3  1.57-3  1.57-3  1.57-3  1.57-3
33?9 1.86+7 0 0 c 0 0 0

2hoy, 1.69-15 5.80-14 6.12-13 6.13-12 6.08-11 5.57-10 2.67-9
23y, 2.91 0 0 0 0 0 0

STy 3.3221  3.ho-1  3.45-1  3.68-1  3.7h-1 3.64-1  2.72-1
23 3,96+5 0 o o 0 0 0

235 1.85+7  1.77#L 1.75+1 1.61+1 7.1k 2.05-2  2.37-7
2hdmy 1.69-15 5.89+1% 6.12-13 6.13-12 6.08-11 5.57-10 2.67-9
amﬂp 3.30+k 0 0 c > o 0
7L

Table 23 {continued)

Radicmctivity? (Ci/metric tor of fuel) after Decay Times (yr) of:

 

 

Nuclide o 10 10% 10° 10" 10° 10°
2%p, 349~ 1.54-% L.79-1% © 0 5 o
2385, 2.7h+3  1.05+2  5,36+1 1.29-1  1.28-19 0O 0
2395y 3.18¢2  1.62 1.67  2.05 %.03 5.65.1  2.37-T
2]"09,,1 L, 77+2 b 47 8,96 8.30 3.30 3.20-4 2.57-9
My 1.05¢5  3.27+2 4.88  3.08-1 1451 7.68-5  1.25-37
31‘2?0. 1.38 6.61-3 7.02-3 7.29-3 7.70-3 5.70-3 1.29-3
23y 3.59+5  2.47-7 2.47-7 2.47-7  2.k7-7  2.46-7  2.37-7
2k 3.69-15 5.90-1h 6.12-13 6.14-12  6.09-11 5.58-20 2.67-9
Hpy 2.23-16 0 0 0 o 0 0
2y, 8.50+1  1,58+2 1.hT+2  3.5+1  1.45.1  7.62-5 0
eham, 4,12 3.93  2.60  L4.30-2 64320 o0 o
2h2y, 7.0+ 393 2.61  h.31-z  6.4320 0O o
k3 1.76+1  1.77+1  1.75+1 1.61+1  T.14 2.05-3  2.37-7
gf‘l‘m -~ 1.30+5  7.67-17 7.96-16 7.96-15 7.92-1% 7.25-13 3.47-12
245 am 5.47-8  1.72-11 0 0 0 0 o
252 3.70¢8 3,23 214 3.,53-2  5.,29-20 O o
3oy 5.62 4.53 6441 2.19-3 0 e 0
2 o 2.58+3 1.76+3  5.60+1 6.90-1% 7.91-1k  7.25-13  3.47-12
zhscn 3. %1 3.34-1  3.30-1 3.07-1 14k 7.61-5 0
Moy 6.70-2  6.69-2 6.60-2 S5.78-2  1.54-2  2.73-8  6.35-31
zl‘Tcm 277 2077 2.hr-7 2.breT 2477 2.46-7 2.37-7
248 8.0L-7 8.00-7 B.01-7 B8.00-7 7.86-7 6.58-T  1.12-7
2490 2.60-2 o 0 0 0 0 0
20 9.35-1%  9.35-1% 9.32-1% 8.99-1% 6.26-1% 1.74-15 4.66-31
zthk 3.63-3 1.1k-6 Q 0 0 0 0
350& 6.50-3 9.35-1k 9.32-14 8.99-1% 6.28-1L 1.7%-15 4. 66-%
249¢p 1.09-6  9.79-6 8.20-6 1.39-6 2.80-14 0 n
20 3.89-5  2.30-5 1.95-7 B8.99-14 6.28-1% 1.74-15 4.57-3L
20lop 2.84-7 2,82-7 2.63-7 1.33-7 2.28-10 O 0
22pp 4.88-5  3.55-6 2.01-16 ¢ 0 0 0
253¢c¢ 4,00-5 0 0 0 o 0 0
2S4ep 1.08-9  7,19.28 © 0 0 0 0
233pe 2.90-6 2.60-58 ¢ 0 o 0 o
Total 3.91+7 2.41+3 3.16+2 T.93+1 2,30+l 2.65+0 3.23+0

 

 

®Read as 1.93 x 1077,

b

At 150 days =fter discharge, 99.5% of U + Pu was extracted,
 

 

 

Table 24, Hezord Measure of Each Actinide Nuclide
After Discharge from a Typical PWR
Hazerd Measure? (m3 oft Hzofmet.ric ton ¢f fuel)
after Decey Times {yr) of:

Muclide 0 10 10° 103 20" 10° 1®
225pe 6.43 1.3 2.4141  2.31+3  1,85+5  3.9146  9.67+6
22 p¢ 5.85-1%  3,7h  1.1g+1  1.30#1  1.9541  L.1gtz  1.6642
228y 3.04-8  2.38-7 h.20-7 1.20-6  1.35-5 2.07-%  2.16-3
2eTq, 3.01+1  2.46+2  T7.82+2  8.53+2  1.28+43  T7.83+3  1.00+k
228111 2.13+2 2.13+1  5.76 1.01-3 .7l 2.66-3 2.73-2
229y, 1.10 1.38 24141 2.31+43  1.B5%5  3.91+3  9.67+6
230q, 8.58 108 1.72+41  1.86+2  1.90+3  1.16+h  2.79+3
23y, 2.6k45 2.85+1 2.86+1 2.91+1 3.8741  1.04+2 1.10+2
22m 1.28-5  1.60-5 1.92-5 5.%1-5  6.08-%  9.31-3  9.73-2
e33m, 1.83-2 0 0 0 0 0 0
23, 1.57+k  T7.85+1 7.85+1 7.85+1  T.85+1  7.86+1  T7.8641
23lp, 2.71+1 2.7l 2.75+1  2.88+1 L.3k41 2.65+2 3.68+2
232pg 1.19+5 0 0 0 0 0 0
z;t:a 3.23+43  3.40+3  3.45+3  3.68+3  3.7M3  3.643  2.7243
23l 3.15-1 1.57-3 1.57-3 1.57-3 1.57=3 1.57-3 1.57-3

Pa. §.15+3 5.24-1  5.2k-1 5,241 524 5,841 5,2L-1
232y 2.02¢2 2,90  1.31.  2.26.% 0 0 0
zzzu L52 5.73-1 B.B7  5.0841  S5.2042  L.28+3  9.62¢3
o3 U - 251+ 2.2z B.68+2  1.55+3 1.50+3  1.19+3  1.h3+2

2y 5.71+2  2.85  2.86  2.91 3.87 10642 11041
2360 9.61+3 L 81«1 L. BT+l 5.63+1 1.03+2  1.34+2 1.31+2
23Ty 2.88411 0 a 0 o 0 0
238, TBS543 3.934L 3.93t1  3.9TL 393 3.931  3.934
239% 1.86+7 o o 0 o o o
2hoy 5.64-11 1.9%6-9 2.04-8 2.04-7 2,026  1.86-5  B8.89-5
23y 9.69¢5 0 0 o 0 0
23Ty 1.1155  1.13*5  1.15+5 1.235  1.25+5  1,21+5  9.06+k
2y 1.33¢11 0 0 0 0 0 0
:39 1.85¢411 1.77+5 1.75+5 1.61+5 7.1k 2.05¢1  2.37-3
240 ) 1.69-15  5.89-1% 6.13-13 6.13-12 6.08-11 5.57-10 2.67.9

Mo 3.30+4 0 o 0 0 0 0
73

Table 2k (coriinued)

 

Hazard Measure® (m3 of H,O/metric ton of fuel)
after Decay Tifies (yr) of:

 

 

Nuclide 0 10 10° 163 10 10 1°
236Pu 1.16+7 5.13+3 1.60-56 c o 0 0
2381711 5.48+8 2.11+7 1.07+7 2.58+% 2.55-14 C 0
239y 6.36+7 32845 3335 L10+5  B.07+5 LS 47he2
2ho,, 9.5447  B.0k45  1.79+6 1.66:6  6.60+5  6.48+1  5.34-h
2l 5.25¢8  1.63+6 2.4Msb  1.583 7.23+2  3.81-1  6.27-3%
2thu 2.7645 1.38+3  1.40+3  1.46+3 1.94+3  1.34+3 2.58+2
23, 1.20¢0  8.24% 8.2kh  8.zh-bh  8.2h%  8.21-h  7.90k
21*1‘1&1 5.65-8 1,97-6 2.0k.5 2,05-4 2,03-3 1.8-2 8.91-2
2h3p, 7.42-5 0 0 o o o 0
zl‘lAm 2.15+7 3.95+7 3.87+7 B.75+6 3.62+4 1.90+1 0
ehem, 1.03#6  9.83+5 6.52+5 1.08:%  1.61-1% O 0
zthm 2.34410 1.314+6 ' B.7c+5 L.kl 2 1halk 0 0
2430y B41+6 hM146  5.38+6  4.03+6  1.78+5  5.13#2  5.92-2
2l 2.60+7  1.53-1% 1.59-13 1.60-12 .1.58-11 1.45-16 6.95-10
245 1.82-3 57271 © 0 0 0 0
zl‘zczm 1.85+9 1.6245  1.07+5 1. 77+3  2.64.15 0 ¢
21‘3&:; - 1.12+6  9.05+5 1.2+5 4.38-k 0 0 0
2o 3.6848  2.5148 8,006 9.71-9  1.13-8  1.04-7  L.96-7
3l‘5c.m 1.11+7 1.11+7  1.10+7 1.02+7  4.81+6 2.54+3 0
21;6@1 2.23+6 2.236  2,2046 1,93+6 5.13+5 9.10-1 2.12-23
24T oy 8.24 5.2% /ﬁ B.24 8.2k 8.21 7.90
thCm 2.67+1 2674  2.67+1 2.67+1 26241 2.19+1 3.73
29, 2.60-2 G 0 c 0 0 0
2500m 3.12-6  3.12-6 3.11-6 3.00-6 2.09-6 5.80-8  1.55-23
2k 1.2142  3.81-2 O 0 0 0 0
250p; 2.17+2  3.12-9 3.11-9 3.00-9 2.09-9 5.80-11 O
k3,0 5.63¢1  3.26+2  2.73+2  L.6541  ¢.32.7 0 0
250¢¢ 1.30+3 7.67+2 6.50  3.00-6 2.06-6 5.80-8  1.55-23
: 9.u& 9.38 8.76 L.38 4,27-3 0 0
2520¢ 1.63+3  1.18+2 6.81-9 0 0 o 0
2530¢ 1.33+2 0 0 ¢ o o 0
25%0p 3.59-2  2.50-20 0O 0 0 ¢ o

- 2535 9.68+1° 0 o 0 o 0 0
Total 6.35+11  3.36+8 T.27+7  2.74+T  1.05+7  3.10+7 6.85+7

 

®Read as k.85 x 107L.

bt 150 days after dischaige,99.5% of Pu + U was extracted.
Tatie 25. Activity of Each Nuclide as a Function of Time
After Discharge After the 60th Recycle

 

Radicactivity? (Ci/metric ton of fuel) after Decay Times {yr) of:

Nuclide 0 10 10° 10° 10t 10° 100

 

1.77-5% 7.66-8  8.37-8 8,747  8.93-5 2.00-3 4.96-3

22T pe 1.00-4  5.12-7  6.50-7 2.3%6-6  1.85-5  1.17-4  1.Lb-h
228, $.13-11 3.90-11 8.11-12 7.42-11 8.43-10 1.03-8 1.05-T
2eTm, 9.9%4-5 5,05-7 6.42.0T 2.32-6 1.82-5 1.15-%  1.42-%
228, 1.95-2  3.32-h  1.50-F  2.60-8  8.43-10 1.03-8 1.05-7
225, 1.53-5 7.66-8 8.37-8 B.74-7  8.93-5  2.00-3 4.96-3
230, 3.10-5  5.37-7  7.19-6  1.38-%  1.)3%3  8.94-3 3.02-3
23, 1.35-5 8.66-5 8.68-5 8,82-5 1.0L-b  Ll.hl-k  1.44-3
23 3.41-10 z.h1-12  8.79-12  T.l2-11  8.43-10 1.03-8  2.07-7
233, 2.41-1 0 o 0 0 0 0

234, | 3.15-1  1.57-3  1.57-3  1.57-3  1.57-3  1.57-3 1.57-3
231py 1.08-k 5417  7.06-7 2.36-6  1.85-5 1.16-% 1.hhh
232py 1.50 0 0 0 0 0 o

233py 7.3-1  2.67-3  3.10-3  5.58-3  6.37-3  6.22-3 3.36-3
33"*“1:3 3.16-1 1.57-3 1.57-3 1.57-3 1.57-3 1.57-3 1.57-3
“<"pa 2.75-2 1.57-6 1,57-6 1.57-6 1.57-6 1.57-6 1.57-6
232y 3.69-2  3.30-%  1.4E-L 2.52-8 0 0 0

@33 5.05-5  3.67-7 1.k7-6  1.93-5  2.57-%  2.19-3 3.63-3
23"‘U T.77-1 4.93-3 1.14-2 1.77-2 1.73-2 1.38-2  1.63-3
235y © 1.73-2  B.66-5  8.68.5 8.82-5  1.00-% 1.1k 1h4h
235, 2.87-1  L.4L-3  1.44-3  1.51-3  1.61-3  2.16-3 2.05.3
23Ty 8.51+5 0 0 0 0 o 0

23y 3.1h-1  1.57-3  L.57-3 1.57-3  1.57-3  1.57-3 1.51-3
23% 1.83+7 o 0 0 0 o 0

2h0y 2.84-9 6,221 b.97-10 L.B49  L.788 1.8  2.37-6
236y, %.53 c 0 0 0 o 0

23Ty 5.25-1  2.67-3  3.10-3  5.58-3  6.37-3 6.22-3 3.36-3
238y, 6.21+5 o 0 0 0 o 0

23% 1.83+7  1.06-1  1.09-1 1.00-1 k.L43-2  1.62-5 3.25-6
;;WNP 2.84-9  6.22-11 4.97-10 L.Bk-9  L.78-8  L.3B7  2.37-6

Np 3.21+4 0 o 0 0 0 0
75

Teble 25 {continued)

 

Radicactivity? (Ci/metric ton of fuel) afier Decay Times (yr) of:

 

 

Noclide 0 10 107 109 10" 10” 10°
236 Pa 9.57-1 L. 21-b 1.31-13 0 0 o o
2385, w43 357+l 1.7T+1 1.64-2  6.91-22 0 0
239y 3.22¢2  1.63 1.63 1.59 1.25 9.83.2  3.39-6
zl‘OPu L,Bi+z  2.46 2.56 2.3 9.30-1  9.17-5 2.10-6
2y, 1.05¢5  3.28t2 k.59 1.70-2  8.08-3 L.26-6 0
2425y 1.38 6.93-3  6.96-3  T7.23-3  8.58-3  T.74-3 1.49-3
2430, 3.5645  3.45-6  3.46.6  3.506  3.54-6  3.52-6 3.33-6
gl‘l‘?u 2.85+9  6.22-11 4. 97-10 4,847 L, 79.8  L.39-7 2.10-6
2455, 3.69-L 0 o 0 c 0 o
Zhlmn 9.11+% 7.11 1.59+1 3.82 B.09-3 L.26-5 0
2hzm, 5,16 2,13-2  L.41-2  2.33-% 34822 o 0
hean 7.3%+4 2,132  1.la-2  2.33-% 34822 0 0
2h3ﬂm 2.16+1 1.09-1 1.09-1 1.00-1 L 432 1.62-5 3.39-6
zmAm 1.56+5 8.09-3k 6.46-13 6.30-12 6.22-12  5.70-10 2.73-9
2hSpm 2.17-2  3.36-8 o 0 o 0 0
220y 4484  1.75-2  1.16-2 1.9k  2.87-22 O 0
2430y 1.7241  6.94-2  9.88-3  3.34-11 @ o o
e, 1.35+h  L72+l  1.50 6.30-12  6.22-11  5.70-10 2.73-9
450 3.53 1.97-2  1.77-2  1.70-2  8,07-3 4.25-6 0
ey, 3.75+1  1.88-1  1.85-1  1.62-1 4.32-2  7.68-B 6.62-26
T o 6.91-k 3,456  3.46-6  12.50-6  3.54-6  3.52-6 3.39-6
248y 121 6.20-%  6.30-%  6.29-h  6.18-%  5.17-%  8.79-5
21’9cm 3.96+3 0 0 0 0 0 0
20 1.95-6  9.74-9  9.71-9  9.37-9  6.55-9  1.81-10 4.86-26
2X9p 1.42+3  2.24-3 0 o o 0 0
20, 2.51+43  9.74-9  6.71-9  9.37-9  £.55-5  1.81-10 4.85-25
H30s 1.02 2.21-2  1.85-2  3.15-3 6.32-11 O o
£Ccr 2.69+1  T.9%-2  6.7h-k  9.37-9  6.55-9  1.81-10 14.86-26
Eler 3.03-1  1.50-3 1.k0-3  7.01-%  6.85-7 0 o
252ae 3.31+2 1.21-1 6.94-12 0 0 o 0
2330 3.4541 0 0 o 0 0 0
25k oe 1.7k-2  5.82-23 0 0 0 0 0
€93 3.41+1 c 0 o 0 0 o
Total 3.80+7  L,.32+2 kel 821 2.39 2.49-1  9.66-2

 

3pead as 1.77 x 107°.

Byt 150 days after discharge, 99.5% of Pu + U was extracted,
76

Table 26, Hazard of Each Actinide Nuclide as a2 Functiozn of Time

After Discharge After the 60th Recycle

4

 

Bazard Measure® (m3 of I-IZO/metric ton of fuel}
After Decay Tizes (yr) of:

 

 

Nuclide o 10 10° 10° 16" 10° 10°
22 5.88+2% 2.35 2.79 2.91+1  2.98+3  6.86+4  1.65+5
B2 e 5.04+¢1  2.96-1  3.25-1  1.18 9.26 5.83t1  T7.19+1
228, i.ci-6 b3k 9.01-8 B.247 9.37-6  1.15-h  1.16-3
22T 1.30+3  L.69%1 2.kl 77541 6,042 3.B3+3  B.73+3
228y, 2.79+3  bL.4Ssl 2.1+l 3.71-3  1.20-4  1.47-3  1.50-2
22, 5.00+2 2.55 2.79 2.91+1 2.96+3 6.66+k  1,65+5
2300y, 1.55+1 2.A3-1 3.60 6.50+1 7.22+2 E h7+3  1.51#3
23ilm, L. 50+5 2.80+1 2.89+1 2.54+1 3.35+1 L 6g+1  Lk.ororl
232, 1714 1.21.6  Lho6 3735 Lozzh 5.15-3  5.2h-2
233m, 2.41.1 0 0 0 0 0 g
234, 1.57+k  T.86+1  T.86+1  T.86+41  7.86+1  T.B6+1  T.85+1
Z3lp, i.16+2  6.01-1  7.85-1 2.62 2.06¢1  1.20+2 1.60+2
23py 5.02+5 0 o 0 o 0 a
233pa 7.3143 267+l 3.10+41 5.58+41 6.37+1  6.22+1 L4 65+1
234 “pa 3.16-1 1.57-3 1.57-3  1.57-3 1.57-3 1.57-3 1.57-3
235y 9.16+3  5.24-1  5.2%-1 5,241  5.2%-1  5.24-1 s5.24-1
232y 1.23+3  1.10+1  L4.87 8.40-i 0 0 o
233y 1.68 1.22-2  &.9l-2 6.42-1  B.57 7.3041  1.6h+2
234y 2.50+k  1.6h+2  3.80+2  5.01+2  S5.78+2  L.61+2 B,50+1
235y 5, 77+2 2.89 2.89 2.94 3.35 k.69 k.79
2368 G.58+3 L.o7o+1 k. 82+1 5.03+1 6.36+1 7.22+1  T.03+1
23Ty 28411 0 o 0 0 0 0
23y 7.86¢3  3.93+1 - 3.93+1  293+L  3.9%1  3.9%L  3.9%1
23% 1.83+7 o 0 0 0 0 ¢
zl*ov 9.18-s 2.07-6 1.66-5 1.61-h  1.59-3 1.46-2 6.99-2
236y, 1.51+6 0 0 0 0 0 0
237y, 1.75+5  B.g2+2  1.03+3  1.B6+3  2.12+3  2,07+3 1.55+3
28y 2.0+11 0 0 0 0 0 0
238 1.83+11  1.09+3  1.09+3  1.00¢3  L.h¥+2  1.62-1 3.39-2
& 2,849  6.22-11 h.97-10 %.84.9 L 78-8 L 387 2.10-6
Ehoﬂp 3.21+h 0 0 0: 0 0 0
Tatle 26 (continued)

 

Hazerd Messure® (m3 of H,0/metric ton of fuel)

 

 

After Decay Times (yr) of:

Muclide o 16 10° 10° 1c? 10° 10°
2365, 3.19+7  1.h0+h 4.37-6 0 0 0 0
238g, 1.50+9  T7.13+6  3.5846  3.28+3  1,38-16 0 c
3%y 6.7 32745 3.26+5  3.18+5  2.40+5  1.97+k  6.7B-1
2hop, 9.6317  4.93+5  5.12+5  14.68+5  1.85+5  1.83+1 4.20-1
2f‘lPu 5.27+8 1.64+6 2.20+%  8.52+1  Llch+1 2.13-2 ¢
232p 2.77+5  1.39+3  1.3+3  1.45+3  1.72+43  1.55+3 2.98+2
243, 1.19+9  1.15-2  1.15-2  1.17-2  1.18-2  1.17-2 .1%-2
zhhpu g9.hg.2 2.07-3 1.66-2 1.61-1 1.60 1.46+1  7.00+1
25y 1.22:2 0 0 o 0 0 g
b, 2.2847  1.78+6  3.9846  Q.54+5  2.0243  1.06 o
Q_hzmnm 1.1246  5.33¢3  3.5h+3  5.8241  B.71-17 O 0

22 2.45+1C  7.11#3  W.71+3  7.7B#1  1.26-16 0 0

2k 3t 5. 47+6 2.Th+h 2.7+ z.50¢k 3.1a+k 4 05 8.48-1
21‘1*.&:1 3.18+7  1.62-11  1.26-10 1.26-9  1.24-8  1.14-7 5.46-7
245 7.22+42  1.12-3 o 0 0 0 o
21‘36::1 2,24+9 8,76+2 5.50+2 9.57 1.43.37 0 0
21"E’c:m 3.L5+6 1.3G+h4 1.98+3 6.68-6 0 0 0

2k 1.98+0  B.75¢6  2.15¢5  9.00-7 8.89-6  8.15-5 3.90-h
25y, 1.18+8  5.86+5  5.8045 5.67+5  z.69+5  Ll.k2+2 0
M6, 1.25+9  6.25+6  6.18+6  5.10+6  1.W6 2.5  2.21-18
257 o 2.30+%  1.12+2  1.15+2  1.17+2 1.18+2  1.1T+2  1.13+2
2he, b2+ 2,104k 2,10+ 2,104%  2,06+k  L1.72+h 2.93+3
2490n 3.96+3 0 0 0 0 0 0
2500, 6.50+1  3.25-1  3.2h-1 3,121 2.18-1  6.05-3 1.62-18
2h8 o B,73+T  TJATHL o 0 0 0 0
250g, 8.36+7  3.25-4  3.2h-3 3124 2.18-%  6.05-6 1.6e-21
2h9¢r 34147 T.37+5  6.1T+5  1.05+5  £.11-3 0 0
250;¢ 8.57+8  2.65+6  2.25+%  3.12-1 =2.18-1 6.05-3 1.62-18
eee 1.0147  5.01+%  1.68+h  2,3hsk  2,2841  1.99-20 O
252ce 1.10+410 40246 2,31k 0 0 0 0
53¢ 1.15+9 0 0 0 0 0 0
25h0p 5.81+5  1.94-15 D 0 G 0 0
25335 1,149 0 0 0 0 0 0
Total T.23+11  3,25+7 1.61+7 7.91+6 2.39+6 1.84+6 1.61+5

 

%Read as 5.88 x 102.

b

At 150 deys after discharge, 99.5% of Pu + U was extractead.
79

APPENDIX III: HAZARD REDUCTION ACEIEVABLE BY ENHANCED
REMOVAL OF ACTINIDE ELEMENTS (REF. 19)

On the basis of the duration of the long-term hazsrd, there is
potential merit in the use of separations processes thet will greatly
reduce the losses of uranium and plutonium to the high-level waste and
separate the high-level waste into fissior product and actinide fractioms.
The required separations processes have not yet teen developed, hut it
does appear that secondary trzatment processes can be developed that will
permit overall recovery of perhaps 99.999% of the wranium, 99.995% of the
plutonium, 99.95% of the neptunium, and 99.9% of the americiuvm apé curium.
Also, approximately 99.5% of the iodine. containing the long—lived'isotope
1291, can potentially be separated Ifrom the remaining fission products on

the basis of its high relative volatility.

Figure 4 and Table 27 illustrate the possible merits of these azepar-
ations by comparing the hazard index* and the hazard measure, respectively,
of the wastes from conventional processing of fuel Ifrom a typical light
water power reactor with those resulting from the postuliated secondary
treatment. For reference, Fig. 4 also shows the hazerd index thet mey
be assoclated with the miperai pitchblende (which occwurs naturally as
pebbles znd rocks in Africa and Cenade) and a uranium cre containing 0.2%
U3O8 (which is typical of the large deposits that cccur in the Colorade
Platea). The hazard index of the waste from conventional reprocessing
decreases rapidly over the first 1000 years {(due to the decay of 905r
and l3705) but remains greater than that of pitchblende for periods
exceeding one million years. The waste resulting from secondary treat-
ment, however, has a hazard index that falls within the range"qf zaturelly
occurring radioactive materials after only several hundred years, a time
span for which it is possible to make relianle extrapolations of the
effects ol geologic, climatic, and other natural phencmena, As can be
seen from Fig. 5 and Table 28, similar conclusions can be drawn if these
separations are spplied to wastes Irom reprocessing ILMFBR fuels. In each
case, the RCGs caleunlated by LaVerne9 were used in place of the default

values recommended in 20 CFR 20.

 

%
Thie hazargd index is based on 2 unit volume of waste. This permits a
direct comparison with the potential hazerd of uranium ores on an
in-situ basis.
 
&o

ORNL DWG.72- 11707

 

 

104 ’
!m» -
 wASTE FROW
o % CONVENTIONAL -
pnocsss:b ]
10% pircusLenne ST
(60% U} §

- WASTE WITH

s SECONDRRY TREAT-
Te) ?:. MENT AFTER SUS-

: TAINED RECYCLE OF

1 HODINE AND ACT-

MEOE S —.
WASTE FROM
10 & COMVERT IONAL
RESROCESSING PLUS SECON- .
DARY TREATHMENT FOR REMOWAL A
OF JODINE AMD ACTINIDES -~
10 % TYPICALU URANIUM ORE =

Bl W e W v AR AR WS Aol K Sl R SR s Mg S W NR P A R a

 
    
 
  

 

 

r . 3

 

HAZARD INDEX , vol of woter 9t RCG/ vol of waste or ore

o ot 0w ot W
AGE OF WASTE, years

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1012} <

HONE ) J

ol

10 WASTE FROM 4
CONVENTIONAL
REPROCESSING

09} -

(60%U) | waSTE WiTh SECOND-

ARY TREATMENT AFTER
o7 SUSTAINED RECYCLE OF |
10" IODINE AND ACTINIDES

106 /

IWASTE FRCOM CONVENTIONAL —
REPROCESSING PLUS SECONDARY
TREATMENT FOR REMOVAL 2F
ICDINE AND ACTINIDES

109 TYPICAL URANIUM ORE 1

I A A AP AN el Ny S AR APy Ll Ay R algl

  
 

HAZARD INDEX, vol of water ot RCG/vol of waste or ore

 

 

 

o 102 0 0% 10° s
AGE OF WASTE, years

Fig. 5. Effect of BAze and Method of Treatment on the EHazard Index
of High-lLevel Wastes from Repiocessing LMFBR Fuels.
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vWhile the indicated separations of actinides and fission products
appear to be feasible, it is apparent that a large development progrem
would be reguired to reduce these concepts to practice. Separations
processes that are used currently, wkhich utilize solwvent extraction and
ion exchange for recovery of small quantities of certain velushle anti-
. 2hi 252
nides ( Cm,

given mmericium, curium, and neptunium recoveries in the ranges of 90

Cf, etc.) from speciaily irradieted materials, have

to 99% and plutonivam recoveries of akuout 99.5%. Since these processes
were originally intended for recovery of the actinide velues and not
for alleviation of the high-lewvel wasie problem, little or no effort
was xade Yo achieve higher recoveries or to reguiste the amounts and
types of chemical reagents for cptimm waste handlinpg.

In order tc evaluate the practicaliiy of sush 2 waste management
system, a comprehensive development program is needed (1) to solve
problems that are obvious frum past experience, (2; to increase the
repoval of these actinide: to the desired level, (3) to determine the
nost desirsble method to integrate the needed process ¢ycies inte cre
overall system, (4) %o choose chemicel processes and reagents that
minimize high-lev2) waste problems, and (5) to Getermine the ccmposi-
tion of intermediate-level waste stresms thet will be generated and
weys to recycle these streams.