Upload
others
View
2
Download
0
Embed Size (px)
Citation preview
Dissolution of high-level nuclear waste solids
Item Type text; Thesis-Reproduction (electronic)
Authors Voss, James Wilson, 1954-
Publisher The University of Arizona.
Rights Copyright © is held by the author. Digital access to this materialis made possible by the University Libraries, University of Arizona.Further transmission, reproduction or presentation (such aspublic display or performance) of protected items is prohibitedexcept with permission of the author.
Download date 09/03/2021 15:50:49
Link to Item http://hdl.handle.net/10150/348047
DISSOLUTION Of HIGH-LEVEL
NUCLEAR WASTE SOLIDS
: . \ ' by
James Wilson Voss
A Thesis Submitted to the Faculty o f the
DEPARTMENT OF NUCLEAR ENGINEERING
In P a rtia l F u lfillm e n t o f the Requirements For the Degree o f
MASTER OF SCIENCE
In the Graduate College
THE UNIVERSITY OF.ARIZONA
1 9 7 6
STATEMENT BY AUTHOR
This thesis has been submitted in p a rtia l fu t f in m e n t o f requirements fo r an advanced degree a t The U n ivers ity o f Arizona and is deposited in the U nivers ity L ibrary to be made ava ilab le to borrowers under rules o f the L ib rary.
B r ie f quotations from th is thesis are alTowable w ithout special permission, provided th a t accurate acknowledgment o f source is made. Requests fo r permission fo r extended quotation from or reproduction o f th is manuscript in whole or in part may be granted by the head o f the major department or the Dean o f the Graduate College when in his judgment the proposed use o f the m aterial is in the in te res ts o f scholarship. In a l l other instances, however, permission must be obtained from the author.
SIGNED:
APPROVAL BY THESIS DIRECTOR
This thesis has been approved on the date shown below:
R. G. Poste c / " f y c
DateProfessor o f Nuclear Engineering
ACKNOWLEDGMENTS
I would Tike to acknowledge the assistance which I have
received from Dr* Norman A. H i!b e rry$ Dr. Morton E. Wackss and
Dr. Roy G, Post. A ll have provided invaluable guidance in th is
work. I would also l ik e to acknowledge the patience o f my
colleagues> especia lly John Boegel, as they gave me many hours
o f th e ir time providing he lp fu l advice. F in a lly s I acknowledge
the strength and patience o f my w ife , Sandra Lee, as she helped
me through th is portion o f my career.
TABLE OF CONTENTS
Page
LIST F TABLES ® ® « * * « » » • » * « e * * * o « « o
LIST OF TLLUSTRATIONS . . , . .. . . , % ............................. .... . v i
ABSTRACT , , . . . . . . . . . . . . . . . v i i
1. INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . 1
2. ANALYTIC EXPRESSION RELATING TO RELEASES OFRAOTOISOTOPES FROM WASTE FORMS . . . . . . . . . . . . . , . 3
Solution Behavior o f Waste Forms . . . . . . . . . . . . 3Temperature D is tr ib u tio n in Waste Solids . . . . . . . . 8
3. RADIOISOTOPE RELEASE FROM WASTE FORM WITHSUBSEQUENT ENVIRONMENTAL TRANSPORT . . . . . . . . . . . . . . 1 2
4. NUMERICAL EXAMPLES . . . . . . . . . . . . . . . . . . . . . . 17
PUREX Waste Description . . . . . . . . . . . . . . . . . 1 7Description o f Reference B o ros ilica te Waste Glass . . . . 21Primary Radioisotope Release . . . . . . . . . . . . . . 25Radioisotope Transport through the Environment . . , . . 27
5. DISCUSSION AND CONCLUSIONS . . . . . . . . . . . 4 . . . . , . 3 4
APPENDIX A: LISTING OF MAJOR RADIOISOTOPE COMPOSITIONIN PUREX WASTES . . . . . . . . . . . . . . . . . 39
APPENDIX B: WEIGHT FRACTIONS OF MAJOR FISSION PRODUCTRADIOISOTOPES . . = . . . . . . . . . . . . . 42
APPENDIX C: WEIGHT FRACTION OF ACTINIDES INBOROSILICATE GLASS . . . . . . . . . . . . . . . . 44
APPENDIX D: TABULATION OF MAXIMUM PERMISSIBLE CONCENTRATIONS (MPC) OF VARIOUS ISOTOPES IN WATER . . . 45
APPENDIX E: RADIOISOTOPIC RELEASES FROM WASTE FORMSVITREOUS FORM, LEACHANT AT 298K . . . . . . . . . 47
i v
V
TABLE OF CONTEMTS-^Contimied
APPENDIX F: RADIOISOTOPIC RELEASES FROM WASTE FORMSDEVITRIFIED FORM, LEACHANT AT 298K, . ,
APPENDIX G: RADIOISOTOPE RELEASE FROM WASTE FORMSVITREOUS FORM, LEACHANT AT 372K , ,
APPENDIX. H: RADIOSOTOPIC RELEASES FROM WASTE FORMS ,MODELED FORM AT ONE DAT OF LEACMING,WASTE AGE — 10 YEARS V ? .
APPENDIX I : NDMERICAL UNITS , . . . . . V . . . ,LIST OF REFERENCES , . . . . . . . . . . . . . , . .
SO
53
56
58
59
LIST OF TABLES
Table : , ■ Page
I . Relative D issolution o f Elements in Glass . . . . . . . . . 5
I I . Chemical Content o f PUREX Process Liquid Wastes . . . . . . 19
I I I . Heat Generation Rate o f PUREX Waste . . . , . . . . . . . . 20
IV. Composition o f Reference Borosi1ica te Glass . . . . . . . . 21
V. Solution Rates o f Reference Borosi1ica te Glass , . . . . 24
VI. Radioisotopic Releases from Waste Forms . . . . , . . . . . 28
V I I . Major Radioisotopes Released from Ten Year Old WasteForms . . . . . . . . . . . . . . . . . . , . . . . . . . . . 29
VIIT. Ion Exchange Holdup Factors fo r "Typica l" Western U.S.Desert Soil . . . . . . . . . . . . . . . .... . . . . . . . 32
IX. Concentrations o f Radioisotopes Dissolved fromD e v itr if ie d Reference Waste as They Are Transportedthrough the Environment . . . . . . . . . . . . . . . . . . 33
v i
LIST OF ILLUSTRATIONS
Figure
T. Steady State Temperature (K) versus Radius (M) fo r Ten Year Old Waste . ................ ... ........................................
ABSTRACT
A work has been done to explore the release o f radioisotopes
from h igh-leve l nuclear waste so lids by d issO fution, I n i t ia l l y , a
generalized expression describing the dissoTutibn o f a waste form
as a function o f time and temperature is derived. Discussion o f the
element sp e c ific d isso lu tion behavior is included. The environmental
transport o f dissolved radionuclides is next discussed. Equations
describing the flow ra te .o f radioisotopes from a waste form, the
d ilu t io n o f radioisotopes as they are transported in a groundwater
environment, and the radioactive decay o f radioisotopes due to ion
exchange w ith the s o ils are developed. F in a lly , the derived
equations are used in a sample ca lcu la tion invo lv ing the d isso lu tion
o f a b o ro s ilica te glass h igh -leve l waste form and the subsequent
environmental transport o f the dissolved radionuclides.
v i i i
CHAPTER 1
INTRODUCTION
Commercial nuclear power has been under development fo r nearly
t h ir t y years. During th is time period, the recycle o f uranium and
plutonium has been considered essentia l to fuel cycle economy and
conservation o f uranium resources. This strategy has been strengthened,
by the lin k between the L ight Water Reactor (LWR) fue l cycle and the
Fast Breeder Reactor (FBR) fue l cycle.
The LWR and FBR fuel cycle re la tio n is a complex one. Some
fuel cycles include a syne rg is tic re la tio n between the two. Others
show the FBR as an evolutionary step past the LWR, w ith the eventual
phase out o f the LWR systems. Within both s tra te g ie s , the need fo r
uranium and plutonium recovery by nuclear fu e l reprocessing isl ?essentia l. 5
The need fo r uranium and plutonium recovery by nuclear fuel
reprocessing may be s a tis fie d by several d if fe re n t means. One Such
method may be the reprocessing o f nuclear fuels by government owned
in s ta lla t io n s . Another p o s s ib ility is fo r the reprocessing step to be
carried Out by commercial reprocessing p lants. Neither is being done
at the present. The choice o f government reprocessing o f spent nuclear
fuel is techn ica lly feas ib le . CommerciaT fue l reprocessing, while
1
being te chn ica lly fe a s ib le , is not le g a lly possible a t the present, as
a va rie ty o f regulatory questions re s tr ic ts the licens ing o f any com
mercial nuclear fu e l reprocessing p lan t.
One regulatory issue which is Currently being addressed by the
Nuclear Regulatory Commission, surrounds the d isp os ition o f h igh-leve l
nuclear wastes. H igh-level nuclear wastes contain nearly a l l o f the
radioactive f is s io n products generated by the fis s io n in g o f nuclear
fue ls . Current reprocessing technology generates these h igh^ieve i
wastes in the form o f liq u id s . -
At present, government regulations require th a t these high-
level l iq u id wastes be s o lid if ie d w ith in f iv e years o f th e ir genera
tio n and be transferred to a Federal Government Nuclear Waste
Repository w ith in ten years o f 1iqu id waste generati on. While
regulations state tha t the s o lid if ie d h igh-leve l waste must be
"chem ically, the rm a lly , and ra d io ly t ie a lly s ta b le ," i t has been
determined th a t S pec ific waste form properties are needed to ensure
, that, the environmental impact o f h igh-leve l nuclear waste disposal be
pred ictab le and environmentally acceptable.
For complete environmental assessment o f the impact o f nuclear
waste d isposa l, a great deal o f inform ation is required about nuclear
waste form behavior. Two items whiCh are w ith in th is need are the
descriptions o f waste form d isso lu tion and o f environmental transport
o f radionuclides. This work addresses these two subjects by developing
generalized models fo r each phenomenon.
CHAPTER 2
ANALYTIC EXPRESSION RELATim TO RELEASiS OF ;RADIOISOTOPES FROM WASTE.FORMS
Radioisotopes may be released fTom a waste form by d isso lu tions
v o la t iliz a t io n s and p a rticu la te d ispersion. This work analyzes the dis»
so lu tion o f waste sol id .
This d isso lu tio n meG.hanism has been re ferred to as the leaching
mechanism. However, the act o f leaching implies oply the percolation
o f a liq u id about a s o lid which says nothing o f the mechanisms involved
in the leaching process. This work views the act o f leaching as the
so lu tion or d isso lu tion o f waste form.
Within th is Sectionj one model characteriz ing the d isso lu tion
o f s o lid is developed. A second discussion evolves a model re la tin g
the d isso lu tion o f a waste form to the environmental transport o f d is
solved radionuclides.
Solution Behavior o f Waste Forms
. The d isso lu tion o f waste forms is a complex phenomenon. TheA
so lu tion process is both temperature and time dependent. In add ition ,' 5the so lu tion process is element s p e c ific ; The element sp e c ific behav
io r is discussed f i r s t .
I t has been observed tha t a lk a li metals, being h igh ly so lub le ,
dissolve most read ily from waste s o lid s .5 ’ ® In a d d itio n , s o lu b il i ty is
seen to decrease:as ion ic radius increases w ith in a valence s ta te .
. : ■: ' ' ■ 3 '
these observations are consistent w ith experimental evidence revealing
advanced d isso lu tion behavior o f sodium and cesium from waste so lid s .
In: contrasts cerium, an abundant element in h igh-leve l nuclear wastes,
is seen to dissolve from waste so lids a t a rate much slower than the 5a lk a li metals.
Mendel compiled a tab le demonstrating the p re fe re n tia l disso
lu tio n o f various elements from po ten tia l waste glasses. Mendel
studied experimental d isso lu tion resu lts obtained by several d iffe re n t
researchers. While several d iffe re n t experimental techniques were used,
the comparison o f atomic ra tio s o f various elements in the glass forms
and in the solutions lead to Mendel1s resu lts shown below in Table I .
The time dependent behavior o f the d isso lu tion o f waste formsshas been described by equation ( I ) *
■ L = A e " 1/2 + B • ( 1)
where L = so lu tion rate o f waste form
8 = time
A = ra te constant re la ted to d iffu s io n o f io n ic species through waste m atrix
B = rate constant re la ted to corrosion o f waste m atrix
Equation ( ! ) is seen to have a S in g u la rity a t e = 0 sec. Since experi
mental measurements to which th is expression has been f i t tend not to
be fo r times less than two hours, the equation is considered not to be
va lid fo r these small times.
This expression has been fu rth e r extended.to the arrhenius form
in equation (2) . ^
5
fab le I . Relative D issolution o f Elements in Glass,
Glass Type Relative D issolution Reference
S ilic a te glass Gs>Sr>Fe>Zr Eliassen and Goldman^
Lead glass
Phosphate ceramic
Cs>Al>Pb>SrsCe :
Nas Cs>Sr>Zr-Nb>Ru,Ce
PaigeR' - ' g ■' ' : . Allemann
Boros11ica te glass .
Borosi1ica te g1 ass
Gs a Na>Zr-Nb>Ru»Ce
Na>Si>B>Cs>Sr>Ge 3 Tb
QAllemann
E l l io t and Auty®
BorosiliCate glass
Phsophate glass
Na>B>Si
Cs>Na>Sr>Ru>Ce
Heimerl e t al . ^
Mendel and McElroy^
where a»b = constants m atrix speGffie)
AHd - activation-energy o f d iffu s io n {element and so lid m atrixsp e c ific )
AHc = a c tiva tio n energy o f corrosion (element and so lid m atrixsp e c ific )
R - gas constant
T - absolute temperature o f so lid m a trix -so lu tion in te rface .
Equation (2) then predicts th a t a t short tim ess the d iffu s io n
o f radionuclide species from the waste surface is the dominant e ffe c t.
A fte r long times s the corrosion o f the waste m atrix becomes dominant.
The separation between short and long times is on the order o f 10 sec-
onds, based s t r i c t l y on an order o f magnitude argument.
The Constants a and b are dependent upon several fac to rs .
These include so lu tion v e lo c ity , so lu tion pH, so lu tion chemical compo
s it io n , and surface condition o f so lid m atrix.
Addressing the surface condition in p a rt ic u la r , i t is observed
tha t the surface energy o f p a rtic le s on a jagged surface is higher than
fo r a smooth surface. Thus, less energy is required to remove those
p a rtic les on a jagged form. This means th a t i f a piece o f a waste form
has a very rough surface, i t w i l l d issolve fas te r than a piece w ith a
smooth s urface.
This surface condition is o f special importance in regard to
measurement o f so lu tion rates Of waste forms. Solution ra te measure
ment methods vary w idely, ye t nearly a l l require a high degree o f waste
form subdivis ion. This may include the grinding Of a waste m atrix to
- - - ; : . . : ' ' \ ^ : / - i ^ ;■ - ; - " ^ ::: 7
a powder3 w ith subsequent d isso lu tion o f a specified set Of waste par
t ic le sizes. Then, in l ig h t o f the previous discussion, i t is apparent
tha t the degree o f suhdivisiOn o f a waste m atrix w i l l have an e ffe c t on
the resu lts o f the so lu tion rate measurement.
The various s o lu ti on rate measurement methods vary the soTutioh
flow rates. Some techniques have s ta t ic so lu tions , some have cycled
so lu tions , w hile others have continuously flow ing so lu tions. This too
w i l l a ffe c t so lu tion ra te measurement re su lts .
Results obtained from the various measurement methods vary.
, Most methods are reproducible w ith in a fa c to r o f ten, ye t since the
values obtained from d iffe re n t methods on s im ila r waste forms vary, the
meaning must be questioned, the d ifferences mentioned above along w ith
many others account fo r re su lt va ria tions .
Calculations using the resu lts o f any so lu tion ra te measure
ments must be performed w ith caution. Recognition o f weaknesses in the
test methods and thus so lu tion rates is demanded.
The actual meaning o f so lu tion ra te measurements is usually
qu ite c lear. As most methods tend to create a set o f “worst possible"
d isso lu tion eonditions, measured so lu tion rates generally serve as up
per 1im its . Thus, ca lcu la tions made using measured values re f le c t th is
same upper T im it behavior.
One p a rtic u la r measurement o f the so lu tion ra te o f b o ro s ilica te
glass, done at B a tte l!e Northwest Laboratory, has been performed to
determine the time and temperature behavior o f the so lu tion ra te . The
measurement has been done using the so ca lled Dynamic Method in which a
1 gram sample o f -45 +60 mesh p a rtic le s (p a rtic le s o f mean diameter be
tween 0.56mm and 0.42mm) o f glass are placed in a s ta in less stee l "tea
bag", and Suspended in d is t i l le d water a t various temperatures. The
water is continuously ag ita ted about the sample and changed a t frequent
in te rva l s. Data from th is measurement have been f i t to equation (2) by
the Least Squares Method to determine the various constants. The v a l
ues obtained are shown below:
a = 1 . 6 2 X 1 0 5 ug/(m2s1/ 2)
b = 2.12 X 103 yg/(m2s)
AHd = 3.0% X 103 J/mole
AHc - 1.20 X 104 J/mole
Temperature D is tr ib u tio n in Waste Solids
The energy balance equation predicts the temperature o f a so lid
with time, as Seen in equation (3).
vkvT+q'' ' = pc (3)
where T - temperature (K)
e - time (s)
K - thermal conductiv ity o f s o lid (W/mK)
q ' ' ' === volumetric heat generation ra te (M/m3)
p = so lid density (kg/m3)
c = s o lid sp e c ific heat (J/kgK)
The surface temperature o f the waste so lid becomes o f fundamen
ta l importance in determining the so lu tion rates o f waste s o lid s , as
expressed e a r l ie f. Determination o f the surface temperature requires
deta iled knowledge o f the environment which the waste w i l l be placed
in to . Even i f a l l necessary inform ation is known9 the descrip tive so
lu tio n o f equation (3) is not a t r i v ia l one.
The energy balance equation has been solved fo r One p a rtic u la r
steady sta te case. A waste cy linde r o f varying thermaT properties des
cribed below is buried in some s o il w ith a l l can ister she ll material
and overpack neglected.
Figure 1 shows the steady sta te temperature d is tr ib u tio n fo r a
waste cy lin d e r and the close proxim ity s o il. The fig u re was developed
fo r a waste form w ith a heat generation rate o f 15,420 W/m which cor
responds to the heat ra te o f a ten year ou t-o f-rea c to r b o ro s ili cate
glass form. The radius o f the cy linde r is 0.1525m, seen to be a t the
in te rsec tion o f the two curves. The ambient temperature is 300K. The
points labe lled 1 corresponding to TEMPI are fo r the thermal conductiv
i t ie s o f the waste form and s o il both 1.OW/mK. The points labe lled 2
corresponding to the variab le TEMPI0 are fo r a waste form conductiv ity
o f 10.0 W/mK and a s o il thermal conductiv ity o f 1.0 W/mK.
The energy balance equation was solved fo r th is ca lcu la tion by
f i r s t assuming th a t the steady s ta te condition existed. The geometry
considered was th a t o f an in f in i te r ig h t cy linder. Four boundary con
d itio n s were used since the problem is o f two regidnS, one region w ith
and one region w ithout heat generation. The f i r s t boundary condition
is tha t the f i r s t rad ia l distance de riva tive o f temperature is zero a t
the center o f the cy linde r. The second and th ird are co n tin u itie s o f
■ ; 10
heat f lu x and temperature a t the c y lin d e r-s o il in te rfa ce . The la s t
boundary condition is the ambient temperature at some distance from the
cy linder.
One meaning o f th is type o f ca lcu la tion , aside from the disso
lu tio n im p lica tions , is re la ted to phase changes which various waste
forms undergo a t elevated temperatures. As discussed la te r , d e v i t r i f i
cation o f a glass waste form may occur a t a rapid ra te w ith elevated
temperatures, as i t s rate is described by an arrhenius equation.
V Another impTication o f th is Calculation is th a t fo r heat con
duction to the environment, the thermal conductiv ity o f the waste form
does not change the in te rface temperature or the temperature p ro f ile in
the surrounding environment. In the f ig u re , the two temperature curves
are seen to merge a t the waste form surface, and Coincide from tha t
po in t on.
A ll o f these fac to rs , phase o f waste form, surface temperature
o f waste form, and temperature p ro f i le in the close proxim ity o f the
waste form, have a marked e ffe c t on the conditions fo r d isso lu tion .
The constants in the d isso lu tion equation w i l l be d if fe re n t fo r each
phase o f each form. Also the temperature fo r d isso lu tion w i l l have
d ire c t e ffe c t on the d isso lu tion rate o f a waste form.
11
1=TEMPI 2=TEMP1C
Figure 1
6.50
6 .03
4 .50
4 .00
3-50
2 .50
0 .00 0 .10 0.20 0.30
Steady State Temperature (K) versus Radius (M) fo r Ten Year Old Waste
CHAPTER 3
RADIOISOTOPE RELEASE FROM WASTE FORM WITH SUBSEQUENT ENVIRONMENTAL TRANSPORT
The mass flow ra te o f the radioisotopes in to the environment *
m, from a waste form is the product o f the so lu tion rate o f the sOTid,
L, and the exposed surface area o f the s o lid . A, provided the mass flow
is independent o f the concentration o f radioisotopes in the Solution^
m = L * A . (4)
Thus, the concentration, C , o f radioisotopes in a so lu tion ' • ' 0 : .
which is Teaching the waste form is the quotient o f the rad io isotope;
mass flow rate and the volumetric flow rate o f the so lu tio n , V^, as
shown in equation (5 ), ’
r - 2L - L ACo ' Vf ~ -Y J - (5)
I t is important to recognize tha t Co w i l l vary w ith time and
waste surface temperature ju s t as the so lu tion rate does. Calculations
which fo llow in th is work w i l l consider constant so lu tion rates and
thus constant concentration, CQ. However, recognition o f the actual
time and temperature va ria tion is made.
Once radioisotopes have been dissolved from waste form, th e ir
subsequent environmental transport may be estimated, and a n a ly tic a lly
described. Equation (5) expresses the concentration o f io n ic radionu
c lide species in a d isso lv ing so lu tion as i t re la tes to the so lu tion
12
■ '■■■ : : -"V . •. ■ ■ . : . . ;■ is. rate o f a waste form. As the radionuclides are transported through the
environment th e ir concentrations are reduced by several in te ra c tio n s ,
o f which only d ilu t io n and exponential decay are considered in th is
work.
Considertng radionuclide d ilu t io n f i r s t , i t is seen by equation
(6) tha t the concentration o f radioisotopes whioh reaches man, C, is
expressed as the concentration a t the waste form, CL, d ivided by the
dilution factor, 0. :
■ c = c0/D • ; (6)
This d ilu t io n fa c to r is the number o f equivalent volumes o f un-
contaminated water th a t the radionuelide mixes w ith during environmen
ta l transport.
Concentrations o f radionuclides are fu rth e r reduced as the
radioisotopes decay exponentia lly w ith time during environmental trans
port. Thus fo r the 1 th rad io isotope, the concentration which reaches
man, C|, is expressed as in equation (7).
C1 = C01 (e ' " h / D ( 7 )
where C = the concentration o f i a t the waste form
A.. = the decay constant of i
t . = the time required fo r i to be transported to man
The transport time is found by d iv id in g the distance which
radionuclides must trave l to reach man by the speed a t which the radio
nuclides move through the environment- The radionuclide ra te o f tra ve l
is somewhat less than the groundwater speed. The reduction o f the
speed is caused by the ion exchange o f the radionuclides as they trave l
through some aqu ife r m ateria l.
Ion exchange is a phenomehon in which some dissolved species
are held by ion ic bonds a t s ite s w ith in a s o lid ; the s o lid being aqui
fe r m aterials fo r the case o f radionuclide transport. There is a com-
p e tit io n fo r the dissolved io n ic radionuclide species as they are moved' ' IA - ' - : • ' " ' ■ '/ - ' ••• ■ . '
through the biosphere. Thus, the dissolved species w i l l be continu-" ■ ■ • • ■ is ■;
a l ly exchanged through environmental transport.
The problem a t hand is to re la te the chemical behavior o f the
ion ic radioisotope species to th e ir transport. I f M+m and A+n repre
sent ions o f radionuclides and n a tu ra lly occurring exchangeable aqu ife r
minerals respective ly carrying charges +m and +n, and X is assumed to
be a reactive Chemical rad ica l in aqu ife r m ate ria l, the equation fo r
the ion exchange reaction is as in equation (8).
nM+rri + m(AXrt) + n.(.MX ') + mA+n (8)
The equ ilib rium constant fo r th is exchange, K, is defined by
equation (9).
[ « g n n C " f (9 )
[M+m] n [AXn] m
Since in r e a l i ty , a very small fra c tio n o f the ava ilab le ex
change s ites in the minerals is occupied, [A+n] and [AX ] may benassumed constant. This allows the d e fin it io n o f a useful d is tr ib u tio n
c o e ffic ie n t.
\v t iv«. i v- « u i w i v i w i c v i u u i w.i i
(1 0 )
W = the weight o f minerals
Consider, then, a species o f rad ionuclides, i n i t i a l l y lo ca liz e d
to a waste form. The flow ra te o f the rad ionuclides, F_, may be ex-• * a
pressed as in equation ( 11),
Since p is not known exactly , i t is convenient to use a value
o f 1, the usual value being 4 or 5. Thus, by assuming in addition th a t
the area o f flow fo r the radionuclides and groundwater are equal,
equation ( 11) reduces to equation ( 12).
where V - the ve lo c ity o f the radionuclides
V = the ve lo c ity o f the groundwater
I - the ion exchange holdup fa c to r
The transport time o f the radionuclides, t , is then the d is
tance, X, which radionuelides must trave l to reach man divided by the
ve lo c ity o f the radionuclides in so lu tion .
where F, = the flow rate o f groundwaterwp = the ra tio Of weight o f minerals to volume o f water
per u n it volume o f aqu ife r material
Thus, the concentration o f each radionuclide transported to man may be
Calculated.
fli l(TX)C = f ^ e V (1 4 )
f ' • . .
Given equation (14), a sp e c ific waste form w ith i t s d isso lu tion
ch a ra c te ris tics , and a maximum concentratioh o f radionuclides which may
reach man, the required is o la tio n o f rad ioactive wastes may be calcu-
la ted. This required is o la tio n would be expressed in terms o f the d is
tance from mans X, and depends on the Velocity o f water through the
environment, Vs the sorption cha rac te ris tics o f the environment. I , and
the quantity o f water which may flow through the environment, V^-D.
A note must be added about th is environmental transport model.
Many sophisticated computer codes modeling th is behavior have been
w ritte n and are in use. This simple model has been derived fo r ease in
hand ca lcu la tions . Assumptions have been made in th is theory. One is
in assuming bulk quan tities such as s ing le sorption cha rac te ris tics o f
an environment. This may be true in a small increment, but ce rta in ly
is not true on a Targe scale. The radionuclides have been modeled as
moving only in one d ire c tio n , while in fa c t they w i l l tend to spread in
a manner s im ila r to m ateria l in a plume dispersion. The important fac
tors in constructing th is simple model are th a t hand c lacu la tions may
be performed, and w ith proper se lection o f averaged p rope rties , the
resu lts w i l l be conservative.
CHAPTER 4
: NUMERICAL EXAMPLES : ^
In order to demonstrate the use and meaning o f the ana ly tic
expressions previously presented, a series o f sample oalculations w i l l
be performed oil a typ ica l waste form. The ca lcu la tions w i l l include an
analysis o f rad io i sotope release by s o lu ti on $ and a sample analysis o f
radioisotope transport through the environment.
The type waste form which w i l l be considered is b o ro s ilica te
glass produced from PUREX waste. This section w i l l q u a n tita tive ly de
fine the b o ro s ilica te glass and the PUREX waste whichs a fte r c a lc i
nation, is mixed w ith the glass.
In add ition , the b o ro s ilica te waste form f i t t e d to the so lu tion
rate expression in Chapter 2 is studied in terms o f i t s radioisotope
release and subsequent environmental transport. For ca lcu la tiona l pur
poses , i t is assumed th a t the radionuclide content to be specified fo r
the reference waste form w i l l e x is t in the modeled form*
■ PUREX Waste Description
This analysis considers nuclear wastes generated by the LWR
fuel cycle* While in the reacto r, the LWR fue l w i l l acquire a f is s i le
fuel burn up o f 2.85. TJ/kgU (33000 MWD/MTU) over a three year p e r io d .^
The fue l considered.w ill have been removed from the reactor 160 days
p r io r to reprocessing.
17
The nuclear fuel is processed in the PUREX process. This PUREX
process uses concentrated n i t r ic acid in aqueous so lu tions and t r ib u ty i
phosphate in kerosene to separate uranium and plutonium from the f i s
sion products in spent reactor fu e l. The re su lting liq u id wastes are
n itra te s . The chemical content o f these liq u id wastes is described in 18
te I I .
The h igh-leve l PUREX wastes are h igh ly rad ioactive and s e lf
heating. The to ta l fis s io n product a e t iv itv in the l iq u id waste isq a ' ' '
0.451 EBq/m (1.22 x TO C i/1 ). The to ta l actin ide a c t iv i ty , assuming - \ ■ . ; . • . - '
0.5% loss o f uranium and plutonium in reprocessing, is 4.28 nn~
(115.8 C i/1 ). Thus the to ta l radioisotope content is 0.455 EBq/m^
(1.24 x 104 C i / 1 ) T h e major radioisotope composition in the PUREX
waste a t 160 days o u t-o f-re a c to r is lis te d in Appendix A. In add ition ,
in Table I I I , the volum etric heat ra te o f PUREX waste as i t varies w ith
time is l i s t e d . ^
The analysis next considers th a t the PUREX liq u id waste is
s o lid if ie d by the ca lc ina tion process. This process drives the waste
n itra te s to oxides, removing the n itra te s and water as o ff-g a s . The
c a l c i n e d product is then suspended in a glass m atrix , s p e c if ic a lly in
a boros H i cate glass m atrix. This fo rm .is discussed in the next
section.
19
Table I I . Chemical Content o f PUREX Process Liquid Wastes*
Component Concentration (kg/m^)
hydrogen 1.058
iron X v: ' .n ickel : 0.265
chrpnium 0.529
n itra te 174.1■
phosphate 2.380 :' • ' a
uranium 12.70
plutonium . . ' 0.106
neptunium 1*270
americium 0.370
curium 0.106
Total f is s io n products 76.19
*based on a waste production rate o f 3.78 x lC fV /k g l l
20
Table I I I . Heat Generation Rate o f PUREX Waste
Time Out o f Reactor (YR) Heat Generation Rate (W/m3)
160 days 5.04 1+041 2.73 E+04
-5:'-. ; ' / ' ■: ' 5.05 E+0310 2.91 E+0320 2.06 E+0330 ■ v-v' .. 1.60 E+03 :40 1.25 E+03
’ so 9.80 E+0260 7.70 E+0270 6.10 E+0280 4,80 E+0290 3.80 E+02
TOO 3.00 E+02120 1.90 E+02140 1.30 E+02160 8.54 E+01180 5.91 E+01200 4.29 E+01220 3.25 E+01240 , 2.58 E+01260 2.14 E+01280 1.84 E+01290 1.74 E+01
V ' . ; . 21
/ v :Descr1ption o f Reference B pros llica te Waste Glass
A b o ro s ilica te glass contains and SiO^ as i t s major
cdnstituents. The boros11ica te glass considered in th is analysis is
Savannah River P lan t's Mix #18. I ts chemical composition is lis te d
in Table I f . 20
Table IV, Composition o f Reference BorosiTicate STass
Compound Wt% o f Glass F r i t
s122 ; ' t : ' . I 52,5
Na20 22.5
B203 10.0
Ti02 10.0
CaO 510
Production requirements fo r bo ro s ilica te glass vary from
1.4 x TO*17 m3/J (1.2 1/1000 MWD) to 5.8 x TO*7 m3/J (5.0 1/1000
This analysis w i l l consider a production ra te o f 2.3 x ld “37m /J
(2.0 1/(1000 MWD)). For th is value, waste oxides are about 22 wt% o f
the b o ros ilica te glass product.
The heat generation rate o f the b o ros ilica te product is
calculated from the heat generation ra te o f the PtiREX waste as shown
... 'V -';;--; : : ; : : . .. ; ■ - ■ 22
in Table I I I . The glass heat generation rate is calculated from
equation (15). ; \
% = <i P V tBV ■ ( , 5 )where q^ = the heat generation rate o f b o ro s ilica te glass (W/m)
qp = the heat generation rate o f PUREX waste (W/m^)
- the production ra te o f PUREX (m^/kgU)
B = the f is s i le fuel burnup o f the fue l (J/kgU)
Rh = the volume o f b o ro s ilica te glass produced per energy generated (m /J )
In th is ana lys iss Rp = 3.78 x 10“ 4 m3/kgUs B = 2.85 TJ/kgU (33000 MWD/
MTU), and Rp is 2.315 x 1 0 "^ m^/J. Hence, qp is described in equation
(16).
qb - q0 ; 3 M -------- .) ,5 .30 (16)p m PUREX waste m buro-glass
Thus, to determine the heat generation rate o f the b o ro s ilica te glass
products the data in Table I I I need be m u ltip lie d by 5.30, as shown
above.
The reference b o ro s ilica te glass has several measured proper
tie s which held describe i t . Solution rates o f th is glass under v a r i
ous conditions are shown in Table ,22
The constants fo r the d isso lu tion equation calcuTated e a r lie r
are used w ith the d isso lu tion equation to calcu late d isso lu tion rates
fo r use in th is section. Temperature conditions selected are 298K and
372K as the d isso lu tion temperatures. The time fo r d isso lu tion is a r
b i t r a r i ly selected to be one day. In th is manner, d isso lu tio n rates
are calculated and shown in Table V.
23
Table V . - So lu t i on Rates o f Reference Boros11ica te Glass
Component Leaching Solution Rate «Leached Condition (ug/mzs) (atoms/m s )*
137cs r 0.013898Sr 1 0.0467
125Sb ./ i
0.261 n h9ma ipria aCLi n i aes
• bulk glassi
1U o U d-U I
603.0 3.96 £+18'137CS : : 2: / : 0.019090Sr 2 0.832 « ■*
; ,25sb ■ 2 . o0.230n m n/ia ipfia: aCuinluGS
. bulk glass. ; c <
2W o u 1U4
535.0 3.50 E+18 ;137Cs 3 0.39490Sr 3 0.749
125Sb 3 1.01alpha actin ides ■ 3 0.0435bulk glass 3 2350,0 1.54 E+l9 ,bulk glass 4 61.2 . 4.01 E+l7modeled glass modeled glass
5 176.3 1.15 E+18• ' • 6 248.0 1.62 E+18
Conditions:
1 - glass in H20 a t 2S8K (23°C)
2 - glass heated to 773K (500°C) fo r 1 month 3 in H O a t 298K (25°C)
3 - glass heated to 873K (600°C) fo r 1 month ( d e v i t r i f ie d ) s in H90a t 298K (25°C) d
4 - glass in H O a t 372K (99°C)
5 - glass in H O a t 298K (25°C) surface temperature fo r e - 1 day
6 - glass in HgO a t 372K (99°C) surface temperature fo r e = 1 day
*Based on a calcine molecular weight o f 190g/mole and a waste glass molecular weight o f 92g/mole.
A few additiona l properties o f the reference.waste glass are
known. These are described below.
The density o f any waste glass varies w ith waste:oxide content.
A ll data ind ica te th a t expected densities w i l l range from 2900 to
The thermal s ta b i l i t y o f b o ro s ilica te glass is well known.
Glass undergoes two steps to thermal i n s ta b il i ty * d e v i t r i f i ca ti on and
m echan ica l;in s ta b ility .
D e v itr if ic a tio n is a phase transformation in glass, in which
the supercooled liq u id s truc tu re o f a glass c ry s ta lliz e s . Four main
factors d ire c tly a ffe c t the d e v it r i f ic a t io n o f a glass; 1) time,
2) temperature, 3) nuciea tion , and 4) in te rna l s truc tu re .
I t has been observed tha t the rate o f d e v it r i f ic a t io n may be27expressed in the arfhenius form as in equation (17).
3500 kg/m3. 23924825$26 This work calculates a density o f 3000 kg/m3
fo r the reference glass
to 1.4 W/mK.
The thermal conductiv ity o f b o ro s ilica te glass varies from 1.0 ,/my 23,24,25,26
(17)
where A - constant
E = a c tiva tio n energy
R - gas constant
T - absolute temperature
:: : ; ' ' / ' ; 25
Thus, d e v it r i f ic a t io n is a k in e tic process which w i l l proceed
to some degree a t a l l temperatures. I t may be expected th a t a nuclear
waste glass w i l l eventually completely d e v it r i fy , even a t low tempera-
v; tures. .' ;
The se lf-hea ting nature o f nuclear wastes provide more advanced
temperatures, and thus higher rates o f d e v it r i f ic a t io n . I t has been
observed tha t a t temperatures over 870K, the d e v it r i f ic a t io n o f the
" reference glass proceeds r api dl y . ^^9 Whi1e the e ffec ts o f d e v it
r i f ic a t io n on a glass are varied, data in Table V demonstrate tha t the
reference glass so lu tion rate may increase w ith d e v it r i f ic a t io n :. V ■■ . ■ .;:v 'v ' ■ : . 24
Mechanical in s ta b i l i ty o f glass begins a t about 973K, as the
glass takes on a mol ten behavi or. This behavior becomes more severe
w ith increasing temperature u n t i l a t 1373K to 1473K, the glass behaves
as a viscous l iq u id . ^
Waste form geometry is Of ye t uncerta in i This work assumes a
canister geometry o f 0.30m in diameter by 3.0m long.
Primary Radioisotope Release
The radioisotope release by d isso lu tion is to be calculated fo r
the reference b o ro s ilica te glass. To demonstrate the possible range o f
releases which may re s u lt from the waste formy fourteen d is t in c t cases
w i l l be considered as shown below:
a - v itreous product w ith leachant at 298K fo r c y lin d r ic a l mono
l i t h and fo r monolith broken in to cube p a rtic le s 0.1 mm on
a s ide, w ith waste age a t one and ten years o u t-o f-re a c to r;
b - d e v it r i f le d product w ith leadhant a t 298K fo r c y lin d r ic a l
monolith and: cubic pieces w ith one and ten years o u t-o f-
reactor ages;
c - yitreous product w ith leachant at 372K fo r c y lin d r ic a l
monolith and cubic pieces w ith and ten years o u t-o f
reactor ages;
d - the modeled glass in c y lin d r ic a l monolith form a t one day
o f leaching w ith the leachant a t 298k and 3721.
the ca lcu la tion o f rad io iso top ic release by so lu tio n , u t i l iz in g
the data o f tab le V and the previously described a na ly tic expressions,
requires some prelim inary explanation. Table V l is t s sp e c ific 1 each
rates o f ^% r, ^ C s , ^ S b , alpha ac tin ides , and bulk glass. Those
radioisotopes not s p e c if ic a lly mentioned w i l l be assumed to be released
from the waste form by corrosion. Thus, the actual rad io iso to p ic re
lease is the release o f the waste forms times the weight fra c tio n which
those radioisotopes make up o f the m atrix. These weight frac tions are
found in Appendices 8 and €.
The a na ly tic expressions presented describing the concen
tra tio n s o f radioisqtopes released from waste forms were o f the general
form shown below in equation (5 ).
Co = V ; (5)
where C. = concentrations o f radioisotopes
m - mass flow rates o f radioisotopes
= volumetric flow rate o f leachant
■ 27
To assure tha t calcuTatipns are not weighted by the estimation o f a
volumetric flow rate o f a leachant, the mass flow ra te o f radioisotopes
alone w i l l be calculated in the units o f (MPC-m^/s)s from Appendix D.
Thus, the po ten tia l mass flow rates o f radioisotopes from the/ •
reference waste glass are shown below in Table VI w ith sp e c ific re
leases shown in Appendices E, F, G, and H.
Calculations performed to a rr ive a t values found in Table; V II
show some addi t i onal i nformation o f i n teres tv Those radi o i sotopes
which are o f major in te re s t a t a waste age o f ten years are revealed.
These are tabulated below in Table V II.
Radioisotope Transport Through the Environment
This section w i l l discuss the releases o f radioisotopes pre
viously calculated as they are moved through the environment. In par
t ic u la r , radioisotope transport in a r iv e r type environment and in a
typ ica l desert environment w i l l be considered.
Equation (14) shows tha t concentrations o f radioisotopes a t
some po in t in the environment is as shown below.
For a r iv e r environment, one in which the decay o f radioisotopes during
the transport may be neglected, th is equation may be reduced to the
form o f equation (18).
28
Table VI. Radioisotopic Releases from Waste Forms
WasteGeometry
Waste . Age (Yr)
Leaching Condition
Mass Flow RatOgOf Release (MPC-m /s )
cy linde r 1 1 9.73 E+2
cy linder TO 1 4.75 E+2cubes 1 1 5.84 E+3cubes TO ; i ■■ 2,85 E+3
cy linde r y-;T;2 ; 2.50 E+5 ’cy linder TO ' ' 2 - 3.97 E+3cubes 1 2 1.50 E+6cubes ’ ; ' ■' .V: 2 . 2.38 E+4cy linde r T 1 3 6.58 E+3cy linder TO 3 3.37 E+3cubes 1 3 3.97 E+4 -cubes 10 V 3 2.02 E+4cy linder TO 4 7.26 E+3cy linder TO 5 1.02 E+4
1 - vitreous waste, 1eachant a t 298k2 - d e v it r i f ie d waste, 1eachant a t 298K3 - v itreous waste, 1eachant a t 372k4 - modeled glass, 1eachant a t 298k, one day o f leaching5 - modeled glass, 1eachant a t 372k, one day o f leaching
29
Table VII. Majoc Radioisotopes Released from Ten Year Old Waste Forms . : /v:.;..--:/; ; : :
Waste Leaching Radio Mass Flow RateGeometry Condition isotope (MPC-m3/s)
c y lin d r ic a l 1 90Sr 236.3
137Cs 0.6
125Sb 0.4 ■
: ,54Eu : y , 0.3
cubic 1 90Sr 1417.8
137Cs 3.5
,25Sb 2.2
154Eu1.9
c y lin d r ic a l 2 , " s r 1699.0
154Eu 259.4
106Ru88.3
147Pm 62.1
14:4Ce 23.0
137Cs 7.5
cubic 2 90Sr 10194.0
154Eu1556.7
106Ru530.1
147Pm 372.4
144Ce 137.8
137Cs 45.0
30
Tab1e V I I9 Gbntinued
Waste Leaching Geometry Condition
Radioisotope ;
Mass Flow Rate (MPC-m3/s )
c y lin d r ic a l 3 sv 16416
T37cs 35.5
134Cs ■ 8.6cubic 3 90Sr 9849,8
137c, 212.8
134Cs 51.3
106Ru 4.4c y lin d r ic a l 4 90Sr 6841.8
137Cs 149.6
1S4Eu 9.2
134Cs 3.2
;CfiRu 3.1cy lin d r ic a l 5 90Sr 9621.0
137Cs 210.4
154Eu 12.9
134Cs 4.6
106Ru 4.3
1 - vitreous waste, leachant a t 298K2 - d e v it r i f ie d waste, leachant a t 298K3 - vitreous waste, leachant a t 372K4 - modeled glass, leachant a t 298K, 1 day o f leaching5 - modeled glass, leachant a t 372K, 1 day o f leaching
. ;• ' . ' . ; 31 .
where m ^ mass flow rate o f radioisotopes
Vg = volumetric flow rate o f body o f water (V^-D)
The concentrations o f radionuclides calculated by th is equation
are contingent on the perfect mixing o f released radioisotopes w ith the
volume o f flow o f water. Thus, using equation (18)s and knowing the
mass flow .rate o f radioisotopes from the waste form, the amount o f
water necessary to achieve proper d ilu t io n o f radioisotopes may be de
termined. • 1
Again re ca llin g equation (14), to demonstrate the use o f th is
equation, a sample ca lcu la tion has been done. The mass flow ra te o f
major radioisotopes from ten year old d e v it r i f ie d cubic p a rtic le s is
shown in Table V II. The ca lcu la tion assumes tha t these radioisotopes
are leached in 10"* m /s o f water, and are undiluted. The radioisotopes
are then transported through a typ ica l desert environment. The ion ex
change holdup factors used fo r each o f these.radioisotopes are l is te d 29 1
in Table V I I I . F in a lly , the concentrations o f these radioisotopes
are lis te d against the time which the groundwater trave ls in the envi
ronment in Table IX.
Values in Table IX and a l l other tables in th is work are lis te d
in System In te rn a tio n a l or SI un its . A l is t in g o f a few less common SI
units is found in Appendix I .
32
Table VM I. Ion Exchange Holdup Factors fo r "Typica l" U.S. Desert Soil
Western
Element r1 Element r1t r i t iu m 1 iodine iberyl 1i urn 3x10‘ 3 cesium Ix lO "3
cerium 4x10"4carbon 1x10-1 promethium 4xT0"4sodium . 2x10"2 samarium 4x10” 4ch lorine " i '■: europium 4x10"4argon 1 holmium 4x10"4potassium 6x10” 3 tha llium Ix lO "1calcium 1x10” 2 lead 6x10"5iron 3x10"4 bismuth 2x10"2cobalt 3x10"3 polonium 9x10"3nickel 3x l0 "3 astatine 1selenium , 1x10"2 radon 1 -krypton ■ 1 francium Ix lO "3rubidium 2x10"3 radium 2x10"3strontium 1x10” 2 actinium 2x10” 4y ttr iu m Ix lO "4 thorium 2x10"5zirconium 1x10" 4 protactinium 6X10"5niobium 1x10"4 uranium 7x10"5molyodenum 4x10"2 neptunium Ix lO "2technetium 1 plutonium Ix lO "4rutheni um 3x10” 4palladium 9x1O” 4 americium Ix lO "4cadmium Ix lO "4 curium 3x10"4t in 9x10” 4 berkelium 3x1O"4antinomy Ix lO "2
33
tab le IX. Concentrations o f Radioisotopes Dissolved from D e v itr if ie d Reference Waste as The^ Are Transported through the Environ^ ment
90Sr V54Eu CT06^trat1H 7 p ; C) H 4Ce 137^ I % (Y r )
1.62E8 L56E7 B,30E6 3.72E6 T.3BE6 4,S0E5 09.95E7 5.28EB 4.90E-4 3.87:3 3.T7E-4 3.S7E5 0.019.71E7 1.79E6 ■ 3,55 = ,a 3 2.83E5 0.029.47E7 6.05E5 3i 46E-3 2,25E5 0.039.24E7 2.05E5 e» =» 1.79E5 0,049.02E7 6,94vE4 '="= 1,42E5 - - 0.058.62E7 2,35E4 ao=e 1.13E5 0.068.58E7 7,95E3 -b i- “ — 8.92E4 0.078.37E7 2.S9E3 •m=e 7,09E4 0.088.17E7 . 9,12E2 “ =» 5.S3E4 0.097.97E7 3.09E2 4.46E4 0.106.2317 6.11E-3 “ *° 4.43E3 0.204 .87E7 —— <«>.co 4.39E2 0.30.3.80E7 =oe” 4.36E1 0.402.97E7 . ‘= CD ' 4,33 0.502.32E7 • '«> «* 4,29E-1 0.601.81E7 — ” » => ™ == = *»=> onac 0.701V39E7 -*> —> . • = *=° == = == = bo CO 0.801.09E7 =* = CO CD coco 0,908.66E6 * =° esco ” — 1.007 . 35E5 CO.OD 2,00€»23E4 —— <=>«= 3.005.29E3 ' co’=D op = COCO = a 4.004.49E2 ” = 5.003, St El «=>” =° 6,003,23 ==™ ' c» « 7.002.74E-1 -------- ■=“ ** coco =° «= 8,00
CHAPTER 5
DISCUSSION A # CONCLUSIONS
A study has been performed to develop the importance o f the d is
so lu tion phenomenon to h igh-leve l nuclear waste management. Theory has
been presented describing the mechanisms o€ radioisotope release from
an a rb itra ry waste form.
Typical borosi11 cate glass waste forms made from PUREX-LWR
waste have been considered. The mass flow o f radioisotopes from these
waste forms has been calculated. F in a lly , expressions have been de
veloped showi ng the re la tio n between the. so lu tion behavior o f waste
forms and the environmental transport o f radioisotopes, w ith numerical
examples displayed fo r the typ ica l waste form placed in possible envi
ronments.
Several im portant observations may be made from the analysis.
F irs t , regarding sp e c ific radioisotope release, i t is apparent tha t the
m ajority o f the radioisotopes $ which may be p o te n tia lly released are
composed o f a few sp e c ific f is s io n products, not the actin ides. The
data in Table V II ind ica te tha t fo r ten year old waste forms, ^ S r ,
^ C s , ^ R u , ^ C e , ^ P m , and ^ % u are the radioisotopes o f major
release. These radioisotopes have h a lf- l iv e s such th a t a fte r several
hundred years, they w i l l have nearly decayed. On the other hand, on
34
th is time scale the summation o f a i l actin ides released is a t r i v ia l
fra c tio n o f the to ta l radioisotope concentration released by th is mech
anism to the environment.
From the analysis comparing vitreous and d e v it r i f ie d products,
i t is cohcluded tha t d e v it r i f ic a t io n is a detrim ental phenomenon based
on the higher so lu tion rates o f d e v it r i f ie d glass forms. Calculations
showed tha t the in i t ia l mass flow ra ta o f radioisotopes from a d e v it r i
f ie d waste form is greater than 10® MPC-m^/s, while a vitreous waste
form in the same condition is about 200 times less so luble.
The statement th a t a d e v it r i f ie d product is worse than a v i t
reous form must be made w ith a note o f caution. This analysis is based
on the d isso lu tion o f waste form sf In the case o f the dropping o f the
waste form, i t may be advantageous fo r the product to be d e v it r i f ie d .
Some evidence ex is ts tha t a d e v it r i f ie d product has a higher impact
strength than the vitreous product. However, th is work does not address
th is subject, ra ther i t is noted as a p o s s ib ility .
On the basis o f ca lcu la tion , i t is concluded th a t i t might be
advantageous to have waste forms in small shapes ra ther than in large
ones. Data presented tha t d e v it r i f ic a t io n or a b o ro s ilica te glass
could re su lt in a 200-fo ld increase in the so lu tion ra te over a v itreous
form. From a heat tra n s fe r po int o f view, lower temperatures might be
expected w ith smaller waste forms thus the rate o f d e v it r i f ic a t io n
would be slower. I f the Smaller waste form did not cause a surface
area increase o f greater than 200 times, then from a rad io iso top ic
release po in t o f view, i t would become advantageous to have such forms.
Before such a decision could be made, an assessment o f the overa ll envi
ronmental impact o f such a decision would have to be done to ensure th a t
the overa ll r is k to man would not be increased.
Regarding the release o f radioisotopes in some transporta tion
accident scenario placing the waste form in to some r iv e r type environ
ment, co rre la tions have been developed showing the type o f environment
flow rate necessary to ensure proper d ilu t io n o f radioisotopes as a
function o f radioisotope release rate from waste forms. In terms o f the
reference b o ro s ilio a te waste form, i t is concluded th a t the best waste
form to ship would be a vitreous form, o f a t least ten years o f age.
This statement is made on the basis tha t th is form has the lowest re
lease ra te o f radioisotope o f those considered in the analysis.
Several s im p lify in g assumptions were made in the course o f the
above mentioned ca lcu la tions. Of primary importance, i t was assumed*
fo r the c y lin d r ic a lly shaped waste form, tha t the e n tire surface area
o f the form was exposed to some leaching environment. In a c tu a lity , i t
is l ik e ly tha t in any transporta tion type scenario, th is condition w i l l
not e x is t. Rather more l ik e ly , several decades o f reduction o f the ex
posed surface area wpuld e x is t. This fa c t alone would g rea tly reduce
the quan tities o f water which would be necessary to ensure proper d i
lu tio n o f radioisotopes.
In regard to the release o f radioisotopes in some disposal
scenario, i t was shown th a t re la tiv e ly high flow rates o f ground water
carrying radioisotopes would be s u ff ic ie n t to ensure proper hold-up o f
rad io isotopes;in a desert type environment. Ca1cu1 atidns showed th a t■90fo r the quan tities o f Sr released from a ten year old d e v it r i f ie d
waste form9 i f the ground water took about e ight years to transport
from the nuclear waste disposal s ite to man's environment, then concen- 90tra tionS o f Sr reaching man would be a t permissible, le ve ls . I t was
also shown tha t i f the ^ S r was con tro lled through i t s environmental
transport, then a ll other radioisotopes would be s im ila r ly con tro lled .
As previously discussed, s im p lify ing assumptions were made en
route to these conclusions. The exposed surface area s im p lif ic a tio n
would force calculated values to be lower. In add ition , th is analysis
assumed very T i t t le radioisotope d ilu t io n . In r e a l i ty , i t is a n t ic i
pated tha t as some 1 eachant moved through the environment, i t would be
greatly d ilu ted . Thus, calculated values would again be lowered in
re a lity .
On the basis o f the theory and numerical examples, i t is con
cluded th a t two factors dominate the concentrations o f radioisotopes
which may reach man from s o lid if ie d HLW. The f i r s t is the temperature
o f the waste form. This is o f importance because o f the arrhenius re
la tio n describing the d isso lu tion o f a waste form. As the temperature
fo r d isso lu tion increases, the d isso lu tion ra te is seen to increase a t
an exponential ra te .
The second fa c to r o f importance is the waste form iso la tio n
distance from man and the ion exchange holdup encountered in tha t d is
tance allow ing rad ioactive decay to reduce the isotope concentration.
The in i t ia l concentration o f radionuclides has very l i t t l e e ffe c t on
. : ; : ; : - 3:8
the tsoTation distance required. As a s e n s it iv ity ana lys is9 the volume
flow rate o f ground water used in the calcuTation o f Table IX was in -4
creased by a fa c to r o f 10 . As th is g rea tly reduced the I n i t ia l con
centra tion o f the radioisotopes, i t was believed th a t the iso la tio n
distance from man would also be g rea tly reduced. However, the decrease
was found to be only a fa c to r o f ten. This is reasonabie sihoe the
is o la tio n distance is a function o f the natural logarithm o f the in i t ia l
concentration9 and the natural log Of 10 is about 10.
The proper is o la tio n o f a waste form may then:be described as
both the temperature contro l o f the waste form to ensure lower disso
lu tio n ra tes , and geometric is o la tio n o f the waste form in a ca re funy
seTected environment w ith favorable sorption c h a ra c te r is tic s . Thus» i t
is concluded tha t w ith the temporary storage o f a s o lid if ie d waste fo r
several years to lower heat generation rates and w ith carefu l s ite
se lection fo r the waste repos ito ry , h igh-leve l nuclear waste disposal
may be ca rried out ensuring minimum exposure o f man to radionuclides.
APPENDIX A
LISTING OF MAJOR RADIOISOTOPE COMPOSITION IN PUREX WASTES
Iso tope...■ .... ' ... v . .. • . . . .. ... .............. ; „ ...
' m wt%
FISSION PRODUCTS
V 8.97E-3 1.54E -3
129I : 6»49E-7 ; 6.39E-6
: 131, 3.07E-5 ' 4 .25E-6
89Sr 1.76 1.03E-1
■ : > ; 90s r 1.74 2.10E+1
90y 1.74 5.50E-3
: 91 y 3.63 2.55E-1
■ 95z , 6.36 5.13E-1
95raNb 1.35E-1 6.28E-4
95Nb 1.19E+1 5.21E-1
: 103Ru 2.25 1.216-1
103mRh 2.24 1.196-4
T06Ru 1 ,06E+1 5.35
106Rh 1.066+1 5.106-6
110mAg 5.91Et2 2.126-2
119mSn 6.29E-4 2.456-2
123mSn 4.57E-2 1.00E-2
39
Isotope wt%
125Sb : 1v52E-1 2.45E-1
1'25mTe 4.78E-2 4 .54E-3
127mte 7.61E-2 1.33E-2
Te : 7.49E-2 . 4 .84E-5
Te 1.T3E-1 6 .28E-3
129Ie T.13E-1 1 o08E-5
Cs 4 ,22 . 5.68
136Cs : 7 ,13E-4 1 . 65E-5
■ 137Cs 2.45 4 . 82E+1
137mBa 2.26 7 . 3TE-6
l40Ba 7 . 22E-3 1 . 69E-4
140La 8 . 30E-3 2 . 55E-5
141Ce 1.49 8 . 94E-2
144Ce 1 . 62E+1 8 .70
143Pr 1 . 278-2 3 . 26E-4
144Pr 1 . 62E+1 3 . 67E-4
Nd T.02E-3 2 . 17E-5
147Pm 2 .94 5.56
148Pm 4 . 71E-3 4 . 91E-5
151Sm 7 . 57E-3 . 4 . 91E-1
154Eu 1 . 805-1 2.12
155Eu 7 . 75E-2 9 . 75E-2
,uTb 3 . HE- 3 4 . 76E-4
41
■Isotope C i l : V Wt%
ACTINIDES
235U 4.00E-8 2.47E-1
238U 7, JOE-7 3.06E+1
238Np : 2.43E-3 1.23E-7
238Pu 1.69 1.28
239Pu 1.63E-1 3.52E+1
240Pu 3.06E-1 1.78E+1
241Pu ; ; ' 7.70E+1 J / 9.10
242Pu 1.46E-3 4.94
241 Am- 1.31E-1 - 5.34E-1
242Ara 2.43E-3 3.97E-8
242Cm 1.9TE+1 7.66E-2
243Cm : 1.171-2 3 .35E-3
244Cm 2.25 2.56E-1
APPENDIX B
WEIGHT FRACTIONS OF MAJOR FISSION PRODUCT RADIOISOTOPES
The absolute so lu tion in BorosiTicate Glass ra te o f each iso
tope from b o ro s ilica te g lass, unless otherwise determined, is the
product o f the bulk waste s o lid 1 each rate times the fra c tio n a l mass
which th a t isotope constitu tes o f the waste.
In th is ana lys is j the waste oxides constitu te about 22 wt% o f
the to ta l glass mass. In the waste oxides, the actual waste isotopes
are about 80 wt% o f the to ta l while the oxide is about 20 wt%. This
means tha t the actual waste radioisotope content is 17.6 wt%.
Appendix A presents the weight fra c tio n o f major fis s io n prod
uct isotopes in wastes» The product o f the weight fra c tio n o f each
isotope o f a l l radioisotopes times the 0.176 mentioned above is the
weight fra c tio n o f each isotope in g la ss ifie d wastes. These frac tions
are presented in Table B - l .
42
43
Table B -L Weight Fractions o f Major Fission Product Radioisotopes in B o ros ilica te Glass
Isotope Weight Fraction Isotope Weight Fraction
89Sr 1.8IE-4 129mTe . I.ITEkS
90$r 3.70E-2 129fe 1,901-8
90y 9.68E-6134Cs . 1.001-2
91Y . 4.49E-4 ,36Cs 2.901-8
91mNb lv iiE -6,37Cs 8.501-2 V
95Nb 9.17E-4 137mBa I . 291-8
103Ru 2.13E-4 140Ba 2.971-7
103mRh 2.09E-7 UOla 4.491-8
106Ru9,42E*3
141Ce 1.571-4
106Rh 8.98E-9 144Ce T. 501-2
110mAg 3.73E-5 ' 144pr 6.461-7
119mSn 4.31E-5 147Nd 3.821-8
123mSn 1.7SE-5 147Pm 9.791-3
1253b 4.31E-4 148Pm • 8,641-6
125niTe 7,99E-6 151Sm 8.641-4
127mTe 2 .34E-5. 154Eu 3.731-3
127Te 8.52E-8155Eu
1.722-4
160Tb 8.382-7
APPENDIX C
WEIGHT FRACTION/OF ACTINIDES IN BOROSILICATE GLASS
Table I I shows the ra tio o f actin ides to f is s io n products in
POREX liq u id waste to be 0.190. In Appendix B, i t is Stated tha t the
fis s io n product isotopes constitu te 17.6 wt% o f the b o ro s i l lc a te glass
sol id . . This means actin ides are (0.176) x (0.190) or 3.36 wt% o f the
boros H i cate wastes. I f th is fig u re i s m u ltip lie d by the values o f
Weight each actin ide constitu tes o f a l1 actin ideS: the weight fra c tio n
o f the b o ro s ilica te waste each a c tin ide represents is determined.
These figures are shown in Table C - l.
Table C - l. Weight Fractions o f Actihides in B o ros ilica te Glass
Isotope Weight Fraction Isotope Weight Fraction
235u 8 .30E-5 241 Pu 3.06E-3
238u 1.03E-2 242PU 1.661-3
238Np 4.13E-11 24lAm 1.801-4238pu 4,30Er4' 242Am 1.331-11
239PU 1.181-2 242Cm 2.571-5
240Pu . 5.98E-3 243Cm 1.131-6
244Cm 8.61E-5
44
APPENDIX D
TABULATION OF MAXIMUM PERMISSIBLE OONCENTBATIONS (MPC) OF VARIOUS ISOTOPES IN WATER
Isotope MPC (Bq/m3) Isoltope MPC (Bq/m3)
3H 1.81E8 . 110mAg 1.11£6
85Kr N/A 119mSn ' 1.1TE6
131mXe N/A 123mSn 1.11E6
,29I 2.22E3 V 125Sb 3.70E6131, T.TLE4 125mTe - 7.40E6
89Sr T.1TE5 127mTe 2.22E6
90Sr 1.11E4 127Te 1.1TE790y 7.40E5 129mTe 1..11E591y 1.11E6 129Te : 2.96E7
95Zr 2 .22E6 134Cs 3.70E6
95mi!b 3.70E6 136Cs 3.33E6
95Nb 3.70E6 137Cs 7.70E5
,03Ru 2,96E6 137mBa 7.40E5
' 03mRh 3.70E8 140Ba T.TIE6
1G6Ru ‘ - 3 .70E5 140La 7.40E5
,06Rh 3 .70E5 l41Ce 3.33E6
144Ce 3.70E5 238u 1.48E7
45
46
Isotope MFC (Bq/m3) Isotope : MFC (Bq/m3)
143Pr 1.85E6 ' 238Np 1.11E3
W p r 3.70E5 238Fu 1,8525
147Nd. 2.22E6 239Pu 1.8525
,47Pm 7V40E6 240Pu 1.8525
"* 43Pm ■ 1.85E6 : 241Pu 7 . 4026 :
i6 ism : 1-48E7 . 242Pu 1.8525
,54EU 7.40E5 241Am 1.4825
: 155eu 7.70E6 242Am . 1.4825
160ib 1.48E6, 242Cm 7.4025235.J 1.1127 243Cm 1.8525
' 244Cm 2.59E5
APPESlX E
M0IO1SOWIG.RELEASES FROM WASTE FORMS VITREOUS: , FORM» LEACHANT AT 2!98K :
Mass Flow Rate fromi . f6\/1 4 %n v*
Waste Formpit (MR0-m3/s )
Isotopev jiin u e r1 Year
Wjf 1 lilUGI10 Years
UUucd1 Year
Cubes■ 10 Years
89Sr 28.9 171.3
90Sr : 294.7 236.3 T 1767.9 1417,8
90y 294.7 216,3. 1767,9 1417.8
91Y 61.9 37T.4
95Zr 5.4 32.2
98mNb 0.1 0.4
95Nb 6.1 36.6
103Ru1.4 8.6
1(33mRh 0.01 0.1
1Q6Ru53,9 0.1 323,2 0.6
106Rh 51.9 0.1 321,2 . 0,6
118% 0.1 ” *= 0,6
1l9mSn 0.1 0.6
123mSn 0.1 0.5
125Sb 3.7 0.4 22,3 2.2
48
linder '■ Cylinder CubesIsotope 1 Year 10 Years 1 Year
Te
Te
Te
Ba
0.1 - 0.4
127je 0.01 — 0.1
0.02 - - 0.1
0.01 ** 0.1
134q 2.6 : 0.1 : 16.5
136cs —- ” ™
T3?Gs 0.7 ; ' 0.6 4.3
0.7 0.6 4.3
140ga 0.01 — 0.1
!40ta 0.02 — 0.1
141c 0.8 — 5.1
144Ce 80.7 0.03 484.5
143pr 0.01 — 0.1
144pr 80.7 0.03 484.5
MdWpm 0.7 0.07 4.5
0.5 * - 2.9Pm
151Sm .
154Eu 0.5 0.3 2.7
1B5Eu 0.02 0.1
160jb — - - 0.02
Cubes 10 Years
0.8
3.5
3.5
0.2
0.2
0.4
1.9
49
Cylinder Cylinder Cubes CubesIsotope 1 Year 10 Yeans I Year 10 Years
238^p ™- .0.02 “ “
Z3Spu 0.01 0.01 0.07 0.07
239pu 1.2E-3 1.2E-3 7.0E-3 7.0E-3
24QpiJ . ; >' 2 .2E -3- 2.2E-3 0.01 0.01
241.pu . 0.01 8.6E-3 0 .0 8 . 0.05
242pu 1.QE-5 1.0E-5 6.3E-5 6.3E-3
241Am 1.2E-3 1 .2E-3 7.0E-3 6.9E-3
242Am 2.2E-5 — 1.3E-4
242q^ 0.03 0 .2
2 4 3 ^ 8.3E-5 6.8E-5 5.0E-4 4.1E-4
244r 8.2E-3 5.8E-3 0.05 0.03
APPENDIX F
RADIOISOTOPIC RELEASES FROM WASTE FORMS OEVITRIFTED FORM, LEACHANT AT 298K .
Mass Flow Rate from Waste Form (MPC-m /s ) Cylinder Cylinder Cubes Cubes
Isotope 1 Year TO Years 1 Year 10 Years
89Sr 207.2 — 1243.3
90Sr 2118.1 / ; 1699.0 12710 . 10194
90Y 2118.1 1699.0 12710 10194
91Y 5183.2 — 31099
95Zr 4500.2 27000
95mNb 58.0 — 347.8
95Nb 5120.1 — 30720
108Ru 1193.9 — 7163.5 r - ' ■
103mRh 1193.9 — 7163.5
106Ru 45230 88.3 271400 530.1
106Eh 45230 88.3 271400 530.1
110mAg 83.8 0.01 502.8 0,05
119mSn 89 0 0.01 533.8 0.06
123rnSn 64.9 — 389.2 - -
125Sb 64.3 6.4 385.7 38.2
125mTe 10.2 — 61.0 —
127raTe 54.1 — 324.8
50
51
Cylinder Cylinder Cubes CubesIsotope 1 Year 10 Years 1 Year 10 Years
U /Te 10.7 64.1
127mTe 160.7 -oo» 964.3
129le 7.3 44.1
134Cs 35.3 1.8 211.8 10.9
136Cs
137Cs : 9.2 ; 7.5 55.4 45.0
137mBa ■ ; 9.2 7.5 55.4 45.0
140Ba 10.2 61.3
143U 6.3 37.5
l41Ce 740.5 4442.8
144Ce 67730 23.0 406400 137.8
143Pr 10.8 65.1
144Pr 67730 23.0 606400 137.8
147Nd 0.7 . 4,3
147Pm 625.7 62.1 3754.0 , 382,4
148Pm 398.4 2390,1
151Sm 0.8 0.8 4.8 4.5
154Eu 383.4 259.4 2300.6 1556.7
155Eu 16.6 0,4 99.5 2,5
160Tb 3.3 20.0
235U •
238u
52
IsotopeCylinder
1 YearCylinder 10 Years
Cufees . 1 Year
Cubes 10 Years
238Np 9.01*4 5.4E-3
238PU 3 .7E-3 3.5E*3 0.02 0.02
239Pu 3.6E*4 3.6E-4 . 2.2E-3 2.2E-3
2M?u 6.8E*4 6.8E-4 4 .1|*3 . 4.1E*3
245 ?u 4.2E-3 2.6E-3 . 0.03 0,02 .;
242PU 3.2E*6 3.2E-6 1.91^5 1.9E--5
2MAm 3.6E*4 : v 3.6E*4 : ' 2.2E-3 /y .;-'-2.2 i»3 ;-:
242a« 6.7E-6 , 4.31*5
242Cm 0.01 0.06
243Cm 2.6£*5 ■ 2.1E-5 1 .5E*4 1.3E*4
244Cm 2.5E-3 1.8E-3 0.02 0.01
APPENDIX 6
RADIOISOTOPE RELEASE FROM WASTE FORMS VITREOUS . FORM, LEACHANT AT 372K
Mass Flow Rate from Waste Form (MPC-m^/s) lin d e r Cylinder Cubes .Cubes
Ru
Ag
Isotope : 1 Year TO Years 1 Year 10 Years
S9Sr 199.8 — 1198.5 -
90Sr 2049.2 ; 1641.6 12295 9849.8
90y 2049.2 1641.6 12295 9849.8
91y 43.2 - . 259.0
95Zr 37.5 — 224.9 -
OSm 0.5 2=9
9 5 ^ 42.5 254.9
T03ru 10.0 59.9 “ “
103m^ 0.08 — 0.5
374.8 0.7 2248.9 4.4
106Rh 374.8 0.7 2248.9 4.4
0.7 — 4 .2
n9mSn 0.7 - 4.5
123nign 0.5 3.2
Sb 0.5 0.05 3.2 0.3
53
54
Cylinder Cylinder Cubes CubesIsotope 1 Year TO Years 1 Year TO Years
e
Te
te
0» 8 =- 5.1
127niT6 0,4 — ; 2.7
0.09 0.5
1.3 8.0
12 9^ 0.06 — 0.4. -
134Cs 166.5 8.6 998.8 ■ 51.3
136cs “ “ Q.02 —-
137€s 43.7 35.5 262.1 212.8
137% 43.7 35.5 262.1 212.8
140gg 0.08 _ ■ 0.5 - -
140^a 0,1 0.9 , - -
14Tce 5.9 35.1 ■“ t
T44Cg 561.9 0 .2 3371.7 : 1.1
143pr 0.09 — 0.5
144pr 561.9 0.2 . 3371.7 1.1
347Nd 0,04 - -
147ps 5.2 0.5 31.3 3.1
T48pm 3.3 - 19.9
T5Tsm ” = 0.04 0,04
154Eu 0.3 . 0.2 1.9 1.3
155Eu 0.1 — 0.8 0.02
160 j b 0.3 — 1.7
Cylinder Cylinder Cubes CubesIsotope 1 Year 10 Years 1 Year 10 Years
£ ODy
238u c4““
238np 0.04 0.3
238Pu 0.2 0.2 1.1 1.0
239Pu 0.02 0.02 0,1 0.1
240Pu ■ 0.03 : 0.03 ; 0.2 0.2
241Pu 0.2 0.1 1.2 0.8
242?u 1.5E-4 1.5E-4 ' • 9..21-4 9.2
24' ari0.02 0.02 0.1 0.1
CM 3.2E-4 1.9E-3 •=”
242Cm 0.5 3,0
243Cm 1.2E-3 1.0E-3 7.31=3 6.0
244Cm 0.1 0.08 0.7 0.5
APPENDIX H
RADIOlSOfGPlC RELEASES FROM WASTE FORMS, MODELED FORM AT ONE DAY OF LEACHING, WASTE AGE — TO YEARS
Mass Flow Rate from Waste From (MPC-m^/s)Isotope Cyfinder - - HgO a t 298K Cylinder — HgO at
90Sr ‘ 6841.8 9621.0
90Y. . 1 0 5 . 4 148.1
1Q6Ru / ; ' 8 . ( 1 9 : .4.35 :
106Rh 0.0117 0.0165
125Sb 0.266 0.318
Cs 3.25 4.6
Cs 149.6 210.4
137mBa 137.9 194.0
147Pm 2.66 3.74
l51Sm 0.0265 0.0372
154Eu 9.2 12.9
155Eu 0.0172 0.0242
235U 2.95E-10 4.15E-10
238U 4.28E-9 6.03E-9
■Pu 0.676 0.949
239Pu 0.0723 0.102
240Pu 0.131 0.185
56
57
Isotope Cylinder --HgO a t 298K Cylinder - - HgO a t 372K
241Pu 0.539 0,756
Pu 6,46E-4 9 .1 0 W
Am 0.0726 0.102
242Am 1.61E-8 2.34E-8
243Cm . 4.28E-3 6.03E-3
244Cm 0.351 0.494
APPENDIX I
; ' : NUMERICAL UNITS
Calculations performed in th is work are carried out in the Sys
tems Internationa 1 or SI un its . This section 11ists some o f these un its
w ith the conventi onal engineering counterparts. In a dd ition , a m u lti-
p ly ing fa c to r to convert the SI u n it to the commntional u n it is in -
eluded.
Conventional M u ltip ly ingSI Unit Unit Factor
Mg/(m2s) g/(cm2day;) 8 .640E-6
W/mK B tu / f th r0F 1.872
kg/m3 g/cm3 1.0E-3
J/kgK B tu /lh m0F 7.755E-4
TJ/kgU MWD/MTU 2.778E5
EBq/m3 Ci/1 2.703E4 '
PBq/ffl3 Ci/1 2.703E1
53
LIST OF REFERENCES
1. McNelly, M. J . , J. A. Liebermann, L. S. Bohl, and A. B. Carson3"Nuclear Fuel Cycle and Energy Parks," Trans. Amer. Nucl. See.,Vol. 22, Nov. 1975, P. 303..
2. Randl, R ., and M. Hagen, "The Concept o f Closing the Nuclear Fuel Cycle in the Federal Republic o f Germany," Trans. Amer. Nucl. Soc. , Vol. 22, Nov. 1975, p,302.
3. United States Nuclear Regulatory Commission, "Rules andRegulations: Licensing o f Production and U til iz a t io n F a c ilit ie s , " 10 CFR 50, Appendix F, June 20, 1975.
4. Wacks, M. E ., Univ. o f A r iz . , and W illiam H ew itt, USNRC, Personal Correspondence, July 30, 1976:
5. Mendel, J, E ., "A Review o f Leaching Test Methods and the Leachi b i l i t y o f Various Solid Media Containing Radioactive Wastes," 6NWL-1965, Ju ly 1973. .
6. E l l io t , N. M., and D. B. Auty,. "The D u ra b ility o f FINGAL Glass,Part I . Discussion o f Method and E ffec t o f Leaching Cond it io n s ," AERE-R-51 51, March, 1967.
7. Eliassen, R ., and M. I . Goldman, "F ixa tion in Vitreous o f High A c t iv ity Fission Product Wastes," pp.576-589, TID-7613, 1961.
8. Paige, B. E., "L e a c h ib ility o f Glass Prepared from Highly Radio active Calcine Alumina Waste," IDO-14672, Ju ly , 1966.
9. Alleman, R. I . , "L e a c h ib ility Tests o f Spray Calcined M elts," Quarterly Progress Report, Research and Development A c t iv i t ie s , F ixation o f Radioactive Wastes, Oct.-Dec., 1964, A. M. P la tt , ed. , m -846M , pp ll6 -18 , Jan 1965.
10. Helmerl, W., H. Heine, L. Kahl, H. W. Levi, W. la tz e , G, Malow, E. Schiener, and P. Schubert, Research on Glasses fo r Fission Product Fi xa ti on Report, Jan,, 1968— June , 1971, Hahn-Mei the r In s ti tu te HMI-B109, S ep t., 1971.
11. Mendel, J. £ . , and J. L. McElroy, "Waste S o lid if ic a t io n Program Vol. 10, Evaluation o f S o lid if ic a t io n Waste Products," BNWL-1666, Ju ly , 1972
59
60
12. Post* R. G., Univ. o f A riz . s Personal Communication,, Aug. 9*1976. : ' / v
13:, Wetsik, J, H ., J r . , B a tte lle Northwest Laboratories, PersonalCommunication, August, 1976.
14. Friedlender, G., J. W. Kennedy and J. M, M in e r , Nuclear and Radiochemistry, 2nd e d ., John Wiley and Sons, In c . , New York, 1955, p .407.
15. H e lffe rich , F ., Ion Exchange, McGraw-Hill Book Company, In c . , New York, 1962, p .6.
16. T e lle r , E ., W. K. Ta lley, G. H. Higgins, and G. W. Johnson, the Constructive Uses o f Nuclear Explosives, McGraw-Hil1 Book Co., New York, 1968, p .118.
17. A llie d Gulf Nuclear Services, "Barnwell Nuclear Fuel Plant: Safety Analysis ReportV" USAEC Docket 50:332, January, 1974.
18. Batte l 1 e Northwest Laboratories, "High-Level Radioactive Waste Management A lte rn a tive s ," BNWt-1900, Vol.. 1, May, 1974.
.19, Angelo, J. A ,, R. G. Post, F. E. Has k in , and C, E. Lewis, "AStudy o f Long-Term Heat Generation in Nuclear By-Products from LWR and LMFBR Systems," IAEA-SM-170/58, 1973. -
20. Kelley, J. A ., "Evaluation o f Glass as a M atrix fo r S o lid if ic a t io n o f Savannah River P lant Wastes," DP-1397, O ct., 1975.
21. Kelley, J. A ., "Evaluation o f Glass as a M atrix fo r S o lid if ic a t io n o f Savannah River Plant Wastes," OP-1382, May, 1975,
22. Mendel, J. E ., W. A, Ross, F. P. Roberts, R. P. Turcotte , Y. B. Katayama, and J. H. Westsik, J r . , "Thermal and Radiation Effects On B o ros ilica te Waste GlassesIAEA-SM-207/100 or . BNWL-SA-5534, March, 1976,
23. National Academy o f Sciences, " In terim Storage o f S o lid if ie d High-Level Radioactive Wastes," 1975.
24. McElroy, J. L . , A. S. B lasew itz, and K. J. Schneider, "Status o f the-Waste 'S o lid if ic a tio n Demonstration Program," Nuclear Technology, Vol. 12, p .69, S ep t., 1971,
61
25. Angelo, J. A ., "Heat Transfer from Radioactive Wastes in Deep Rock," D issertation in preparation, 1974.
26. Pittman, F ., "High-Level Radioactive Waste Management A lternat iv e s ," WASH-1297, May, 1974.
27. Van Vlack, L. H ., Physical Ceramics fo r Engineers, Addison- Wesley Publishing Company, Reading, Massachusetts, 1964, p .66.
28. Ross, W. A ., "Impact Testing o f Vitreous Simulated High-Level Wastes in Canisters," BNWL-1903, May, 1975.
29. Burkholder, H. C ., M. 0. Cloninger, D. A. Baker, and G. Jansen, "Incentives fo r P a rtitio n in g High-Level Waste," BNWL-1927,Nov., 1975.
30. United States Nuclear Regulatory Commission, "Rules and Regula tio n s : Standards fo r Protection Against Radiation," 10CFR20,30 A p r il, 1975.