THE SOVIET JOURNAL OF ATOMIC ENERGY VOL. 9 NO. 4

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• Publication Date: September 1, 1961
• Content Type: REPORT

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r to
THE SOVIET JOURNAL OF
TRANSLATED FROM RUSSIAN
CONSULTANTS BUREAU
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TECHNIQUES IN
FLAME_
PHOTOMETRIC
ANALYSIS
by N. S. Poluektov.
TRANSLATED FROM RUSSIAN
Original published by the
State Scientific-Technical Press
for Chemical Literature, Moscow
This volume contains a practical and comprehensive survey of the
techniques employed and the instruments required for this important
rapid analysis method, as well as a brief account of the theoretical prin-
ciples involved. Experimental procedures and the design of apparatus
are discussed at length, making the relevant chapters a valuable manual
for all chemists concerned with analytical problems:
The second,half of the book deals with the determination of individual
elements in a variety of materials, :providing much factual information,
together with a complete and international bibliography. There are
many useful tables, and two appendices which consider wavelengths of
brightest spectral lines" and peaks of molecular bands of elements
excited in flame, and recommended wavelengths of spectral lines and
peaks of molecular bands for determination of elements with the aid of
a spectrophotometer employing glass optics and using an air-acetylene
flame. I
Scientists working in the varied fields in which flame photometry is
today, applicable - in geochemistry and mineralogy for the analysis of _
bolic bases in soils and the -analysis of fertilizers and plant materials, in
biochemistry and medicine for the investigation of urine, blood serum;
and tissues, in industry for production control by analysis of metal
alloys, glasses, cements, refractories, reagents, etc. - will find
TECHNIQUES IN FLAME PHOTOMETRIC ANALYSIS
an invaluable handbook.
waters and minerals, in agrochemistry for the determination of.meta-
230 pages- $9.50
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EDITORIAL BOARD OF
ATOMNAYA ENERGIYA
A. I. Alikhanov
A. A. Bochvar
N. A. Dollezhal
D. V. Efremov
V. S. Emel'yanov
V. S. Fursov
V. F. Kalinin
A. K. Krasin
A. V. Lebedinskii
A. I. Leipunskii
I. I. Novikov
(Editor-in-Chief)
B. V. Semenov
V. I. Veksler
A. P. Vinogradov
N. A. Vlasov
(Assistant Editor)
A. P. Zefirov
THE SOVIET JOURNAL OF
ATOMIC ENERGY
A translation of ATOMNAYA ENERGIYA;
a publication of the Academy of Sciences of the USSR
(Russian original dated October, 1960)
Vol. 9, No. 4
September, 1961
CONTENTS
PAGE
RUSS.
PAGE
Neutron -Flux Distribution in a Homogeneous Boiling-Water Reactor. B. Z. Tor 1 i n ......
787
257
Effectiveness of a System of Rod Absorbers in a Reactor Fitted with a Reflector. V. I. N o s o v.
795
262
The Metallurgy of Uranium. N. P. G a l k i n ........... ......................
804
270
The Solubility Product of the Hydroxide of Tetravalent Uranium. M. A . S t e p a n o v
and N. P. Galkin ............................................
817
282
The Sorption Extraction of Uranium from Pulps and Solutions. B. N. Las korin ........
822
286
LETTERS TO THE EDITOR
Energy Dependence of the Differential Cross Sections and Mechanism of the (d, p) Reaction.
V. B. Belyaev, B. N. Zakhar'ev, and V. G. Neubachin...........
833
298
High Energy y-Ray Beams. V. S. Barashenkov and Hsien Ting-ch'ang ........
835
300
Effects of the Leakage Fields of a Sectional Magnet on the Double Focusing of a Beam.
Yu. A. Kholmovskii ........................................
838
301
Beam Loss at the Limiting Radius in a Phasotron. V . P. D m i t r i e v s k i i, B. I.
Zamolodchikov, and V. V. Kol'ga ................ ..........
841
303
Comparison of Cascade Circuits for Producing Large Currents with Little Ripple.
A. A. Vorob'ev and S. F. Pokrovskii ..........................
845
305
Hydraulic Resistance to the Flow of a Liquid along a Bundle of Rods. V. I. S u b b o t i n ,
P. A. Ushakov, and B. N. Gabrianovich .......................
848
308
Investigation of Heat Exchange in Connection with a Turbulent Flow of Mercury in an
Annular Duct. V. I. Subbotin, P. A. Ushakov, and I. P. Sviridenko ..
851
310
On the Separation of Boron Isotopes by Chemical Interchange. B. P. K i s e 1 e v .........
854
312
Demarcation of Petroleum-Bearing and Water-Bearing Strata with the Application
of Electron and Photon Beams. V. I. Gomonai, I. Yu. Krivskii, N; V.
Ryzhkina, V. A. Shkoda-Ul'yanov, and A. M. Parlag . . . . . . . . . . . .
855
313
Investigation of Attenuation Functions in Water for Neutrons from Isotropic and One-
Directional Fission Sources. V. A. Dulin, Yu. A. Kazanskii, V. P.
Mashkovich, E. A. Panov, and S. G. Tsypin ....................
858
315
Energy Distribution in Water of Fast Fission Neutrons. V. A. D u l i n , V. P. Mash -
kovich, E. A. Panov, and S. G. Tsypin ........................
861
318
Calculation of the Dose Created in an Irradiated Object Moving in the Radiation Field
of a Line Source. U. Ya. Margulis, S. M. Stepanov , and
V. G. Krushchev ......... ...................................
864
320
Simple Calorimetric Method of Measuring the Absolute Energy Dose Received from Powerful
Sources of Ionizing Radiation. M. B. Fiveiskii, Yu. S. Lazurkin,
M. A. Mokul'skii ..........................................
865
321
Annual subscription $ 75.00
Single issue 20.00
Single article 12.50
? 1961 Consultants Bureau Enterprises, Inc., 227 West 17th St., New York 11, N. Y.
Note: The sale of photostatic copies of any portion of this copyright translation is expressly
prohibited by the copyright owners.
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CONTENTS (continued)
RUSS..
PAGE PAGE
The Dosage of Outdoor y Radiation from Radioactive Fallout during 1959. V. P.
Shvedov, G, V. Yakovleva, and M. I. Zhilkina .............
868
323
The Increase in Radioactive Fallout in Gradets Kralove (Czechoslovakia) as the Consequence
of Nuclear Tests in Sakhar. V. S a n t go l z e r . . . . . . . . . . . ........ ........
870
324
[Nuclear Power Engineering in France. Source: Nucl. Power 5, No. 48 (1960) .... .......
327]
The PM-2A Nuclear Package Power Facility .... . . . . . . ,
873
329.
[PRTR Reactor with a Repetitive Plutonium Cycle ..............................
331]
[A Nuclear Training Facility for Greenwich College ..... . .... . . . . . . . . . . . . . . . . .
332]
Beryllium (Present Status of Beryllium Technology and Research). N. M i r o n o v and
S. Kostogarov ..............................................
875
334
[The Uranium Mining Industry in the USA ..................................
337]
The Second Azerbaidjan Republic-Wide Conference on the Uses of Radioactive Isotopes
and Nuclear Radiations. A. M. M a m e d o v ............................
878
338
[Electronic Measuring Equipment and Computers at the Annual United Kingdom Exposition
Source: Nucl. Power 5, 104 (1960) ....................................
340]
[The Berylometer-A New Research Tool. Source: Econ. Geol. 54, 1103 (1959). .........
340]
Decontaminating Enclosure. G. N. Lokhanin and V. I. Sinitsyn ........ . ....
880
341
New Leaktight Glove Boxes for Handling Alpha- and Beta-Emitting Materials.
G. N. Lokhanin and V. I. Sinitsyn.............................
883
344
Brief, Communications ................
886
347
BIBLIOGRAPHY
New Literature ....................................................
888
349
The Table of Contents lists all material that appears in Atomnaya Energiya. Those items that
originated in the English language are not included in the translation and are shown enclosed in brack-
ets. Whenever possible, the English-language source containing the omitted reports will be given .
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NEUTRON-FLUX DISTRIBUTION IN A HOMOGENEOUS
B. Z. Torlin
Translated from Atomnaya Energiya, Vol. 9, No. 4, pp. 257-261, October, 1960
Original article submitted January 25, 1960
A methodof calculating a homogeneous boiling-water reactor in which the density of
the medium varies with the height is calculated in the diffusion and age approximation. For
a cylindrical reactor, the solution can be reduced to elliptical integrals. It is shown that, for
a reactor whose radius is much greater than its height, the solution is expressed in terms of
elementary functions.
In the calculation of reactors in which the heat is removed by the boiling of the moderator we have to
deal with the moderator nuclei and the diffusion of neutrons in a medium of variable density. If the vapor
bubbles of the moderator are small in comparison with the mean free path of a neutron in the medium, we can
regard such a medium as being continuous. If, in addition, the medium is homogeneous, i.e., the atoms of the
fissionable material are uniformly distributed among the atoms of the moderator, then at each point Y of den-
sity y (r) the neutron scattering length is (r), the capture length la(t ), and the total mean free path l (r) are
related to the corresponding quantities 1so, lao, and to in a medium of density yo by
lso Sao to v (' )
where a (r) is a continuous function. Hereafter, it shall always be assumed that
I grad d (r) I ? 1.
Under these assumptions, it is not difficult to obtain (following the method adopted in [1, 2]) in the diffusion
approximation the equations of moderation and diffusion of neutrons in such a medium,`
V(PVN)Lfl~_ nLP)
u
49 n
Co \ l
V (NV n) 8
and the initial conditions n (r , 0) = kN W. Here N is the thermal neutron density; n is the density of the
neutron moderator; Lo is the diffusion length .of thermal neutrons in a medium of density y o; ro is the age
' It can be assumed here that the moderator is stationary, since during the lifetime of a neutron (approximately
10-g sec) in the active zone there is no appreciable displacement of the medium.
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of a moderated neutron in a medium of density yo; TOT is the age of the thermal neutrons in a medium of
density yo; k is the multiplication factor.
We shall consider the case in which the extrapolated boundaries on which N and n vanish coincide.
Under the assumptions made above, the solution of the moderation equation can be represented in the
-
form
The system of equations (2) then reduces to
where a2 is determined from the condition
n = kNe-a2.ro.
PV(PVN)-i-a2N0,
9. + I oat = Ice TOT.
If 8 does not depend on the radius and the angle [8 (r) = 8 (z)] , then for a cylindrical reactor the solution of
Eq. (3) can be found in the form
In place of Eq. (3), we then obtain
N (r) = Nr (r) NZ (A
aZ a Z-f-(a2-cc?P2)NZ0
Nr = Jo (arr),
where J0 is the zero-order Bessel function of the first kind; ar = 2.4/Reff (Reff is the extrapolated radius of
the active zone).
For the further investigation of Eq. (5) it is necessary to determine the form of the function or, at least,
the relation between 8 and Nz. This relation is not difficult to establish if we make the following assumptions:
1) The moderator density is constant over the reactor radius and varies only with the height of the layer;'
2) all the heat released below any cross section is expended on the production of steam, which, in rising
to the top, passes through this cross section;
This assumption is obviously valid for a homogeneous boiling-water reactor, where the -intense, turbulent
agitation of the medium inevitably tends to equalize the vapor content over the radius. The. effect of a small
change in the vapor content in the radial direction is probably slight. In fact, if it is assumed that in the upper
half of the active zone of the reactor, where the vapor content is the highest, the medium density close to the
axis is one-half that at the walls of the vessel, then, according to perturbation theory, one can estimate the
A a2
change in the Laplacian of the system. For the example cited (see figure) these estimates are 2 2- 3%.
a
Of course, in a strongly turbulent flow such a large difference in densities cannot actually occur. From the
estimates it follows that the difference in densities of the medium at the center and at the periphery does not
exceed 1001o.
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3) the quantity of steam passing through the cross section, as in the case of bubbling, is proportional to
F/(1 - F) (F is the fraction of the cross section occupied by steam in the layer) _[3];
4) the steam density y" is small in comparison with the water density y', and hence
(6)
Under the assumptions above,
Nzdz_a F -a
where 8 o = ?O ; a is a constant. Introducing the change of variables'
dz _
dy
and, after differentiating expression (7), we obtain
a do .
dy
For a boiling water reactor in which a?P2 ? a2 over the entire interval of variation of.S , Eq. (5) will
have the form
d dN`
P iz (_t dz ) + ao2N = O.
Inserting relation (8) into Eq. (10), we obtain
d2N 2
dye _+ aoN = 0.
The solution of Eq. (11) can be represented in the form
N = A sin aoy.
Using the condition at the upper boundary of the reactor, we. find
(12)
Integrating expression (9) under the condition 8 (0) = 8 Q, we obtain
3-~o=C(1-Cos aoy)?
' The advantage of this change of variables in our problem was kindly pointed out to the author by Ya. V.
Shevelev.
(14)
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For y = yH ,
C= PH-0o
2
After integration of expression (8),
Z _ 1H+1o sH-Po sin a
2 y - 2a0 oy?
Inserting into expression (15) the value of yH from formula (13), we obtain the expression for the height H of
the active zone:
;t
H=0
2 ao .
From expression (12) it follows that the maximum neutron density occurs for Ymax = yH/2, or for
Zmax= RH-f-Ro n - RH-00 = H
2 2a0 2ao 2 \
2 RH-Po
t NH+QO
i.e., the maximum of the neutron density is displaced downward with respect to the median plane of the reactor;
the greater the difference between the moderator densities in the lower and upper cross sections of the reactor
and the greater the over-all height of the reactor, the greater is the displacement.
The mean density of the moderator is determined by the expression
1 c dz = Yo
Y= H ) H yH
0
Substituting into expression (18) the value of yH/H from expression (16), we find
2Yo
1H?13o
Inserting expression (1), and then (6), into (19), we obtain
- , 1__FH
vV 1-FH/2'
(20)
where FH = F (H).
Since the amount of steam flowing through any cross section of the reactor is proportional to F/(1- F),
the reactor power W is proportional to FH/(1 - FH). If we express FH in terms of 'y and yo by means of
expression (20), we obtain
(21)
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The mean steam content F is expressed in terms of the mean density of the moderator y by means of the
relation
v=y'(1-F).
Inserting (22) into (21), we obtain
Inserting into Eq. (16) the expression
where
W = 2a F
1-F
K= n
a
Expressions (23) - (25) relate the size and power of the reactor to its mean steam content. Here, the
height of the boiling-water reactor proves to be exactly the same as the height of a reactor with a constant
density equal to the mean density of the moderator in the given solution. For a reactor of finite radius, as we
shall show, similar results are obtained. Moreover, it
TABLE 1 follows from relation (23) that the power of the boiling-
Comparative Characteristics of Boiling-Water water reactor is double that of a reactor whose steam
Reactors with Constant and Variable Steam Content content is constant over its height for the same mean
over the Height for the Same Mean Density of the steam content. A check of this circumstance for a
Medium cylindrical reactor of finite radius shows that the ratio
a2.104,
YH,
Y.
H0,
F10-F)
CM-2
g/cn?
g/cm3
cm
2 FH/(i-FH)
19,1
0,846
0,933
131
131
1,04
19,.1
0,733
0,850
168
170
1,09
25,8
0,846
0,935
96,8
96,6
1,02
25,8
0,733
0,857
114
112
1;05
25,8
0,550
0,688
208
212
1,14
29,4
0,846
0,935
86,7
86,5
1,02
29,4
0,550
0,703
158
158
1,10
31,4
0,846
0,935
82,3
82,0
1,02
31,4
31,4
0,550
0,525
0,703
0,674
143
160
140
165
1,101,14
I Note: In the calculations we took as = 8.10'4
m, y o =1.10 g/cm3, and y'.= 1.05 g/ cm3?
2 " " is somewhat above unity and increases.
FHl(1- FH) ~'
with the mean.steam content (Table 1). From this
table it also follows that even for F c 0.35 the error
resulting from formula (23) in the determination of
the power W of the reactor does not exceed 15%.
We shall now consider the case of a cylindrical
boiling-water reactor of finite radius, which, although
more complicated, is of greater practical importance.
Inserting expression (8) into (5), we obtain in the new
variables
d -I- (a2 - ar02) CIO = 0
v
(24)
(26)
with the initial conditions 8 (0) = 8 o and dB = 0
Y
at the upper and lower extrapolated boundaries of the reactor. After a single integration of Eq. (26), we find
P ~R{--ff~ RR
P do - 2 Cl-a2YT 3 arF'3~
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where
Integrating Eq. (27), we obtain
d
p-dy
CIP + C2 a2P2 F 6
Z- Pdp
P
4; (28)
The determination of the constants of integration Cl and C2 can be fundamentally simplified if we assume
that the boundaries of the boiling-water zone coincide with the extrapolated boundaries of the active zone; This
assumption cannot introduce a serious error in the calculations, since the effective additions are small in com-
parison with the height of the active zone and their contribution to the heat balance of the reactor is negligible.
With this assumption we find that
P = OP, - N) (PH "- K) (N - PO) (N - M+
where SH = S (H); H is the effective height of the active zone;
(30)
2 R a R Qli' R / p
02H 6 a~ - 4 ~~o + NH~ - F'oNH -~o F i N2 - - I 6 ar 4 ~No ~ NPH ~02~H
In order not to obtain a negative number under the radical sign of expression (30) for a change of S from So to
SH, it is necessary that a, , SH. This determines the limiting value of SH:
/2a2_2 11
SH lim= Po 0oa2 9 3 /
According to expression '(9),
NZ = a p.
The dependence of 1 on a is expressed by the formula
_ p 2 V6u
Y ~(R1-130(RHa,
where u = F (gyp, k) is the elliptical integral of the first kind [4], Here
~_ 1/(RH--I2) (R-R(~o) . k= 1(PH-Po) (01-R~{2)
sin (RH--Po) (li-t32) ' ((131-00) (11H-r2)
For$ =SH
21/6K
YH - Y (~1-~o)(~H-~r)ar
792
(31)
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where K is the complete elliptical integral of the first kind. The expression for z has the form
1(I3o-)3x) 0(a)H(a) rO'(a) -1-k2 In 0(u-a)
1 j~ 1-ka 0i (a) H1 (a) L ?(a) 2u h9 (u+a). +
sin af3,-Po
Y Pi-NS
0, H, 01, and Hl are Jacobian theta functions [4]. In this case
H- r 8o-8s O'(a)H(a) (36)
-yFt t v1-ki01(a)I1 (a)+~o
If BH - 8H lim ' then _y tends to infinity. Hence BH lim is the maximum value of 8H at which the
reactor remains critical only for an infinite height of the active zone.
Neutron-Flux Distribution Over the Height of a
Boiling-Water Reactor
z, cm I y , g/ cm, I p, cm-1 I Nz/Nmax
0
25
34
43
51
60
70
81
93
109
129
162
208
I1,crn
200
N
1,05
0,94
0,88
0,83
0,78
0,74
0,70
0,67
0,64
0,61
0,58
0,56
0,55
0,0000
0,0007
0,0116
0,0126
0,0129
0,0127
0,0121
0,0112
0,0099
0,0082
0,0061
0,0033
0,0000
0
0,75
0,89
0,97
0,99
0,98
0,93
0,86
0,76
0,63
0,47
6,25
0
Note: The calculations were performed for values
of the parameters given in the caption to the figure.
The minimum value of BH is 8a. The minimum
height of the reactor's active zone which corresponds
to this. value of 8 H is
PH
Hlim = lim 0 8
PH-a0 P
o Ro
Using the mean-value theorem, we obtain
0 6,5 1,0
N/Nmax, y, g/cm3
The distribution of the thermal
neutron density N and the moderator
density y over the height H for the
case a 2 = 25.8.10'4 CM- z, a r=
= 8.10-4 cm-2, YO=1.10 g/cros,
y' = 1.05 g/cm3, yH = 0.550 g/cm 3,
0.688 g/cros.
RH
Hlrm= lim [ d
a,?V(RI-R*)~(R*-132) so Y(RH-R)(R-Ro) _
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Here a * is some value of 8 lying inside the segment [6 o; 8H].
The mean value of the moderator density is determined from the expression
~= 1 H ydz= yn-
As 8H - 8 0, we have, according to expression (18), y -. y'. The minimum mean density of the moderator
in the reactor will, of course, occur for 8H -i SH lim? Since'y can be defined as
V=Volim i ,
Z- H
then, by inserting into the expressions fort and z the values SH = 8H limy we obtain, upon integration,
z
Y = 8H lim
8o+8H lim
1
limy=
z->H Z PH lim
_ Vo
'lim RH lim
Hence, for a given 8o (moderator density or temperature), 8H (reactor power), ar (effective radius of the
active zone), and c 2, determined by expression (4), one can calculate by means of expressions (18), (31), (35),
and (36) the -height of the reactor's active zone, the mean density inside it, the critical charge, and also the
density distribution of the neutron flux and medium over the height.
The characteristic, distribution of the thermal-neutron density over the height and the density of the medium
are shown in the figure and in Table 2. Table 1 gives the results. of the calculations of the height H from formula
(36) and the height Ho from the formula for the height of a reactor with a medium of constant density in the
active zone equal to the density y of the boiling- water reactor. Comparison of H and Ho indicates good agree-
ment of both quantities in the case in which the reactor height does not exceed its diameter and the average
steam content F does not exceed 35%.
The author expresses his sincere gratitude to Ya. V. Shevelev and B. L. Joffe for valuable advice and
A. D. Galanin for discussion of the results of the work and valuable comments.
LITERATURE CITED
1. A. Akheizer and I. Pomeranchuk, Some Problems on the Theory of the Nucleus [in Russian] (State
Technical and Theoretical Press, 1950, 2nd edition).
2. S. Glasstone and M. Edlund, The Elements of Nuclear Reactor Theory [Russian translation] (Moscow, IL,
(1954).
3. A. I. Filimonov et al., Teploenergetika 10, 22 (1957).
4. I. M. Ryzhik and I. S. Gradshtein, Tables of Integrals, Sums, Series, and Products [in Russian] (Moscow-
Leningrad, State Technical and Theoretical Press, 1951), 2nd edition.
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IN A REACTOR FITTED WITH A REFLECTOR
Translated from Atomnaya Energiya, Vol. 9. No. 4, pp. 262-269, October, 1960
Original article submitted November 19, 1959
The conditions of criticality and the distributions of the neutron flux for a homogeneous
thermal-neutron reactor with a system of absorbing rods are obtained in the two-group approxima-
tion . The rods extend through the entire depth of the reactor and are situated around the circum-
ference of the active zone or radial reflector at a uniform distance from one another. The results
of the calculation are presented.
In a number of articles in the literature [1-3], the effectiveness of a system of rods located in the active
zone of thermal reactors is calculated in a simplified way for the case of reactors without reflectors. However,
in some cases such an approximation is not sufficient, and it is required to calculate the effectiveness of the
system of rods in the active zone of a reactor with a reflector, where the control rods can be located in the
reflector itself. This article is devoted to the consideration of the. effectiveness of a system of rods in the two-
group approximation.
We shall write the two-group equations for the moderator density q(r) and the thermal-neutron density
n (r) [4]:
xLq(r)-q(r)+ krn(r) =09
keff pi
L2 An (r) - n (r) -I- pIq (r) = 0,
The solution for 9 and n
where T is the neutron age; L2 is the square of the thermal-neutron diffusion length; 1 is the lifetime of thermal
neutrons in an infinite medium; .P is the probability of avoiding resonance capture; km is the neutron multi-
plication factor in an infinite medium; keff is the effective multiplication factor; q(r) and n W are functions
of
in a cylindrical reactor without end reflectors has the form
q=S1,j+SON; n=`1,+V2.
Here Sl and S2 are the coupling coefficients;
V, = [Alneln (Vxi - (n/Hex)2 r) + BtnYn (Vxi - (n/Hex)2 r)] [cos ncp +Etn sin nW]; (2a)
n=0
795
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2 = [A2nIn ( (/x2 + (n/Hex)2 r) + B2,Kn( 1/"X2 + (n/Hex)2 r)] [cos n(p + E2,, sin ncp], (2b)
n=0
where An, Bn, and E are constants determined from the boundary conditions; c7 n, Yn, In, and Kn are Bessel
functions [5]; K i , K 2 are the roots of the two- group equation of criticality
= (I + alt) (1 -{- a2L2);
Hex is the reactor height and includes the extrapolation distance and effective additions from the end reflectors.
In obtaining relations (2a) and (2b), we assumed that it is possible to separate the variables and that the first root
k
of Eq (2c) is positive, since in the reactors of interest to us one usually has kef ff > 1.
The solution for the neutron flux with a system of rods in the reactor can be represented as the superposition
of two partial solutions, one of which is always regular, while the other (irregular) has singularities at the absorbing
rods:
= S (R) (ir) S (i(R) + *(ir)) ( 1R) + iir)) + ( IR) .2ar))
where the superscript R refers to the regular solution and it to the irregular solution.
The solutions (3) for q and n should satisfy the following boundary conditions:
1) q and n are bounded at each point of the reactor and vanish at the extrapolated radius of the reactor
Rex?
2) At the boundary between the active zone and the reflector the following relations hold:
dnl dn71 I Il dqI dgll
dr = TO dr ; q = Yt9 , dr =1'2 dr
where I and II refer to the active zone and reflector, respectively.
3) At the surface of the rod, the conditions for g and n have the form [6, 7]
dq/de 1
q - d
do/de 1
Q=a' n 1' Q
where a is the geometrical radius of the rod.
We shall now proceed directly to the derivation of the equations of criticality.
System of Rods in the Reactor Reflector
In a reactor with a system of rods located at the circumference of the reflector at a uniform distance from
one another (Fig. 1), the solution for the thermal neutron density and moderator density has the form'
' If the addition theorem is used, the terms with Im (v p i) and Im (p p i) can be combined with the terms con-
taining InN (v r) and InN (pr), respectively.
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III = L, [A7 (xr) + A2nI,,N (Pr)] cos nNw;
n=0
I = Y, ISIA1nclnN (xr) + S2AinInN (pr)] cos nNT;
CO N I
nil = J ICInjfN (?r) + DInKfN (?r)] cos n4p + Y 'V BimKm (t ) cos mwi + (6)
M=0 i= I
?O CO N
i
+ I. IC2njnN (vr) + D2nKnN (vr)] cos nNq) ? 1, 1 B
n=0 m=0 i=i 2m xm (vei) cos mwi;
CO co N
9tI S3 {n o IC2njnN (vr) 4- D2nKnN (vr)] cos nNq) + I Y B2mK,n (vQi) cos mwi}.
M=O i=1
Here N is the number of rods; i is the ordering number of a rod;
x -'~xl - \ Hex J 2 ; N = /xz + (Hex )2;
2 ;
?-Y(LII)+ 2
(Hex )
1 Tr X12
v= 7II +\-Ile x) .
Fig. 1. Position of the rods in
the reactor reflector.
In obtaining, the solution (6), we took into account the fact that in
the system chosen for designating the angles, thesolution should be
symmetric with respect to rp and. wi and periodic in (p:
n(r, W)=n(r, T-}-cti); 4(r, p)=4(r, q +ai)
From the boundary conditions for the moderator density (4) between
the active zone and the reflector, it follows that
00
A2n = Alnl n - N 2 B2mDnm;
to=0
where we have introduced the following notation:
CO
= Aln(fn + N - B2mxnmt
a
JnN- I (aRaz ) - ~7hN+l (a Ra. )
c7nN (aRaz a 2 ; IfN (a Raz
KnN (aRaz )_ -a KnN-I (aRaz )+KnN+i (aRaz )
2
a InN-I (aRaz )+InN+i (aRaz. )
2
Y,nN_ I (aRaz )-YnN+I (a.Raz )
fN(URaz)=a 2
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an clnN ('KRaZ)]
Si [J., (xRaz )--1-2
!n - S2 y2 in I,,N(PRaz)-InN(PRaz)
L Yl a., R )] 1
+ S2fnInN (PRaz )] [VIcc++San] ';
(Pn = [S1J,N (XRaz) +
anm S3y2 1 rnm a't KnN (vRaz )-KnN (vRaz ) ]+(-1)'nvnrn [InN (vRaz )-InN (vRaz )1 }
fin? l n
2 Sr
2 L Y2 an Inv (PRaz )-Inm (PRaz) ]
r " (7a)
Xnm { S3V1 2n [KnN (vRaz ) rnm-(-1)minN (vRaz) vnm)-S2DnmInN (NRaz)} [Y1S3an] InN (vRex) InN (vRex)
hn = InN (vRaz an (vRaz) IC vR '
an =InN (vRaz) -KnN (vRaz) KnN (vReX) nN ( ex)
rnm = InN+m (vRc) +- InN-m (vR0) ; Vnm = KnN-m(vRc) + KnN-+ (vRc);
b n = 1 for n = 0; S n = 2 for n > 1 (Raz is the radius of the active zone; Rc is the radius of the circum-
ference along which the rods are situated).
In obtaining formulas (7) and (7a) we used the addition theorem for Km (? p i) and Km ( u p i) and the
relations for the constants Din and Den obtained from the boundary conditions for q and n at the reactor
surface:
Din= -
InN(PRex) S
An
I
I
N B
R
R
in KN (A Rex)
nN-+. (11
o) +
nN-m (!t
[
1nt
c));
M=O
00
D2n - C2n KnN(URex) N I B2m 2 [I nN?m (vRc) InN-m (vRc)J
From the condition of continuity for the thermal- neutron density at the boundary between the active zone
and the reflector, we obtain
Ain = N (BimRnm + B2mT nm); C1n = N E (BimQnm + B2mAnn) 9
m=0 m=0
fin ln[(-1)n'Inv(?Raz )anm-Knv(#tRaz )bnml-[(-1)minN(PRaz )amn-KnN'(N?Raz)bnmlto
Rnm - 0 2 Yolnzn-untn
Sn [(-1)'nlnv (ftRaz) anm-KnN (flRaz ) bnml un-[(-1)mInN (PRaz ) anm-KnN (ftRaz) bnml'}'ozn
Qnm = 2 Yotnzn - untn
(xnman+(DrrminN(PRaz) 2 rninKnN(vRaz)+(-1)nl2n InN (v Raz )vnm] un -
VoInzn -'untn
Z Xnmhn+onml' (R R K, vR ) r -(-(-1)m 2n I' (vR vnm )
rLY0 [ nN N az) 2 nN ( aZ nm 2 N az )
Votnzn - untn
Tnm = Zn-' [ 0nmtn-1- Xnman+ (PnmInN (PR az ) - 2n- rnmKnN (vR az )+(- 1)m 2n. InN (v R az ) vnm ]
anm = .KnN-m (t1?) + KnN+m ([tRc); vnm =InN { m (?R,,) +InN-m (NR,);
1'N (~R ) - KnN (tRaz) InN (f1 hex) to = InN (?R az) KnN ([tRaz) IfnN (fiRex)
n~^j az R nv ([t eX)
ttn = V nN (xRaz ) + fn1Nm (3Raz ) - y0(Pnhn; Zn = J nN (xRaz ) + fnlnN (NR az) - (Pnan'
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Finally, if we use the boundary conditions (5) for q and n at the surface of any rod (for example, the first)
and the corresponding addition theorem for the Bessel functions for p < Rc and p < pil, then, after some cal-
culations, we arrive at the following equations:'
(B1m [N (g 1 Hnmk + gZhgnmk) + SmhLlit + g11 Alm,,) +
k=0 m=0 n=0
+ B2m [N (glhonmh + g2h3nmh) + 6mhL2k + g2hA2h)) cos ka) = 0;
CO CO
om [ JJ TT
1,~ 1 (BlinNg g11mk + B2m [Ngd2 Onmh + Smhi-!21, + gd A2mk)) cos kw = 0.
h=0 m=0 n=0
Here we have introduced the following notations: S mk =1 for m = k, and S mk = 0 for in k;
glh= IhY?a) -Ih(?a); g2hd= IY(d) -Ik(va);
L1k = Kh (p a) = K'h (?a) ; L2hd KI, (a) - Kh (va);
Aimh = Zh I (Km+h (tt il) cos (m - k) Pi + K.-I, ([tQil) cos (m + k) F'il;
i=2
N
A2mh = Zh N [Km+h (vQil) cos (m - k) Ni + Km-k (veil) cos (m + k)
i=2
H h b 1)hanh QI;,N (l1Rex) a-- n b
nmh - - L [Qnm nh-(- C nm KnN (1LReX) nm) ]
bh r k N (yRex) l .
gnmh = 2 Bnm~n L rnk - (- 1) vnk AN K.1v (vRe7) J '
co sh. n (b - Inv(P1ex) 1 ha
nmk = Z nm L nh KnY (LRex) ) nk ]
n ah [ T r I~,N (vRex) q kz, ) - 1 kv r ] .
nmh = 2 ( nm~n + xnm) ( nk - KnN (vRex) (- ) nk - ( ) nk nm
From Eqs. (10) we can obtain the condition of criticality of the problem if we limit ourselves to the
approximation of the kth order; we then obtain a system of 2 (k + 1) of linear homogeneous algebraic equations
which should be solved for the 2 (k + 1) unknowns (all terms with m > k should be discarded). The condition
that the determinant of this system of equations equal zero will also be a condition of criticality for the problem.
The effectiveness of the system of absorbing rods is then characterized by the difference in keff with and without
the system of rods.
System of Rods in the Active Zone of the Reactor
For a reactor with a system of rods placed in the active zone of the reactor with a reflector, the solution
for the thermal- neutron. density and moderator density will have the following form (the position of the rods and
the notation is the same as before):
In obtaining Eq. (10), it was assumed that d and y are independent of the angle.
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m Co N
I ~ ~
n = [Al.,rynN (xr)+A2nInN (Pr)] cos nN y+ Y,~ I. [B1mYm (xQi)+B2nsKm (PQ,)] cos mW
n=0 M=0 j=1
41= I [S1 A1I ,7nN (xr) + S2A2nI fN (Pr)] cos nNcp +
n=0
00 N R 7
+ [S1B1mYm (xe1) + S2A2mKm (Nei)l cos rn o ;
m=0 i=1
n"1= 11 [C1nInN (?r) + D1nKnN-(?r)] cos nNq + [C2nIn?' (vr) + D2.KnN (vr)] cos nNp;
n=0 n=0
``a
41I = S3 11, [C2nIfN (vr) + D2nKn.N (vr)] cos nNp'.
n=0
From the boundary conditions (4) for q it follows that
aD
A2n = A1n/n + N L' (Blm nm - B2nt(nm); . C2n = A1ngn + N LJ (B1mFnm - B2mLnm), (12)
,n=0 m-0
q)nm - Sn
2
Y2 hu
S1 gn --1 YnN (xRaZ ) - Y1 an YnN (xR az) ]
Vnm = Sz 2 Yi an InN (P R az 1-InN (PR az)
rnm = (- 1)m,7nN+m (XRc) + el nN-m (xRc) ;
Y2 hnK+iN(RRaz)-K,N(PRaz )
bnm i a h 1 , Unm = InN--m (OR.) + InN-m (tRc);
Yl an InN(PRaz )-''N (PRaz )
Fnm = ['S1 2n YnN (x Raz) rnm + S2Vnm InN(YR az )] (Y,S3an]-1;
L..=S2 - - 2 KnN (~ R az) Unm'~ InN (~R az) rnm (Y,S3an]-1.
The functions f n, Vn, an, hn have the same form as the functions f n, Vn, an, hn in (7a).
From the condition of continuity of the thermal-neutron density at the boundary between the active zone
and the reflector,it follows that
0. 00
Ain = N 10 (B1mRnm + B2mrnm]; Cin N 110 [BimQnm+ B2mAnm],
M= M=
bn
Yo1n [ 2 YnN(Raz )rnm+'FnminN(RR
1'Raz )-anFnm] -
-- to [ 2-" YnN (x R az ) rnm+*nminN (PR az )-YohnFnm ]
Rnm = untn-VotnZn
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Vol. an 2 KnN (SR az ) bn
L
(Dnminm (OR az )+ anLnm ] -
-in 2n KfN(pRaz )bnm-(Dnminm(PRaz )+YoLnmhn
Tn,n =
untn-YolnZn
Q8n
nm = [ ZnRnm + 2 YnN (xR az ) rnm + lPnml nN (NR az ) - CtnFnm
Anm = [ zn7'nm + 2n KnN (P R az ) bnm - DnmI nN (P R az ) + anLnm j
to ,
tn1.
The functions un, tn, l n, and zn have the same form as the functions un, tn, i n, and zn in (9).
Finally, if we apply the boundary conditions (5) to q and n, we obtain, after some calculations, the
following equations:
{Blm [N (gikHnmh + g2kgnmh) + 6mhL lh + glkA imk] +
k=0 m=0 n=0
+B2.. [N (glkwnmh + g2ktnmh) + amkL2k + g2kn.2mh]} cos kw = 0;
I I I {Bim [N (S1g;hHn:nh +S2g2kgnmk) +Si
h=0 m=0 n=0 smhLik+'SigikAlmh]
B2m [N (S1 gi hwnmk + S2g2kJtnmh) + S2smhL2h 'I" S29YhA2mh]} cos kw = 0.
Here, we have introduced the notation
= k (d a) l
g1 k d J
- cl h (xa); g2k '= 1 (d a)
- I, (Na);
L1 k d = Yy, da) - Y, (xa); L2kd = Kk (da) - d Kk (Pa);
n
Almh = 2k Y [ 'm+k (xeil) Cos Ni (m - k) + (- 1)k Ym-k (xeil) Cos Pi (m + k)];
i=2
n
A2mh = 2h Y [Km+h (POil) cos Pi (m - k) + Km_k (N il) cos 1'i (m + k)];
1-2
Hnmk = 2 Rnmrnh; 4 k
nmh - 2 (Rnmfn + nm) bnk;
'~
wnmh = 2' Tnmrnk;
7tnmh = 2k (7~nmfn - (Dnm) bnk?
Hence, as in the case already considered, we obtain for the kth order approximation a system of 2(k + 1)
linear homogeneous algebraic equations, which should be solved for the 2 (k + 1) unknowns.
Results of the Calculations and Conclusions
In conclusion, we present some results of the calculations and remarks on the solution of the obtained
equations of criticality.
Figures 2 and 3 shows the variation of the effectiveness of the system of rods and of a single rod, respectively,
as a function of the rod position in the reflector and of the rod radius. Figure 4 shows the change in the inter-
ference coefficient as a function of the number of rods.' As seen from the curves of Figs. 2 and 3, a rather weak
increase in the effectiveness of the system of rods is observed with an increase in their size, and a very sharp drop
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is observed with an increase in the radius of the circle on which the rods are situated in the reflector.
The results were obtained under the assumption that the rods are absolutely black to thermal neutrons and
do not slow down or absorb fast neutrons. The calculation for the rods was limited to the first-order approxi-
mation: k = 0; k =1.
N=36
,c=1,08Raz
/
/N=24
/kc,1 n
N=36
N=24
i N=12
35
n=0-2
. N=
'~-
.~N-24
/
!
\
.
M=12
~i ..~
4=1,254az
/,Rfc=0at the reac
tors center
/
Rc=1,08Raz
n=0
n=0-3
n=0
Rc1,178
Fig. 2. Change in the effectiveness of a system of Fig. 3. Change in the effectiveness of a single rod
rods as a function of their position in the reflector as a function of its position in the reflector and
and of the rod radius (n is the number of terms of radius (n is the number of terms of the series used
the series used in the calculations; Rp = 1.5 Raz). in the calculations; Rp = 1.5 Raz)'
As was to be expected, for small rod sizes it was possible to limit the approximation to the zero order
(k = 0) in the equation of criticality with a good degree of accuracy (in Figs. 2 and 3), the curves for k = 0 and
k = 1 merge), i.e., to neglect the angular dependence of the neutron flux on the rod surfaces. Moreover, from
Fig. 4. Change in the interference coefficient
g as a function of the number of rods N
(Ak eff is the effectiveness of the entire system
of rods; i keff is the effectiveness of a single
rod; Rp = 1.5 Raz; Rc = 1.08 Raz; a = 1.25 cm).
small, the effectiveness of the rod in the system
same place.
the calculations it follows that, for a s fficiently large
number of rods situated at some distance from the edge
of the reactor, one can also neglect in the equation of
criticality the angular dependence of the neutron flux by
setting n = 0. For example, for N = 12, the magnitude of
the reactivity, when the next two higher terms of the
series are taken into account, increases by only approxi-
mately 21o (see Fig. 2). At the same time (see Fig. 3)
for a single rod, the neglecting of the higher terms of the
expansion leads to a considerably smaller value of the
effectiveness (approximately 5% less than the value ob-
tained when the next three higher terms of the series in
n are kept).
The interference coefficient (see Fig. 4) can be
greater or less than unity [8]. This results from the fact
that in the neighborhood of each rod, the relative magnitude
of the neutron density is reduced, and at large distances,
the neutron density is enhanced (as compared to a reactor
without rods). Therefore, when the number of rods is
exceeds the effectiveness of a single rod located in the
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The author expresses his gratitude to Ya. V. Shevelev for discussing basic 'questions connected with this
work, and also to N. N. Ponomarev-Stepnii and E. S. Glushkov for aid, and critical remarks. Numerous and
difficult calculations were performed by R. V. Kulev, to whom the author is verylgrateful.
1. R. Avery, Nuclear Sci. and Eng. 3, 5, 504 (1958).
2. I. Carlvik and H. McCririck, Report No. 152 presented by Sweden at the Second International Conference
on the Peaceful Uses of Atomic Energy (Russian translation) (Geneva, 1958).
3. G. V. Sinyutin and V. G. Semenov, Proceedings of the Second International Conference on the Peaceful
Uses of Atomic Energy, Geneva, 1958 [in Russian], Report of Soviet Scientists, Vol. 2 - Nuclear
Reactors and Nuclear Power (Moscow, Atomic Energy Press, 1959) p. 613.
4. A. D. Galanin, Theory of Thermal-Neutron Nuclear.Reactors [in Russian] (Moscow, Atomic. Energy Press,
1957).
5. G. N. Watson, Theory of Bessel Functions [Russian translation] (Moscow, IL, 1949).
6. D. F. Zaretskii, Proceedings of -the International Conference on the Peaceful Uses of Atomic Energy,
Geneva, 1955 [in Russian]. (Moscow, Acad. Sci. USSR Press, 1958) Vol. 5, p. 624.
7. R. Murray and J. Niestlie, Nucleonics 13, 2, 18 (1955).
8. United States Atomic Energy Commission Reports, Nuclear Reactors, Vol. 1, Physics of Nuclear Reactors
[Russian translation] (Moscow, IL, 1956) p.. 271.
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 Declassified and Approved For Release 2013/02/14: CIA-RDP10-02196R000100060004-4
N. P. Galkin
Translated from Atomnaya Energiya, Vol. 9, No.
Original article submitted April 18, 1960
, pp. 270- 281, October, 1960
This review gives general information on the history of the development and the modern
state of uranium metallurgy outside the Soviet Union. There is a brief description based on
practice in the USA) of methods for producing metallic uranium. The article deals with the
production of the starting materials (U02, UF4), the theory and practice of magnesium-thermic
reduction and the process for refining the heat. There is a discussion of the problem of obtaining
metal from uranium hexafluoride, enriched and poor in U 235 isotope. In conclusion, there is a
discussion of the possibilities for development in uranium metallurgy.
Metallic uranium has been produced on a commercial scale for not more than 15- 20 years, although the
industrial treatment of uranium ores started more than half a century ago. The demand for uranium increased
very rapidly after the demonstration of uranium nuclear fission under the action of neutrons and the use of the
energy of this fission on an industrial scale. The fragmentary data existing at that time on uranium metallurgy
could not be used directly for the production of large amounts of metallic uranium with physical and chemical
properties satisfying the requirements of nuclear physics. Metal for nuclear reactors should be nuclear-pure, i.e.,
free from harmful impurities which are able to capture neutrons to a large extent and which have high density.
Metallurgists were faced with the problem of developing techniques for producing uranium which would remove
the harmful impurities.
Thermodynamic Characteristics of Uranium, Magnesium,and Calcium Compounds Used in Metallothermic
Processes [3]
Com-
Melting
?
eat of forma-
ion at 25 ?C,
Free energy of for-
mation at 1500?C
Com-
Melting,
Heat of forma
tion at 25?C,
Free energy of
Formation at
pound
point,
C
cal/g? atom
kcal g
pound
point, C
kcal/g? atom
500 ? C,
metalloid
v
atom
metalloid
kcal/g? atom
metalloi
metalloid
Decomposes
UaO8
above 900?C
107
-
-
MgO
2800
143,8
90
U02
2800
129,6
2,18
94
CaO
2550
151,7
110
UF4
1036
111
3,6
83
MgF2
1263
131,7
92
UFa
1430
113
3,5
81
CaF2
1418
145,3
111
UC14
590
63
1,9
44
MgCl2
714
76,6
45
Ucla
935
71
1,9
44
CaC12
782
95,2
68
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In a comparatively short time in various countries (USSR, USA, Britain), uranium salts and metallic uranium
were produced which were nuclear-pure, as well as essential chemical reagents and construction materials.
The production of nuclear fuel was then rapidly organized. The world production of uranium [1] has now reached
.considerable proportions, comparable with the production of other rare and nonferrous metals (Fig. 1); and as
regards costs of annual production,it considerably exceeds many of them.
Uranium has a density of 19.05 g/cm3, a melting point of
1130? C,and high chemical activity. It reacts readily with all
metalloids, forming stable compounds. This property determines
the special features of uranium metallurgy.
Nuclear-pure compounds of uranium must be reduced with
powerful and noncontaminating metal reduction agents. The
apparatus and medium in which the metallurgical operations are
100000 i f i- -171 J ;-/-~ Vi-{ carried out should be inert to uranium or protected from reaction
with it.
The history of uranium metallurgy [21 shows that the more
f0000 1 i i r-ri Iti ' T!i or less successful attempts to obtain the pure metal have amounted
to the development and testing of the following methods: " 1) re-
duction of the oxides; 2) reduction of the halides; 3) electrolysis
of the melte; Al thermal decom
osition of the rmmnnnnrle
p
The types of neutral medium used were: vacuum, helium,
argon,and sometimes fused salts of alkali and alkali-earth metals.
100 - - ---H--F--h -~-~ Experience has shown that MgF2, CaF2, BeO, MgO, CaO, Th02
are stable to heated and fused uranium and (at temperatures which
are not too high) graphite, used as refractories and cladding
materials. The usual reduction agents (hydrogen, carbon, silicon,
1800 90 1900 /0 20
Years
alulllluulll/ Yvt.lt. ualauuaul4. t.lult.l .Wilt. lt. l.llt.u lllt.lllt.lt.llt.y Wl lt.
the fact that they contaminated the metal. Good results were
obtained by the metallothermic method for reducing oxides and
halides of uranium by lithium and the alkali earth metals and also
the reduction of chlorides by sodium and potassium. For economic
reasons, only magnesium and calcium were suitable for production
Fig. 1. Comparative graph showing the purposes. Table 1 gives the characteristics of the main compounds
increase in world production of uranium, of uranium, magnesium,and calcium needed for the thermodynamic
molybdenum, nickel,and aluminum. evaluation of metallothermic reduction and electrolysis.
In uranium metallurgy, the first three methods are used for various purposes. The method of thermal de-
composition [4], checked under laboratory. conditions (the decomposition of uranium tetraiodide), was not de-
veloped further due to the complexity of the process and apparatus.
The reduction of uranium oxides (U02, U308, UO3) by calcium in the presence of a flux (CaCl2, excess of
calcium) or by magnesium gives a powder of round particles of metal [5]. This method is used in the powder
metallurgy of uranium.
The metallothermic reduction of halide salts was the simplest, quickest,and cheapest method for preparing
uranium. The reduction of UCI4 by calcium, giving good results [6], did not find widespread application due to
operating difficulties with the strongly hygroscopic chloride. Most metallic uranium is now produced in the form
of ingots by reducing the tetrafluoride with calcium or magnesium.
The electrolysis of melts cannot compete economically with the metallothermic reduction of uranium
tetrafluoride and is only used for preparative purposes (for producing very pure metal from UCI3 [7]). However,
further investigations are being carried out to improve this method and make it cheaper for use both in the
production of the pure metal [8] and also for refining [9].
. Before the First International Conference on the Peaceful Uses of Atomic Energy,it seemed that the main
raw material for producing metallicuranium was the tetrafluoride, reduced by calcium or magnesium. The
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Second International Conference on the Peaceful Uses of Atomic Energy showed that in foreign practice greater
emphasis is placed on the magnesium-thermic method, which is now used not only in the USA,but also in
Canada and Britain [10, 11]. In Italy the double fluorides are reduced by magnesium [121. In France [13] and
Sweden [14] uranium tetrafluoride is reduced by metallic calcium.
The advantages and disadvantages of using these metal reduction agents are as follows:
1. The specific consumption of magnesium per unit of reduced uranium tetrafluoride is 1.6 times less
than calcium.
2. The cost of magnesium is 5-10 times less than calcium.
3. Methods for producing magnesium have been developed to a high level and can ensure a minimum
content of impurities in the pure metal.
4. Reduction by magnesium must be carried out in a closed bomb and the charge must be heated before
the start of reaction. This is due to the fact that, in the first place, magnesium boils at 1130 ? C. i.e., below
the melting point of the slag (the melting point of MgF2 is 1263 ? C), and. in an open apparatus it would vaporize,
without reacting with the uranium tetrafluoride; in the second place, because the heat of reduction of uranium
tetrafluoride by magnesium is insufficient (- 83.5 kcal) to melt the reaction products.
5. The 1 ium-thermic reduction of uranium tetrafluoride can be carried out in an open apparatus
without preliminary heating of the charge. The extraction of uranium from slags based on calcium fluoride is
also much easier than from magnesium fluoride.
The Calcium- Thermic Reduction of Uranium Tetrafluoride. As yet,there is no literature on detailed in-
vestigations of this process not requiring complex apparatus.
In most countries where the calcium-thermic method of reduction is used, the main efforts have been
directed at improving the reaction apparatus and the ignition device [13, 14]. There are descriptions [13] of
reduction heats in which crude ingots of metallic uranium are obtained weighing not more than 100 kg. These
ingots have a perfectly satisfactory surface state and sufficient purity, which means that the metal can be used
to make rods and fuel elements for reactors without further treatment.
In Sweden [14],where the heats are much smaller (the ingots only weigh 30 kg),the crude uranium is sub-
jected to vacuum,refin ing- remelting with simultaneous casting of the uranium rods.
The Magnesium-Thermic Reduction of the Double Fluoride of Uranium. The double fluoride is obtained
by precipitation according to the reaction [12, 15]
UO2(NO3)2 + HCOOH + 2SO2 + 5NaF = NaUF5 + 2NaSO3 + 2N aHCO3 + C02-
It is readily precipitated from solutions and is easily dehydrated at about 120 ? C.
The possible advantages of this method are:
1) the elimination from the production cycle of intermediate precipitation of the higher oxide and re-
duction to uranium dioxide;
2) complex apparatus is no longer needed in the furnaces for reduction and fluorination;
3) the elimination of reactions with expensive pure hydrogen fluoride, which is replaced by the cheaper
solid fluoride of an alkali metal;
4) lower consumption of reagents;
5) a complex composition slag is obtained, melting at about 1000? C. The magnesium-thermic reduction
of the double fluoride uses the same method as the reduction of uranium tetrafluoride. The charge is heated to'
750 ? C. Up to 60 kg of uranium is obtained in the heat.
The double fluoride can serve as the starting product for the preparation of uranium hexafluoride.
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The Magnesium-Thermic Reduction of Uranium Tetrafluoride. This process is used in its most highly
developed form in the USA [16, 17]. The reduction (see the flow diagram, Fig. 2) is carried out in two variants.
The crude metal of the usual reduction heat (140-180 kg) is subjected to refining by remelting in vacuuin,or
large scale reduction smelting is carried out (approximately 1500 kg) providing the pure metal directly, i.e.,
direct reduction is carried out.
The starting material is uranium trioxide, obtained by the continuous roasting of pure uranyl nitrate hexa-
hydrate at temperature of 510-538? C. Table 2 gives the specifications for this material. The impurities to be
checked are divided into groups characterizing the role of the impurities in the process. Metallurgical impurities
affect the physical properties of uranium, its workability and behavior in subsequent heat-treatment and radiation
in the neutron field of the reactor. Technological impurities appear due to contamination of the semifinished
uranium products by the material of the apparatus (stainless steel, monel, etc.) and reagents (for example,
magnesium) and partially affect the quality of the metal, in the same way as the metallurgical impurities. The
forbidden impurities reduce the possible degree of burn-up of U zss in the nuclear reactor.
TABLE 2
Specifications for Semifinished Product and Metallic Uranium [171
Impurity, 10-410
t
i
l
M
Metallur
gical
Technological
Forbidden
Addi
i
l
er
a
a
t
ona
H I
C
N
[St
Fe
Cr
Ni I
Mg
B
Gd
Dy
Cd
Ag
requirements
U0$
-
100,
-
20
30
10
15
-
0, 2
0,05
0,1
0,2
1
Weight of shakedown equal
to 4.2-4.3 g/cros
UF4
-
-
-
-
35
5
15
-
> 96,5 weight %;
oistur0,07 wei lit %;
UOZF> 2,3 weig t To;
ranium oxide > 1,2
eight %**
U
1
400
50
50
50
20
40
5
0,2
I -
-
0,2
1 1
Density 18.96 g/cros
Direct reduction 4,5 30 10 20 45 6 20 10 0,15 - - - - 3
Density 19.01 g/cm
U
Determined by dissolving in water.
Determined by dissolving in ammonium oxalate.
Uranium trioxide is reduced by hydrogen or by cracked ammonia to uranium dioxide at a maximum
temperature of 625-650 ? C in various types of continuous reduction reactors.
Uranium dioxide is subjected to hydrofluorination in continuous reactors. In an advanced type of such a
reactor with a moving layer ,the sintered U02 is hydrofluorinated [18], moving through a vertical tapered monel
tube countercurrent to the hydrogen fluoride flowing upward. The temperature in the upper part of the reactor
is 300-400? C and in the lower part,500-600? C. The consumption of HF is 110% of the theoretical. The
product contains up to 9801o UF4 (see Table 2).
807
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Preparation
UO2
Preparation
UF4
Mixing and
charging
t
Reduction in the
bomb
H2/(N2+3H2)
Regeneration
H2
Regeneration
HF
Lining the body
of the bomb
Slag of MgF2,
housing of bomb
Purification
of direct reduc-
tion of ingot
Slag to regene-
ration
Machining
of ingot
Wastes to re-
generation
Finished
metal
Purification of
ingots gots and weigh
I...no.............uo..ue..e.. .......nn..................uu........
Charging into
crucible
Remelting in
vacuum
Burning wastes
from crucible
Painting crucib1
with MgO
Knocking out
~- ingot
Cutting ingot
and weighing
To remelting
...............
Finished metal-
Fig. 2. Flow diagram for the preparation of metallic uranium [17]. The thick line shows the direct reduction
operations.
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The reduction of uranium tetrafluoride by magnesium is carried out in a closed,but not hermetically sealed,
bomb, carefully lined with dry magnesium fluoride obtained from the ground slag of previous heats. To obtain
the crude metal.a 4% excess of magnesium is taken. The charge is mixed in a double cone mixer and carefully
packed in the body of the bomb with a special charging machine [17]. The density obtained is almost 15%
higher than with ordinary charging. A layer of magnesium fluoride of not less than 30 mm thickness is poured
and tightly tamped on the layer of tamped charge. The bomb is then closed with a steel lid fastened by bolts
placed in a furnace and heated to 700 ? C for about 41/2 hours. Depending on the changes in the composition
Of the charge, the apparatus, etc., the heating of the furnace can vary between 677 and 760? C.
The reduction begins in the range 600-650 ? C by reaction of magnesium vapor with tetrafluoride and
occurs in steps:
UF4 + 0, 5Mg --> UF3 -{- 0, 5MgF2,
(1)
UF3 +1,5Mg -> U +1,5MgF2.
(2)
In order to melt the reaction product the charge must be heated to not less than 200 ? C. The heating is carried
out so that the reaction is excited from the bottom. The reaction front travels at a rate of about 5 mm/sec..
The heating system is determined by the thermal conductivity of the charge and its volume. The larger the
volume of the charge and the lower its thermal conductivity, depending on the density of the charge, the slower
the charge should be heated. If the heating is carried out very rapidly (high temperature), the reaction begins
before the inner layers are heated to the required temperature and insufficient heat is introduced. At a tem-
perature below the optimum in the charge, secondary reactions will rake place over a long period (see below);
the partial reduction [reaction (1)] and the melting will be unsatisfactory.
In Britain [19- 21], instead of a powdery charge they use brickettes made at a pressure of 1.1-1.4 tons/cm2,
with diameter 60-100 mm, height 50 100 mm, and density of 3 g/cros.
Investigations [11, 17J have shown that the yield of metal as ingots depends not only on the type of heating
of the charge but also on the content of impurities in the uranium tetrafluoride (moisture, oxide,and uranyl
fluoride) and magnesium (see Table 2).
The role of these impurities can be clearly seen from the example of secondary reactions of uranyl fluoride
U02F2 + H2O U03 -E- 2HF,
Mg -}- 2HF < MgF2 -I- H21
U02F2 + H2 U02 -{- 2HF,
U02 + 2Mg t U + 2Mg0.
(3)
(4)
(5)
(6)
The reaction cycle (4) + (5) and others in the presence of the smallest traces of moisture (3) lead to the formation
of a film on the particles of MgF2 and to an increase in the time of heating of the charge (increase in temperature
at the start of the reaction) due to difficulty in vaporizing the magnesium. The quantity of heat transmitted to
the charge then increases. Reduction of the uranyl fluoride [reactions (4) + (5) + (6)J gives twice as much heat
as the main reaction (1) + (2), which has a favorable effect on the thermal balance of the process.
The magnesium oxide (6) which forms in the reduction of the uranyl fluoride dissolves in the slag, Small
amounts of it reduce the melting point of the slag (by 40 ? C at 5.5 weight % MgO) ; large amounts increase the
temperature and seriously hinder the separation of the metal.
Experience shows that the best yields (> 97%) are obtained with 1 weight 1o U02F2 and 1 weight % of
uranium oxides. A content of U02F2 greater than 2.3-2.5 weight % is undesirable since the yield of metal in
this case falls below 95 weight To.
Impurities of oxides give a small amount of heat. The thermal effect of reaction (6) is almost a third of
that of the main reaction. The presence of more than 2 weight To of oxides in the tetrafluoride is therefore not
permissible, since it spoils the thermal balance and the magnesium oxide which forms interferes with the separation
of the slag and metal.
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0 /0 20 1 30 40 50 cm
Fig. 3. Diagram of induction furnace for the vacuum remelting of crude
uranium [17): 1) top cover; 2) valve for inspection window; 3) insulator
cover; 4) cover; 5) screen; 6) crucible; 7) base insulator; 8) slides;
9) top of mold; 10) lever clamp; 11) bottom insert; 12) cooling jacket
for mold; 13) bottom of mold; 14) baffle; 15) support for thermal in-
sulation; 16) ring for crucible base; 17) support ring for zirconium screen;
18) external insulator ring; 19) disc with slot; 20) spiral; 21) plug;
22) cupped support for mold.
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When there is moisture in the charge,the following secondary reactions take .place:
UF4 + 2H20 UO2 + 4HF9
Mg+H2O MgO+H2,
H2 + U U, saturated H2.
With a dolomite lining,the following reaction takes place
CaO -}- 2HF < CaF2 + H2O. (10)
The reaction of hydrolysis (7) and the cycle of reactions (7) and (10), commencing at a temperature of 400 ? C,
oxidize the uranium tetrafluoride. If the charge is kept for a long time while heating above this temperature,
the yield of the metal is considerably reduced [11]. For this reason magnesium fluoride is now used instead of
a dolomite lining,and the water content in the initial products of the heats is kept to a minimum. The hydrogen,
contaminating the uranium is removed during the vacuum remelting.
To improve the thermal balance, some investigators [ 22] suggest carrying out the magnesium-thermic
reduction of uranium tetrafluoride in an atmosphere of oxygen. The success of the magnesium-thermic method
is due to the fact [10] that it uses a method for producing uranium tetrafluoride which does not involve moisture
and oxygen.
The ingots of the crude uranium are subjected to vacuum remelting to refine the metal. At the same
time, billets are cast for further rolling or extrusion [17].
The refined remelted uranium is mainly bottom poured (Fig. 3). The basic material for crucibles and
molds is graphite. In order to improve the refining process and to break up oxide films on the metal, before
the start of the vacuum melting operation or during it, metallic calcium or magnesium should be added [23].
Vacuum remelting is the main method for alloying uranium, preparing its alloys and casting components
from it [ 24- 28].
A characteristic tendency of modern vacuum melting is the increase sizes (the furnace capacity is as high
as 4 tons of metallic uranium) and greater accuracy is. casting (deviations of t 15 p with an external diameter
of components 41 mm) [29, 30].
The direct reduction of 2000 kg UF4 is carried out with an excess of magnesium (0.5 weight %) giving a
satisfactory yield of metal and, which is more important, reducing the vapor pressure of magnesium in the bomb.
The mixing and loading of the charge in this process is the same as in the smelting of the crude metal. The
bomb is placed in a 400 kw electric resistance furnace (Fig. 4) with three heating zones. During the first five
hours all three zones are at 621 ? C. The upper and middle zones are then switched off, and after nine hours the
reaction occurs. The temperature in the furnace can increase, depending on the physical and chemical properties
of the charge, the quality of the lining, etc. to 673? C. There can also be fluctuations in the time the charge
is heated before the start of reaction. The hot, low thermal conductivity slag of magnesium fluoride which
collects over the molten metal forms a "hot cap" in the receiver, which retards the cooling of the ingot and c
causes directional crystallization in the metal from the bottom to the top. This results in very good separation
of the slag and very pure metal is obtained (Fig. 5) with a yield of 97% and more. However, the side surfaces
and especially the top of the ingot are contaminated with slag inclusions. The surface of the ingot is therefore
machined on a lathe and a slug of pure metal is obtained with a direct yield of 82-85%. The direct reduced
metal is purer than the ordinary crude metal as regards most of the impurities (see Table 2),but contains more
hydrogen [see reaction (9)], which is undesirable since hydrogen interferes with the sealing of the uranium blocks
and spoils their performance in the reactor. More work must therefore be done on the direct reduction process.
Examples of measures for reducing the amount of hydrogen in the metal are [171 the thorough removal of
moisture from the lining and the components of the charge, the removal of moisture by selecting a system for
heating the charge (experience shows that if the whole charge is heated to above 370? C before the reactions,
the amount of hydrogen is reduced to the normal value), sweeping the heated charge with a neutral gas before
the start of reaction, etc.
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Fig. 4. Magnesium-thermic direct reduction bomb
on a trolley in an electric furnace.
Fig. 5. Structure of direct reduced
uranium; metal not etched. Bright
field (x 125).
Preparing Uranium Enriched and Poor in U235 Isotope. The starting product for the preparation of enriched
uranium is uranium hexafluoride supplied by the gas diffusion plants. When working with uranium compounds
and metallic enriched uranium, it is essential to carefully observe a new condition - nuclear safety [17]. If the
nuclear safety limit is exceeded ,the critical mass of U is reached and a uranium fission chain reaction starts
with the liberation of a large quantity of heat and nuclear radiation, extremely dangerous for the operating per-
sonnel and contaminating the expensive material and surrounding apparatus. The detailed rules for nuclear
safety are established in each actual case by specialist physicists and will not be discussed here. The general
rules limit the masses, volumes,and concentrations of enriched uranium which can be reprocessed, its arrange-
ment,and the composition of the surrounding medium (water, light materials, etc.) in the operating buildings
and during transport. They also recommend the best shapes for the apparatus and containers for enriched materials.
Uranium hexachloride is reduced to the tetrafluoride and is converted to metal according to the above
described arrangement for producing the crude metal and refining it in vacuum. The method for reducing uranium
hexafluoride to the tetrafluoride has not yet been finalized.
Attempts to use as reducing agents sulfur dioxide, ethylene, trichlorethylene, ammonia, etc. have had
very little success, since they have not been used on an industrial scale [31, 32]. An indication of the lack of
success in this field is the fact that in Britain and the USA for a number of years they have been using a multi-
stage hydrolysis method, the serious drawbacks of which are well known.
The search for methods for reducing uranium hexafluoride is very important not only to increase the out-
put of the gaseous diffusion plants, but also to reprocess uranium oxides and uranium concentrates to uranium
hexafluoride by direct fluorination [33-35]. Furthermore, these methods are essential for reprocessing uranium
hexafluoride which is poor in the light isotope. Serious attention is therefore paid to the reduction of uranium
hexafluoride by carbon tetrachloride [31], both in batches and continuous.
This simple and effective method, giving high-purity uranium tetrafluoride, has all the advantages of a
single-stage process. When it is used for highly enriched uranium, the requirements of nuclear safety can readily
be fulfilled. Furthermore, the comparatively small apparatus can have a very high output. At the same time
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in the unfinished production there is a small amount of the expensive product. Also of interest is the method
'for reducing uranium hexafluoride with hydrogen [32, 36].
Uranium tetrafluoride obtained by the reduction of uranium hexafluoride with carbon tetrachloride or
hydrogen contains hardly any water and has the theoretical composition with regard to fluorine and uranium.
Good yields are therefore obtained in the metallothermic reduction. The by-products of the reduction of
uranium hexafluoride (hydrogen fluoride, fluorine and chlorine derivatives of carbon) can be regenerated and
used in the nuclear engineering industry, and the fluorine and chlorine derivatives can be used separately as
freons.
Possibilities for the Development of Uranium Metallurgy. It is rather difficult to predict the trends in this
new and rapidly growing branch of metallurgy. However, some idea can be obtained. of the development trends
in the metallurgical processes by considering the leading investigations and the changes in requirements for
purity of the metal.
At the first stage of reactor construction ,exceptionally rigid requirements were made on the content of
impurities in the metal. This made it necessary to accurately carry out all preparatory and main operations in
the reduction of uranium. The apparatus for these processes was very complex. The reduction was carried out
in an atmosphere of inert gas - argon, and the sizes of the reduction heats did not reach 100 kg metal.
With development in reactor construction and the development of various types of fuel elements, large
quantities of uranium were required. This made the problem of reducing the cost of metallic uranium particularly
acute; that of simplifying the technological processes for producing the metal and the possibilities of using.these
processes directly at ore plants or in fuel cycles (when reprocessing nuclear materials). also became acute.
It is possible that the production of metallic uranium will develop in parallel with the development of new
methods for producing uranium tetrafluoride, the technology of which is complex and expensive.. It is also possible
that in the near future the use of successive. reactions of hydrofluorination and fluorination in the processing of
concentrates [33, 34] to obtain very pure hexafluoride will make it possible to avoid the multistage .processes of
so-called fine chemical purification, i.e., dissolving the concentrates in nitric acid, extraction with organic
solvents, reextraction and precipitation of the uranium in the form of an insoluble salt with subsequent roasting
of this precipitate to the trioxide. When these operations are eliminated there will no longer be huge quantities
of harmful waste solutions requiring additional treatment.
On reaction with reduction agents (H2, CC14, NH3, C2H2) the obtained uranium hexafluoride is easily con-
verted to uranium tetrafluoride,and the latter is reduced to the metal.
Another very promising trend in the technology of uranium is a method based on the conversion of uranium
oxides to chlorides [37]. The charge, consisting of uranium oxides, barium sulfide,and the eutectic mixture
NaCl + KCl or KCl + BaC12 is heated to 750 800 ? C. Gaseous chlorine is delivered to the heated charge through
a quartz tube. The process of.conversion of uranium oxides to uranium tetrachloride can be represented by the
following reactions:
2BaS + 2UO2 + 6C12 . - .
(11)
--~ 2BaC12 + 2UC14 + 2SO2, 2UC14 + BaS ---> BaCl2 + 2UC13 + S,
(12)
3BaS -{- 2UO2 -{- 6C12 --- 2UC13 3BaC12 + 2502 -{- S.
(13)
After all of the sulfide has changed to chloride, with subsequent chlorination uranium tetrachloride will again
be formed:
2UC13 + Cl2 ---> 2UC14. (14)
The obtained alloy is treated with magnesium,and metallic uranium separates out. This method is very interesting
and it is possible that it will be introduced on an industrial scale after further development.
At the present time up to 2000 kg of uranium tetrafluoride is reduced in one calcium-thermic heat [17].
However, in principle it is quite possible to have a continuous process of reduction of fluorine or chlorine compounds
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of uranium, which will considerably reduce the volume of the apparatus and completely automate the process.
Development work has been done on this continuous process for reducing uranium since 1953 [17], but it is not
yet used on an industrial scale, and it is not likely to be used in the near future. In the continuous process it is
difficult to separate the slag from the metal, which reduces the yield of uranium. Difficulties also arise in
selecting constructional materials for the reduction apparatus. Furthermore, at the present time the scale of
uranium production is not so great as to make it imperative to introduce a continuous process on an industrial
scale. In the near future the attention of scientists and technologists will most likely be concentrated on
developing large-dimension apparatuses which will increase the scale of simultaneous reduction of uranium
From the economic point of view it is perhaps better to produce uranium in the form of an ingot weighing
several tons than to produce it in the course of several days by a continuous process in small apparatuses. The
continuous process would require 24-hour operation at high temperatures, which needs close attention.
A further detailed study of the reduction smelting and refining of the metal in one apparatus should con-
Ma
ly reduce the cost of this metallurgical method.
In connection with the development of vacuum metallurgy and improvements in vacuum apparatus, it may
sible to reduce the cost and introduce into production. the carbothermic process for producing uranium from
the flow sheet of which was recently published [17].
Also very interesting are investigations connected with the use during reduction of a combined charge
fisting, for example, of a mixture of compounds of two or three metals and a reducing agent [171. As fuel
ments in modern reactor construction, more use is being made of alloys of uranium with other metals (diluents).
ny alloys can be produced by metallothermic reduction. There is reason to suppose that work on the prepara-
n of alloys by the metallothermic method will be successfully developed.
Of considerable interest is the development of a process for the direct reduction of uranium hexafluoride to
the metal. In the Oak Ridge National Laboratory the possibility has been shown on a small scale of reducing
uranium hexafluoride by calcium and iodine, the result being compact metallic uranium [17]. Studies have also
been made of the reduction of uranium hexafluoride by magnesium vapors at temperatures of 1000-1500?C. In
this case the product was finely powdered uranium [17]. The promising features of these processes have led to
extensive investigations in this direction. Studies are being made of processes for reducing uranium hexafluoride
by sodium vapors according to the reaction
UF6 + 6Na U + 6NaF,
and also by a mixture of sodium and magnesium vapors; in the latter case compact metal and a very easily
fusible slag can be obtained. However, this process involves increased consumption of the metal reduction
agent; it is therefore necessary to carry out research work to provide cheaper reduction agents.
There is no doubt that the direct reduction of uranium hexafluoride to the metal combined with the above-
mentioned processes of hydrofluorination and fluorination has tremendous technological and economic advantages.
LITERATURE CITED
1. V. Ripar Jaderna Energia 4 (5) 128 (1958).
2. J. Van Impe, Chem. Engng Progr. 50, 230 (1954).
3. A. Glassner, ANL 5750 (1957).
4. C. Prescott, et al., AEC Report MDDC-437 (1946).
5. J. Buddery, Metallurgy and Fuels 1, Pergamon Press (1956) p. 24.
6. J. Goggins, et al., Ind. Eng. Chem. 18, 114 (1926).
7. B. Blumenthal and R. Noland, Metallurgy and Fuels 1, . Pergamon Press (1956)
L. Niedroch. Nucleonics 16, 64 (1958).
9. J. Antill et al., Metallurgy and Fuels 2, Pergamon Press (1959) p. 38.
p. 62.
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10. Melvanin, Proceedings of the Second International Conference on the Peaceful Use of Atomic Energy
(Geneva, 1958). Selected reports of non-Soviet scientists, Vol. 7. The Technology of Nuclear Raw
Materials [in Russian] (Moscow, Atomic Energy Press, 1959) p. 436
11. J. Harper and A. E. Williams, The Extraction and Purification of Rare Metals [Russian translation] (Moscow
Atomic Energy Press, 1960).
12. A. Cacciari et al., Report No. 1399 presented by Italy to the Second International Conference on the.
Peaceful Use of Atomic Energy (Geneva, 1958).
13. Decropet al., Proceedings of the Second International Conference on the Peaceful Use of Atomic-Energy
(Geneva, 1958). Selected reports of non-Soviet scientists, Vol. 7. The Technology of Nuclear Raw Materials
[in Russian] (Moscow, Atomic Energy Press, 1959) p. 485.
14. Gelin et al., Proceedings of the Second International Conference on the Peaceful Use of Atomic Energy
(Geneva, 1958). Selected reports of non-Soviet scientists. Vol.. 7. The Technology of Nuclear Raw Materials
[in Russian] (Moscow, Atomic Energy Press, 1959) p. 417.
15. Brodsky and Pagny, Proceedings of the Second International Conference on the Peaceful Use of Atomic
Energy (Geneva, 1958). Selected reports of non-Soviet Scientists, Vol. 7. The Technology of Nuclear
Raw Materials [in Russian] (Moscow, Atomic Energy Press, 1959) p. 494.
16.
H. E. Thayer, Proceedings of the Second International Conference on the Peaceful Use of Atomic Energy
(Geneva, 1958). Selected reports of non-Soviet scientists. Vol. 6. Nuclear Fuel and Reactor Material
[in Russian] (Moscow Atomic Energy Press, 1959) p. 5.
17.
C. Harrington and A. Ruehle, Uranium Production Technology, Van Nostrand Company (1959).
18.
Yu. V. Gagarinskii, Atomnaya Energiya 6, 2, 124 (1959).*
19.
Metal Ind. 94, 7, 127 (1959).
20.
Williams, Improvements in the Production of Uranium, British Patent No. 780,974 (1957).
21.
Atomic World 10, 3, 99 (1959).
22.
Wotek, Improvements in the Production of Uranium, British Patent No. 806,031,
(1958).
23.
Colbeck, British Patent No. 806,001 (1958).
24.
W. Hayward and P. Corzine, Proceedings of the Second International Conference on the Peaceful Use of
Atomic Energy (Geneva, 1958). Selected reports of non-Soviet scientists. Vol. 6. Nuclear Fuel and
Reactor Material [in Russian] (Moscow, Atomic Energy Press, 1959) p. 561.
25. K. J. Turner and L. R. Williams, Proceedings of the Second International Conference on the Peaceful Use
of Atomic Energy (Geneva, 1958). Selected reports of non-Soviet scientists. Vol. 6. Nuclear Fuel and
Reactor Material [in Russian] (Moscow, Atomic Energy Press, 1959) p. 570.
26. J. Storr, M. Englander, and M. Gotron, Proceedings of the Second International Conference on the Peaceful
Use of Atomic Energy (Geneva, 1958). Selected reports of non-Soviet scientists. Vol. 6. Nuclear Fuel
and Reactor Material [in Russian] (Moscow, Atomic Energy Press, 1959) p. 515.
27. M. D. Jepson et al., Proceedings of the Second International Conference on the Peaceful Use of Atomic
Energy (Geneva, 1958). Selected reports of non-Soviet scientists. Vol. 6. Nuclear Fuel and Reactor
Material [in Russian] (Moscow, Atomic Energy Press, 1959) p. 96.
28. Stivenson, J. Inst. Metals 87, 6, 174 (1959).
29. H. Hardung, Vakuum-Technik 7, 6, 135 (1958).
30. Flancke, Vacuum, April, p. 59 (1959).
31. Nairn et al., Proceedings of the Second International Conference on the Peaceful Use of Atomic Energy
(Geneva, 1958). Selected reports of non-Soviet scientists. Vol. 7. The Technology of Nuclear Raw Materials
[in Russian] (Moscow, Atomic Energy Press, 1959) p. 553.
32. Smiley and Brater, Proceedings of the Second International Conference on the Peaceful Use of Atomic
Energy (Geneva, 1958). Selected reports of non-Soviet scientists. Vol. 7. The Technology of Nuclear
Raw Materials [in Russian] '(Moscow, Atomic Energy Press, 1959) p. 561.
33. Smiley and Brater, Proceedings of the Second International Conference on the Peaceful Use of Atomic
Energy (Geneva, 1958). Selected reports of non-Soviet scientists. Vol. 7. The Technology of Nuclear
Raw Materials [in Russian] (Moscow, Atomic Energy Press, 1959) p. 587.
34. Lawroski et al., Proceedings of the Second International Conference on the Peaceful Use of Atomic Energy
(Geneva, 1958). Selected reports of non-Soviet scientists. Vol. 7. The Technology of Nuclear Raw Materials
[in Russian] (Moscow, Atomic Energy Press,.1959) p. 615.
`Original Russian pagination. See C.B. translation.
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35. Powell, 'Proceedings of the Second International Conference on the Peaceful Use of Atomic Energy
(Geneva, 1958). Selected reports of non-Soviet scientists. Vol. 7. The Technology of Nuclear Raw Materials
[in Russian] (Moscow, Atomic Energy Press, 1959) p. 641.
36. Cuthbert et al., Proceedings of the Second International Conference on the Peaceful Use of Atomic Energy
(Geneva, 1958). Selected reports of non-Soviet scientists. Vol. 6. Nuclear Fuel and Reactor Material [in
Russian] (Moscow, Atomic Energy Press, 1959) p. 551.
37. Gibson and Buddery, The Extraction and Purification of Rare Metals [Russian translation] (Moscow, Atomic
Energy Press, 1960).
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M. A. Stepanov and N. P. Galkin
Translated from Atomnaya Energiya, Vol. 9, No. 4, pp. 282-285, October, 1960
Original article submitted March 18, 1960.
This article gives a calculation of the solubility product of the hydroxide of tetravalent
uranium, a knowledge of which is essential for the efficient operation of many processes in the
treatment of uranium and methods for determining it. The calculation is based on experimental
data obtained in the potentiometric titration of a chloride solution of tetravalent uranium by
alkali. The activity of the hydrogen ions was determined by a tube potentiometer LP-5; the
active concentration of the quadruple charged ion of tetravalent uranium was calculated from
the analytically found concentration of tetravalent uranium with an allowance for hydrolysis
and the value of the ionic strength. It is shown that the required solubility product is equal to
(1.10 ? 0.72) ? 10-52.
The hydroxide of tetravalent uranium is a slightly soluble compound. Knowing the value of the solubility
product of this type of material, it is possible to determine its thermodynamic parameters.
The information in the literature on the solubility product of tetravalent uranium hydroxide is very con-
tradictory. The value of the equilibrium constant given in [1]
U(OH)4 i U4+ +4011-
is )
equal to 10-4. This value was found by a calculation method (by comparing the ionic radii and the solubility
products of thorium and tetravalent plutonium) and therefore requires experimental confirmation. There are also
data [2] on the solubility product of tetravalent uranium hydroxide, calculated on the basis of the reaction
UO(OH)2 U02+ + 20H-.
The constant of the given equilibrium is equal to 1.3 ? 10- 8, and the change in free energy is equal to 41
kcal/mole. However, the existence of the UOZ+ ion is open to doubt [3]. The low solubility of tetravalent
uranium hydroxide is the basis for the analytical methods for separating tetravalent and hexavalent uranium [4]
and also a number of technological processes [5-8]. It is therefore important to determine the solubility product
of tetravalent uranium hydroxide.
The basis for determining the solubility product was the potentiometric method, one of the most reliable
in this type of experiment [9]. The point of inflection of the potentiometric (more accurately the pH- metric)
curve corresponds to the start of formation of the hydroxide precipitate. All the pH-metric measurements were
made on a LP-5 tube potentiometer, which gives an accuracy in this range of ? 0.05 units. All hydrolysis of
tetravalent uranium occurred in the pH range between 1 and 6, i.e., in the region where a linear relationship
is observed between the potential and pH. The potentiometer was adjusted and checked with a series of acetate-
chloride buffer solutions prepared by the method described in [9].
817
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Chloride solutions of tetravalent uranium were used, prepared by dissolving uranium metal of 99.761o
purity [10]. The insoluble residue was filtered off after being allowed to stand. The tetravalent uranium solution
was kept in an opaque flask, which reduced the rate of oxidation [11]. The air was removed from the flask by
oxygen-free argon [12]. The oxidation of the tetravalent uranium was checked by reaction with potassium ferro-
cyanide [13, 14]. During 15 days no oxidation of uranium was detected.
0 1 2 3 4
Molar ratio OH;U
A typical curve of a pH titration of
chloride solution of tetravalent uranium
with alkali. Concentration of tetravalent
uranium about 1.10-2 - 1. 10-3 g ? ion/
/ liter.
The concentration of tetravalent uranium was deter-
mined by titration with potassium bichromate. All solutions
of reagents were made from analyzed chemicals and boiled
distilled water. the content of excess acid was found by
titration with alkali using phenolphthalein as the indicator.
Allowance was made for the amount of alkali used in forming
the hydroxide of tetravalent uranium. The concentration of
the initial solution was 0.590 M with respect to uranium and
1.02 M with respect to hydrochloric acid. To obtain smaller
concentrations of uranium,the solution mentioned, was diluted
with distilled water.
The precipitating agents were solutions of ammonium
hydroxide, caustic soda,and caustic potash (graded 'pure for
analysis') of concentrations 0.464, 1.992, 2.184 N, re-
spectively.
When carrying out the experiment,a fixed amount of
initial solution was placed in a reaction vessel, into which
portions of the precipitation agent were added from a micro-
burette with continuous mixing. The pH values of the reaction
medium were determined with a glass electrode in the system,
3 and 8 minutes after introducing the portion of precipitation agent. There was practically no difference between
the first and second measurements of pH. When pH = 7-9 was reached,the experiment was stopped since further
additions of precipitation agent led to an increase in alkali content which was not connected with hydrolysis of
uranium. The dependence of pH on the amount of precipitation agent added to the system was shown graphically
(see figure). The inflection points of the curve near the molar ratio of alkali to uranium equal to two corresponding
to the start of formation of the hydroxide of tetravalent uranium.
It was found by the pH titration that in a solution containing 5.75.10-g g ? ion/liter of tetravalent uranium
and 3.28.10-2 g -ion/liter of chlorine, the formation of uranium hydroxide begins at pH = 2.38 (the average from
several experiments using the above precipitation agent). On the basis of these data a calculation was made of
the solubility product SP of the hydroxide of tetravalent uranium: SP = [U4+] [OH -14, where [U4+] and [OH -]
are the equilibrium concentrations of the quadruple-charged ion of tetravalent uranium and hydroxyl. The cal-
culation of the solubility product from the pH where the hydroxide begins to form is described in the literature
[9, 15-171; in particular this method was used to determine the solubility product of the hydroxide of tetravalent
plutonium [1].
After substituting the ionic product of water and the concentration of hydrogen ions and converting to
logarithms, we obtain a trinomial
pSP = npJ + pM - npH, (2)
in which pSP is the negative logarithm of SP; pJ is the negative logarithm of the ionic product of water; pH is
the pH value for the start of formation of the hydroxide; pM is the negative logarithm of the cation concen-
tration at the instant of formation of the hydroxide; n is the charge on the cation.
In calculating the equilibrium concentrations, allowance was made for the processes of hydrolysis and
complex formation of tetravalent uranium in the solution. The following equations were used:
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U4+ + Cl- = UC13+, k j = 1,91;
t = [U4+)[Cl_) ,
U4+ + 2Cl = UC12+ K2 _ ][Gj )
2 , _ - [U4+)[Cl_72 =1,27;
U *;
U4+ + H2O = UGH3+ + H+, K3 = [UO[ ],H+) - 7,25. 10-2
U4. + 2H20 = U(OH)2++ 2H+, K4 = [U(OH)rlEH+)2 = 5,25.10-3 *.
[181
Using the laws of acting masses and conservation of mass together with the above equations, we obtain the
polynomials
('u = [U4+](1-[- K1[CI ] +K2[Cl-]2 -[-K3/[H+] +K4/[H+]2),
Cci [Cl ] -[- K1[U4+]tC1 ] + 2K2[U4+][C1 ]2,
(3)
(4)
where CU and CCl are the total, analytically determined concentrations, respectively, of the tetravalent uranium
and chlorine.
We assume that q = 1 + [H3l+j . Solving Eqs. (3) and (4) relative to [Cl-1, we obtain the identity
K2 [CI -J3 + (K1 + 2K2Cu - K2 CC,) [C1 ]2 4- (q + K1Cu - K1Cci)[Cl ] - gCC1 _ 0.
(5)
After substituting the constants of the reactions and the values of the corresponding concentrations in identity
we confirm that [CI-] = CC,. Solving the system of polynomials (3) and (4) relative to the [U4+], we obtain
[U4+] = 1,79.10'5, g. ion/liter [UGH"] = 3,16.10-4 g- ion/liter [U(OH)Z+] = 5,42.10-3 g? ion/liter
The equilibrium active concentrations must be used in calculating the solubility product [201.
We determine the activity of the U4+ ion from formulas based on the Debye-Hiickel theory [21],
a= Ig A A= 1,823.10en1?n2.
1' [U4+]; V Y -ft; (ST)3/2
where nl and n2 are the charges on the ions forming the salt; T is the absolute temperature; a is the dielectric
constant. of the solvent (for water this is 80 [22]); a is the ionic strength of the solution; Ci is the total con-
centration of the ion i with charge zi.
In calculating the activity of tetravalent uranium the following values were assumed: pJ = 14.07 [23];
pH = 2.38; [U4+] =1.79.10-5 g ? ion /liter; T = 293? C; A = 2.075; ? = 4.73.10-2 (concentrations of chlorine
and ammonium ions equal to 3.28 ? 10-2 g? ion/liter); y = 0.353. Consequently, the activity of the quadruple-
charged ion of tetravalent uranium is equal to 6.32. 1.0-6 g ? ion/ liter (pMactive = - log a = 5.20). After
? The value of the data for the equilibrium constants were obtained by extrapolation of values given in [19], to
zero ionic strength and subsequent potentiometric measurement.
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To calculate the absolute error in the solubility product for tetravalent uranium hydroxide we use the
mula given in [24]:
Ay_ (8a,) Aal+(bat) Aa2 ...,
ere Ay is the error in the value of 1, being a function of the variable al,
8f - the partial
derivative of the function f.
In the case under consideration, when the solubility product of the hydroxide is determined by Eq. (1),
we have
ASP = (J ? 10PH)4 {A tU40] + 9,2[U41 I AAH}. (6)
As mentioned above, the accuracy in measuring the pH is 0.05 units of pH. The relative error in deter-
mining the content of tetravalent uranium is due to the value of the least accurate calculated constant of
hydrolysis (error 2016); consequently, A (U4' J=1,26.10-e g - ion/liter. In this case ASP= t 0.72.10-52.
When calculating ASP it is found that the error is mainly due to the error in, measuring the pH of the medium.
The solubility product of tetravalent uranium hydroxide has therefore been determined by pH titration
and the calculation of the equilibrium-active concentration of the quadruple charged ion of tetravalent uranium
during hydrolysis. Its value is (1.10 ? 0.72)- 10-52.
1. W. Latimer, The Oxidation States of Elements and Their Potentials in Aqueous Solutions [Russian translation]
(Moscow, Foreign Literature Press, 1954).
2. K. Gayer and H. Leider, Nucl. Sci. Abstrs., No, 3321, 398 (1954).
3. K. Kraus and F. Nelson, J. Am. Chem. Soc. 72, 3901 (1950).
4. Tsubaki and Hara, Japan Analyst 4, 357 (1955).
5. I. Ya. Bashilov, An Introduction to the Technology of the Rare Elements [in Russian] (Moscow, State
Chemical and Theoretical Press, 1932).
6. N. P. Galkin, A. A. Maiorov, and U. D. Veryatin, The Technology of Uranium Concentrate Processing
[in Russian] (Moscow, Atomic Energy Press, 1960).
7. R. Gelin, G. Mogard, and B. Nelson, Proceedings of the Second International Conference on the Peaceful
Use of Atomic Energy (Geneva, 1958). Selected reports of non-Soviet scientists. Vol. 7. The Technology
of Nuclear Raw Material [in Russian] (Moscow, Atomic Energy Press, 1959) p. 417.
8. D. Kaufman and S. E. Bailey, The Extraction of Uranium from Ores. U. S. Patent No. 2,780,519 (1957).
9. H. Britton, Hydrogen Ions [in Russian] (Moscow, United Scientific and Technical Press, 1936).
10. The Analytical Chemistry of Uranium and Thorium. Edited by Rodden [Russian translation] (Moscow,
Foreign Literature Press, 1956).
11. V. N. Ushatskii and Yu. M. Tolmachev, Trudy Radievogo Inst. 7, Khimiya i Tekhnologiya, 98
(1956).
12. K. V. Chmutov, Techniques of Physicochemical Investigations [in Russian] (Moscow-Leningrad, State
Chemistry Press, 1948).
13. N. A. Tananaev and P. A. Panchenko, The Drop Method for Detecting Titanium and Uranium [in Russian]
(Kiev, 1926).
The value for the solubility product of tetravalent uranium hydroxide which we calculated previously [6] differs
from. this value since in the previous work no allowance was made for the hydrolysis processes.
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14. N. A. Tananaev, The Drop Method [in Russian) (Moscow, State Chemistry Press, 1954).
15. I. M. Korenman, Zhur. Obshchei Khim. 25, 1859 (1955).
16. I. M. Korenman, F. S. Frum, and A. S. Kudinova, Collection of Papers on General Chemistry, Vol. 1
[in Russian] (Moscow, Acad. Sci. USSR Press, 1953) p. 83. .
17. V. L. Zolotavin and V. V. Sergovskaya, Trudy Ural'skogo Politekh, In-ta, 57, 66 (1956).
18. R. Day, R. Wilhite, and F. Hamilton, J. Am. Chem. Soc. 77, 3180 (1955).
19. S. Hietanen, Acta Chem. Scand. 10, 1531 (1956).
20. A. F. Kapustinskii, Zhur. Priklad. Khim. 16, 51 (1943).
21. V. A. Kireev, A Course of Physical Chemistry [in Russian] (Moscow, State Chemistry Press, 1955).
22. D. Kay and T. Labey, A Handbook of Practical Physics [Russian translation] (Moscow, Foreign Literature
Press, 1949).
23. Yu. Yu. Lure, Calculation Tables for Chemists [in Russian] (Moscow, State Chemistry Press, 1947).
24. N. K. Vorob'ev, V. A. Gol'shmidt, and M. Kh. Karapet'yants, Practical Physical Chemistry [in Russian]
(Moscow-Leningrad, State Chemistry Press, 1950).
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 Declassified and Approved For Release 2013/02/14: CIA-RDP10-02196R000100060004-4
THE SORPTION EXTRACTION OF URANIUM FROM PULPS
AND SOLUTIONS
Translated from Atomnaya Energiya, Vol. 9, No. 4, pp. 286- 296,
October, 1960
Original article submitted April 14, 1960
Extensive use is made of ion-exchange techniques in the uranium industry. Sorption from
acid and carbonate solutions and pulps is used in the treatment of uranium ores. and the production
of pure compounds of uranium. This article deals with the basic regularities of these processes.
It gives the characteristics of various types of ion-exchange materials, which can be used for the
selective absorption of uranium, and it also deals with a number of accompanying elements. The
article also describes the various types of apparatus used in the sorption of uranium from pulp,
making it possible to process all pulps containing up to 4056 solid material. Further improvement
in sorption techniques is connected with the use of ion-exchange materials having high kinetic
characteristics and high selectivity with regard to uranium.
Experience in the industrial treatment of uranium ores has shown that the extraction systems, as well as
the sorption systems. are very efficient economically.
In the leaching of uranium,a large quantity of accompanying impurity goes into solution, such as iron,
aluminum, magnesium, alkali metals, etc. Methods have recently been developed which considerably reduce
the quantity of impurities which are dissolved when uranium ores are leached. However, in a number of cases
the content of the salts in solution for various types of ores can vary over wide limits from 2 to 30%.
When most of the ore mass is separated by filtration methods .large floor areas are taken up with the filtering
apparatus. In order to reduce the losses of uranium when separating the hydrate cakes of iron and aluminum, many
repulping operations are needed. The most important problem in the hydrometallurgy of uranium ores is,therefore,
to develop methods allowing the selective absorption of uranium from solutions and ore pulps having complex salt
composition, and also the intensification of the filtration process or the development of filtrationless methods,
eliminating the most laborious and power-consuming operations. These problems are best solved on the basis of
sorption techniques.
Ion-Exchange Resins for Uranium Extraction
Cation-exchange materials and various types of ion-exchange materials are used in the Soviet Union for the
absorption of uranium [1]. As first shown in [2], in sulfate solutions, particularly important for the hydrometallurgy
of uranium ore, the uranium is present in cationic and anionic forms, the relationship between which depends on
the concentrations of the sulfate ions and uranium. It can therefore be assumed that the uranium will be sorbed
from sulfate solutions by cation-exchange and anion-exchange materials. This was confirmed in the study of sorption
of uranium from pure solutions containing no impurities. One of the production problems which arises is sorbing
uranium from solutions with a complex salt composition. For the selection and correct evaluation of sorbents
which selectively absorb uranium, it is essential to have data on the effect of the salt composition on the sorption
of uranium.
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Only sorbents which retain an absorption capacity for uranium at high salt composition of the solutions
can be of any practical interest. Figure 1 shows the relationship -between the sorption capacity for uranium of
a sulfo- cation- exchange material KU-2 and a carboxyl cation-exchange material type SG-1 on the concentration
of sodium sulfate at pH = 3.5. As can be seen from the diagram, with a sodium sulfate concentration in the
solution of more than 10 g/liter,the sorption capacity for uranium of the sulfo-cation-exchange material KU-2
(sulfonated copolymer of styrene with divinylbenzene) approaches zero, which indicates the sharp drop in
sorption capacity for uranium of strongly acid cation-exchange materials with increase in the salt composition
of the solution. On the basis of similar facts, workers in various countries [3] have decided on the unsuitability
of cation-exchange materials for the sorption extraction of uranium. Carboxyl resins, and also other ion-exchange
materials with weakly acid ion-exchange groups,have been dismissed due to the low degree of dissociation of the
corresponding weakly acid ion-exchange groups in acid medium and, consequently, their low value of exchange
capacity.
25 50 75 /00 125 /50
Concentration of Na2SO4 , g/ liter
Fig.. 1. The dependence of the sorption capacity
for uranium of a sulfo-cation-exchange material
KU- 2 (?) and a carboxyl cation-exchange mat
material type SG- 1 (0) on the concentration of
sodium sulfate .
00
160
120
Cd
0 80
4.
15 40
Cd
a
Cd
U
0 2 4 6 8 10
Quantity of binder, %
-13,0
Fig. 2. The dependence of the swelling factor (?)
and sorption capacity (o) for uranium of copolymers
of methacrylic acid on the quantity of binder.
In 1948, the author found that carboxyl resins
based on acrylic and methacrylic acids selectively
absorbed uranium from weakly acid solutions containing high concentrations of salts (150 g/liter of chlorides,
nitrates and sulfates). Due to the low degree of dissociation, the ion-exchange properties under these conditions
are considerably depressed, and therefore the ballast impurities sorb in small quantities and the uranium,together
with the ordinary ion-exchange, can be absorbed due to the formation of complexes as in the case of reaction with
carboxylic acids such as oxalic or acetic acids. The industrially produced ion-exchange materials based on
acrylic and methacrylic acids (KB-4, amberlite IRC-50, etc.) are intended for use in a medium close to neutral
or even weakly alkaline. They contain a large amount of binder*, and due to the low capacity they are not suit-
able for the extraction of uranium from acid industrial solutions. The sorption capacity for uranium of copolymers
of acrylic and methacrylic acids depends on the chemical nature of the binder and its quantity. This dependence
was studied by the author with co-workers for various specimens of carboxyl resins synthesized on the basis of
acrylic and methacrylic acids using various binders.
Figure 2 shows the dependence of the swelling factor and the sorption capacity for uranium of methacrylic
acid copolymers on the quantity of binder (dimethacryl ester of ethylene glycol). It can be seen from the figure
that the swelling factor increases linearly with increase in the content of binder between 1 and 12 %, whereas
the maximum sorption capacity for uranium is achieved with a content of 3.5-516 binder in the copolymer - an
ester of methacrylic acid and ethylene glycol or di- and triethylene glycol.
The SG-1 resin is made by the method of pearl polymerization and has high mechanical strength. It has
been shown that SG-1 type copolymers in large amounts begin to sorb uranium at pH = 1.7-1.9; with further
increase in pH to 3-4,the sorption capacity for uranium increases several times.
* The amount of binder means the content of transverse-binding reagent such as divinylbenzene, the dimethacryl
ester of ethylene glycol, etc. (as percentages of the monomer).
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Figure 3 shows the dependence on pH of the sorption capacity of SG-1 resin for uranium and some frequently
encountered accompanying impurities. As follows from Fig. 3, pH = 2.8-3.5 is the optimum value in the absorption
of uranium by SG-1 type carboxyl resins; in this case there is high sorption capacity for uranium and the selective
properties of the sorbent are retained. With increase In the pH of the solution (up to 3.8-4.0) due to increase in
the dissociation constant of the ion-exchange groups,there is an increase in the exchange capacity of the SG-1
resin. There is an increase in the absorption of accompanying impurities, which leads to a reduction of the
selectivity of the copolymer for uranium.
The trivalent iron forms rather stable complexes
with copolymers such as the SG-1 resin, which are poorly
x/00 desorbed under ordinary conditions of regeneration.
N However, at values of pH = 2.8 - 3.5 in solutions and
80 pulps, the content of the ferric ion does not exceed
10-15 mg/liter, which is due to the low sorption
60 -- capacity for it in the working range of pH values.
0
40 sorbed in small quantities and this is not reflected in
m the value of the sorption capacity for uranium. Aluminum
20 -- is sorbed by the SG- 1 carboxyl resin in the range pH =
= 2-5.0. With comparable concentrations of uranium
and aluminum in the solution the latter is sorbed by
2 3 4 5 6 Ay
SG- 1 resin in much smaller amounts than uranium.
Fig. 3. The dependence of the sorption capacity The uranium is able to extract from the resin the
of SG-1 resin for uranium and accompanying im- previously sorbed aluminum.
purities on the pH of the medium. Content of
impurity (g/liter): ?) UO2+ - 1.0 (uranium);
x) 0.05 Feat; 0) 2.5 Fe 2+; 0) 1.0 Al s+,
?) 0.02 Cu 2+; p) 5.0 Mn 2+.
All elements which are present in solution only
in the anionic form, are hardly sorbed at all by the SG-1
type carboxyl resins. The presence in these solutions of
noticeable amounts of phosphate, arsenate, fluorides, and
oxalates can lead to a considerable reduction in the sorption capacity for uranium of the SG-1 type resin. It should
be noted that during lime neutralization of production solutions of pulps these interfering impurities are precipitated
and hence do not affect the sorption capacity of the SG- 1 resin for uranium. As regards the effect of the anionic
composition of the solution on the sorbability of uranium by the SG-1 resin, it should be mentioned that the
presence in the solution of even 100-150 g/liter of the nitrate ion does not affect the sorbability of uranium.
From solutions containing 50-100 g/liter acetate, uranium is also well sorbed by the SG-1 resin; with the equili-
brium concentration of uranium at 0.08-0.2 g/liter and the sorption capacity for uranium being 100-110 mg/g.
Figure 4 gives the isotherms for the sorption of uranium by SG-1 resin from various solutions. The copoly-
mers of SG71 are chemically stable and are not decomposed during prolonged action of solutions with high p
and y activity. The desorption of uranium from SG-1 resin is accomplished by dilute solutions (2-316 sulfuric,
nitric or hydrochloric acids).
Figure 5 gives a typical curve for the regeneration of SG-1 resin by a solution of sulfuric acid. In the
process of regeneration. 90-95% of the sorbed uranium is extracted in volumes comprising 0.8-1.0 of the volume
of the regenerated sorbent (the whole of the regeneration process finishes after the passage of four to five volumes),
which ensures minimum consumption of chemicals for the regeneration and a high degree of concentration.
The low cost of the SG-1 type resin, its high mechanical and chemical stability, and also its large sorption
capacity for uranium in its extraction from sulfate, nitrate,and chloride solutions and pulps (at pH = 2.8-3.8) have
ensured the widespread use of this type of cation-exchange material.
For the sorption of uranium from sulfate, phosphate, and carbonate solutions and pulps.various types of anion-
exchange materials are also used.
A statement has been made [3) on the suitability of only strongly basic anion-exchange materials such as
quaternary ammonium bases for the extraction of uranium from sulfate and carbonate media. We have shown
that in a number of cases, weakly basic anion-exchange materials can be successfully used to extract uranium
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from solutions of complex salt composition with a large excess acidity, and also with a high content of phosphates
and other interfering impurities, firmly linked to the strongly basic anion-exchange materials.
Depending on the salt composition of the solution and its excess acidity,it is possible to use weakly basic
anion-exchange materials such as AN- 2F, medium basic types EDE-1OP,or strongly basic quaternary ammonium
bases such as AM and AMP.
5 10 15 20 25
Volume of regenerated material per
volume of resin
Fig. 4. Isotherms for the sorption of uranium by SG-1
resin from solutions: 0) sulfate; 0) sodium acetate
(UO2 : CH3OOO - = 1: 10.8); x) sulfate with 50 g /liter
sodium sulfate.
2 6
Volume of regenerated material per
volume of resin
Fig. 5. Curve for regeneration of the SG-1
resin by a sulfuric acid solution: ?) sulfuric
acid; 0) uranium.
The anion-exchange materials AM-2F, a product of the- polycondensation of methylol derivatives of phenol
and polyethylenepolyamines, can be obtained in the form of beads by the method of pearl polycondensation.. It
has high selectivity and sorption capacity for uranium in the absorption of uranium from solutions with a high ex-
cess acidity. The anion-exchange material EDE-10P is the product of the polycondensation of epichlorhydrin and
polyethylenepolyamine. It has high sorption capacity for uranium in the extraction of uranium from phosphate
solutions. The anion-exchange materials which we prepared are polymerization resins of the quaternary ammonium
base type, containing a trialkylamine, pyridine or its derivative. The anion-exchange material AM is a product
similar to amberlite IRA-400 or Dowex-1. The AM and AMP materials can be used in a number of cases for
sorption from acid solutions with a small excess acidity (pH = 1.0-2.0) and in all cases for sorption from carbonate
solutions and ore pulps.
Figure 6 shows the dependence of the sorption capacity for uranium of some anion-exchange materials on
the excess acidity. With a sodium sulfate concentration of 50 g/liter, sulfuric acid 30-40 g/liter,and an equili-
brium concentration of uranium of 1 g/liter the sorption capacity for uranium of AN- 2F is twice that of the
strongly basic AM.
The AN- 2F type materials can be successfully used to extract uranium from colored solutions containing
20-50 g/liter of free sulfuric acid. With increase in the free acidity, there is a sharp drop in the absorption of
trivalent iron; hence the selectivity for uranium increases. Since the sorption complex of uranium is held much
more strongly than the complex of trivalent iron, then to obtain uranium concentrates of high quality before the
desorption of uranium,the AN- 2F material should be washed with a 1-2% solution of sulfuric acid; the uranium
is hardly desorbed and the iron is almost completely removed.
Figure 7 shows the dependence of the sorption capacity for uranium and iron of AN- 2F on the excess
acidity. With increase in the excess acidity from 10-30 g/liter,the sorption capacity for uranium changes very
little, whereas the capacity for iron falls from 60 to 10 mg/g.
o
iC
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Another example of the
materials for the extraction
300
04 3 2 I
its sorption from solutions containing large amounts of phosphate ions.
efficient use of weakly basic (AN-2F) and medium basic (EDE-10P) anion-exchange
pH 0,1 0,3 0,5 0,7
Concentration of H2SO4, mole/liter
Fig. 6. The effect of excess acidity on the sorption
capacity for uranium of the anion-exchange materials
AN-2F (x); EDE-1OP (c); AM (A) and amberlite
IRA-410 (O) (concentration of uranium 0.5 g/liter).
/0 20 30 40
Concentration of H2S04, g/liter
Fig. 7. The effect of excess acidity on the sorption
capacity for uranium (?) and iron (o) of the anion-
exchange material AN- 2F.
sorption of uranium (concentration 0.74 g/liter) from
solutions containing vanadium (concentration 0.42
These solutions where the concentration of phosphoric
acid reaches 150-350 g/liter are usually obtained in
the treatment of uranium-containing phosphorites.
Figure 8 shows the relationship between the
sorption capacity for uranium of various anion-exchange
materials and the concentration of phosphoric acid in
the solution.
In all cases the sorption capacity for uranium
of EDE-10P is more than twice the capacity of strongly
basic materials such as AM (amberlite IRA-400).
Molybdates and vanadates have a very noticeable
depressing action. This difference is especially
noticeable when comparing the capacity of AN-2F
and AM. The vanadate ion has hardly any effect on
the sorption capacity of uranium on weakly basic
AN- 2F type materials, whereas the capacity of the
strongly basic material AM under these conditions is
reduced by more than a half. For example, in the
50 /00 /50 200
Concentration of P205, g/liter
Fig. 8. The effect of concentration of phosphoric
acid on the sorption capacity for uranium for various
anion-exchange materials: O) EDE-1OP; ?) WO tit
L-150; &) AMP; x ) AN-2F; ^) AM (con-
centration of uranium 1 g/liter, H2SO4 10 g/liter).
g/iiter), the sorption capacity for uranium of AN-2F is reduced by 7%, and AM or amberlite IRA-400 by 50
and 5656, respectively. The selection of the ion-exchange material to treat a given type of uranium ore should
be made after determining the optimum conditions for leaching and after ascertaining the salt composition of
the solution.
The various types of absorbents for uranium differ considerably from one another in their kinetic character-
istics. For the SG-1 cation-exchange material,acceptable equilibrium is reached after 60-80 minutes, whereas
for the AN-2F anion-exchange materials 40-50 minutes are needed and for AMP .20-30 minutes. During de-
sorption the equilibrium of all absorbents for uranium is established after 15- 20 minutes.
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The types of ion-exchange materials differ considerably in their behavior during regeneration. As can be
seen from the data given in Fig. 9, 90-95% of uranium absorbed by the carboxyl resin SG-1 is desorbed by
0.8-1.0 volume of regenerating agent per volume of sorbent; and in the regeneration of AN-2F type anion-
exchange materials and AMP, two to three volumes per volume of resin, respectively. The regenerating mixtures
2 4 6 8 10
Volume of regenerated material per
volume of resin
Fig. 9. Regeneration curves: o) carboxyl
resin SG-1 (by a 3% solution of sulfuric acid);
e) AN- 2F (by a nitric mixture).
ao
40
0
30
0
20
V
Cd
a,
v 10
U 10 20 30 40
Concentration of anions, g/ liter
Fig. 10. The relationship between the sorption
capacity for uranium of ED>4-1OP and the con-
centration of carbonates (x ), bicarbonates of
sodium (e) and ammonium (o).
used for the anion-exchange materials are nitrates or chlorides of sodium with the addition of the corresponding
acids. The addition of small amounts of complex-forming materials somewhat improves the regeneration process
and reduces the consumption of regenerating solution.
For the sorption of uranium from carbonate and bicarbonate solutions and pulps, as already mentioned, AM,
AMP, and EDE-1OP type resins are used. The sorption capacity of these anion-exchange materials, other con-
ditions being equal, depends on the concentration of carbonate and bicarbonate.
Figure 10 shows the dependence of the sorption capacity for uranium of EDE- lOP on the concentration of
carbonates and bicarbonates of sodium and ammonium. If the carbonate solution obtained during the leaching
of uranium ore is carbonated to change the carbonates to bicarbonates, the sorption capacity for uranium of the
EDE-10P type anion-exchange material increases.
Figures 11 and 12 show the dependence of the sorption capacity for uranium of EDE-lOP and AM materials
during absorption from a carbonate solution on the concentration of depressing ions (initial concentration 1 g/liter,
excess carbonate 5 g/liter, grain size + 0.3 to - 0.5 mm). It can be seen from the figures that the sorption of
uranium is most reduced by chloride ions. In a 0.5 N solution of nitric ions,the sorption capacity for uranium of
AM anion-exchange material is reduced by a quarter,and with the same concentration of chloride ions it is re-
duced to less than a fifth. The regeneration of anion-exchange materials after sorption of uranium from soda
solutions and pulps should therefore be carried out by dilute solutions of soda with sodium chloride, which con-
siderably improves this process.
The table gives a comparative evaluation of anion-exchange materials made in various countries.
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0,3 46 40 1,5 2,0
Concentration of depressing ions,
g- eq/liter
Fig. 11. Dependence of the sorption capacity for
uranium of EDE-10P on the concentration of de-
pressing ions: m -POr ; O-NOj ; X-SOq-; A -CF
O,3 0,6 1,0 45 2,0
Concentration of depressing ions,
g ? eq/liter
Fig. 12. Dependence of the sorption capacity
for uranium of the anion-exchange material
AM on the concentration of depressing ions:
0-F-; O-NO3; a-PO 4-; X-SO4 ; O-CI-?
The Sorption of Uranium from Clarified Solutions
In the treatments of uranium ores,clarified solutions can be obtained by percolation, filtration .or counter-
current decantation. The presence of small amounts of suspended particles does not hinder the process of sorption
in a stationary layer. In some cases,to remove the suspended particles control filtration is used. The dynamics
of sorption of uranium by various types of cation-exchange and anion-exchange materials can be described with
sufficient accuracy for practical purposes by the equation of [4], used in 1929 [5] to describe the dynamics of
sorption from solutions of activated carbon:
W = K (L - h). (1)
In the sorption of uranium by a stationary layer of ion-exchange material after a certain volume of solution
passes through the layer of sorbent.stationary transfer of the front of depletion of the ion-exchange material is
established. The length of the ion-exchange layer lq, at which the concentration of the absorbed ion changes
from the initial value to c = 0 during the steady process, is called the length of the working layer. The stationary
transfer of the front of depletion of the ion-exchange material will only occur on the condition that the length of
the ion-exchange layer in the sorption column L is much greater than the length of the working layer Lo. In
accordance with Eq. (1) the relationship of the volume W of the filtered solution of uranium on L (with the con-
dition L >> Lo) is expressed by a linear equation. The angular coefficient characterizing the slope of the straight
line is equal to the coefficient of filtration action k , determining the rate of movement of the depletion front
per unit of length of the ion-exchange layer. The section h cutting the straight line at one of the axes, is equal
to the loss in filtration action and is directly connected with the kinetic characteristics of the ion-exchange
material. In a general form the volume of the filtered solution of uranium depends on L, the static sorption
capacity a, the concentration of uranium in the initial solution c 0, the rate of filtration v , the salt composition
of the solution,and the pH. The theory of sorption dynamics was further developed in [6-8]. The equation of
the 'output curve' can be obtained from the equation of balance of the sorbed material and has the form
PLO
c=coe
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where c is the concentration of the material behind the layer of ion-exchange material after passing through
the filter; S is a kinetic coefficient directly proportional to the diffusion coefficient (D) of
the sorbed material inside the grain of the ion-exchange material and inversely proportional to the grain diameter
(d n). The length of the working layer can be determined from (2)
n
L0 In~~=eDInC
Sorption Capacity of Various Anion-Exchange Materials for Uranium
comp
ositions
sol
utions
`O'
CO
Active ionogenic
0
[d
ff,
Cd 'td
Q
il
O
Cd Cd Nyy
Anion-exchange material
Country
groups
z
Cd
Cd x
r
4 zz x
r-a
CD
O
~+ )
t)
H
4) a)
to 0) F
b9x
bo
~
~0
a'
C> 'I
e-1 tf)
boO
,-i CO
0
7
b~O
.-i MU .-1 CO
AN-2F ? ? ? ? ? ? ? ?
USSR
-NH2; =NH; = N
84
64
32
24
20
91)t-10P ? ? . ? ? ? ? . ?
USSR
= NH; = N; N
80
70
70
60
60
AMP . ? ? ? . ? ? ? ? ? ? ?
USSR
-N
86
82
20
10
10
x
N
AM ? . . . . . . . . . . . . .
USSR
-
(CH3)3
73
64
13
8
8
x
N
W ofatit .. . . . . . . . .
E. German
; = N; = NH
=
96
70
67
60 -
55
Wofatit . . . . . . . . .
E. Germany
-N(C2H5)3
68
40
8
4
5
x
N
Amberlite . . . . . . .
USA
(CH3)3
-
82
72
20
16
18
x
Dowex 1 -X8 . ? . ? ? . .
USA
- i (CH3)3
88
80
20
15
15
x
+ C2H4OH
Dowex 2-X8 ? ? ? ? ? ? ? ?
USA
- N \\
82
76
20
14
15
I (CH3)2
x
N
Deacidite FF . . . . . . .
Britain
(CH3)3
-
88
80
20
7
20
x
From the shape of the output curve (sloping steeply or gently), it is therefore possible to obtain an idea
of the kinetic characteristics of the ion-exchange material or, by determining the value of 8, to determine Lo
from Eq. (3). Equations (1) to (3) can be used in the analysis of data on the dynamics of sorption of uranium
from acid and carbonate solutions on carboxyl resin and anion-exchange materials and in calculations of
sorption apparatuses. '
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In accordance with Eqs. (2) and (3) in the sorption of uranium from clarified solutions by a stationary layer
of ion-exchange material,it is essential to use the smallest possible grain size of the ion- exchange4material. It
has been found that during the sorption of uranium from solutions in large industrial columns,the ion-exchange
material grain size should be between + 0.2 and - 0.8 mm. The optimum rate of flow of the solution during
passage through a layer of ion-exchange material changes somewhat depending on the salt composition of the
solution, the concentration of uranium,and the type of ion-exchange material and is between 6 and 12 m /hr.
For the maximum use of the sorption capacity for uranium of the ion-exchange material,the concentration
of uranium behind the layer of ion-exchange material before the column is disconnected for regeneration should
be equal to the initial concentration in the solution, i.e., c/c 0 = 1. To ensure a closed sorption- desorption cycle
(providing continuous treatment of the solutions) with a minimum number of sorption columns, L should be not
less than 1.5-2.5 m. The number of columns used in the sorption- desorption cycle can also vary, depending on
the salt composition and the excess acidity of the solution. Depending on the output of the unit,sorption columns
of 30-35 m2 cross section can be used. The use of small cross section sorption columns (3-5 m2) in installations
with high productivity is economically undesirable, since apart from the increase in capital expenses, there is an
increase in the operating costs and the automation of the whole production process. The desorption of uranium
for various types of ion-exchange materials is carried out at a flow rate of 3-8 m/hr. With increase in L there is
a decrease in the volume of the desorbed solution. The latter is divided into three or four fractions; the main
fraction, containing 90-95% sorbed uranium, is taken out for further treatment, the remaining two or three
fractions are returned and used for the desorption of uranium from the next column. The main impurity in the
concentrates when using SG-1 resin is aluminum, the content of which can reach 10-12%, and when using the
anion-exchange materials AN-2F, AM, and AMP it is iron, the content of which reaches 5-12%. To obtain high-
purity uranium compounds it is possible to use sorption or extraction purification of commercial fractions of
regeneration materials using tributyl phosphate (TBP) or trioptyl amine (TOA).
The Sorption Extraction of Uranium from Ore Pulps
In 1953 the author proposed a filtration less method for treating uranium ores, consisting of the suspension
of precipitate or finely ground material in which the uranium was distributed between solid and liquid phases.
A granular sorbent was added which selectively absorbed the product to be extracted. After absorption of uranium
by the sorbent from the solid and liquid phases of the pulpit was separated by means of a mesh or in a rising
stream. Almost at the same time and independently,the method of sorption of uranium from pulps by strongly
basic anion-exchange materials was developed by American scientists [9].
We have now tested several variants of sorption. filtration- less systems for processing uranium ores, differing
in the arrangement of the apparatus. The main regularities of sorption of uranium by ion-exchange materials
from solutions are fully preserved for processes of sorption from pulps containing up to 40-50% of solid material.
Sometimes a difference is observed in the kinetic characteristics due to the fact that the attainment of equili-
brium in the system depends not only on the concentration of uranium in the solution,but also on the kinetics of
its desorption from the solid phase of the pulp. Uranium which has been put into solution during leaching is
quantitatively extracted from all types of ore pulps; in a number of cases additional extraction is observed
during the sorption treatment of pulp. The treatment of pulp by a resin which selectively absorbs uranium is
similar to very efficient repulping.
To achieve maximum efficiency in the sorption process during the preparation of the ore pulp, it is essential
to ensure the maximum possible concentration of uranium in it, minimum excess acidity or carbonate content and
minimum amount of salts which are also dissolved during leaching.
The process of sorption from pulps can be carried out in reactors with an impeller stirrer, Pachuca tanks,
or in special sorption apparatuses.
The variants of the method for sorption of uranium from ore pulps can be put into two main types: the
static method and the dynamic method of sorption of uranium from pulps in fluidized bed of sorbent.
The sorption of uranium from ore pulps by the static method is carried out in ordinary reactors. Pachuca
tanks,or in special sorption apparatuses. The resin and pulp are separated on screens of various designs. At each
part of the sorption a stage is reached which is close to equilibrium (by analogy with the static sorption experi-
ment of this type the arrangements are called static). The larger the stages of sorption, the larger the value of
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the sorption capacity, since the equilibrium concentration approaches the initial value. However, the maximum
permissible number of sorption stages is limited by economic considerations. Continuous and batch sorption
processes for treating uranium ores have been established.
I
Crushing and grind-
ing
I
Acid leaching I
1
Separation of sand
sludges
41
Treatment with
lime to pH = 2.8-
3.2
Sorption from pulp (15-201o of soli
material) on SG-1, AN-2F, AM,
AMP resins
Treatment with
lime to pH=8-9
To the dam
Desorption solution
(regenerating
material)
Extraction purification
of TBP, TOA
Fig. 13. Flow sheet for the treatment of ore pulps in a fluidized bed of sorbent.
Reextraction with
ammonium carbon-
U8O8
The sorption of uranium from a pulp is carried out according to a continuous countercurrent 3- or 4- stage
system. With the dynamic method of sorption of uranium from pulps in a fluidized bed of sorbent (Fig. 13), the
preparation of the ore pulp (grinding, leaching and classification) is carried out as in the case of other sorption
systems. The pulp,after careful separation of sands and control screening is fed to the sorption columns from the
bottom to the top at a rate which ensures the formation of a fluidized bed of sorbent. The uranium is sorbed
during passage of the pulp through a fluidized bed of sorbent. The coefficient of expansion depends on the
density and viscosity of the ore pulp and also on the speed of the arising' stream. When using the process of
sorption from pulp in a rising stream,it is very important to have an ion-exchange material with a high specific
gravity. In production specimens of ion-exchange materials the main fraction held is that with a specific
gravity of 1.20-1.25 g/cm3. Ion-exchange materials of higher specific gravity are more difficult to obtain.
With the method of sorption in a fluidized bed of sorbent it is possible to process ore pulps containing 10-20%
of solid material.
Further improvement in the sorption methods for extracting uranium can be achieved by introducing the
following measures:
1) reducing the consumption of resin in the continuous arrangements for sorption of uranium from ore pumps;
2) intensifying processes of sorption and desorption due to the use of ion-exchange materials with high
kinetic properties;
3) developing sorbents with high selectivity and sorption capacity for uranium;
4) reducing the consumption of chemicals when using new methods of desorption and separating pure
compounds of uranium;
5) combining processes of leaching and sorption of uranium from pulps.
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1. D. I. Ryabchikov and M. M. Senyavin, Material of the International Conference on.the Peaceful Use of
Atomic Energy (Geneva, 1955), Vol. 8 [in Russian] (Moscow, Metallurgy Press, 1958) p. 328.
2. C. Diffrich, Phys. Chem.. 29, 499 (1899).
3. Preuss and Kunin, Material of the International Conference on the Peaceful Use of Atomic Energy (Geneva,
1955), Vol. 8 [in Russian) (Moscow, Metallurgy Press, 1958) p. 60.
4. N A. Shilov, L. K Lepin', and S. A. Voznesenskii, Zhur. Russ Khim. Obshchestva 51, 1107 (1929).
5. S. A. Voznesenskii, Physicochemical Processes in the Purification of Water [in Russian] (Moscow, State
Construction Press, 1934).
6. M. M. Dubinin, Physicochemical Fundamentals of Sorption Technology [in Russian] (Moscow, State
Chemical and Technical Press, 1932).
7. A. A. Zhukhovitskii, A. N. Tikhonov, and Ya L. Zabezhinskii, Zhur. Fiz. Khim. 19, 253 (1945).
8. N. N. Tunitskii and E. P. Cherneva, Zhur. Fiz. Khim. 24, 11, 1350 (1950).
9. Hollis and McArthur, Material of the International Conference on the Peaceful Use of Atomic Energy
(Geneva, 1955), Vol. 8 [in Russian) (Moscow, Metallurgy Press, 1958) p. 71.
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LETTERS TO THE EDITOR
ENERGY DEPENDENCE OF THE DIFFERENTIAL CROSS SECTIONS
AND MECHANISM OF THE (d, p) REACTION
V. B. Belyaev, B. N. Zakhar'ev, and V. G. Neubachin
Translated from Atomnaya Energiya, Vol. 9, No. 4, pp. 298-299, October, 1960
Original article submitted February 10, 1960
In a number of articles [1], the ratios of the reduced widths of the y 2 levels obtained from the (d, p)
reaction data are used as spectroscopic characteristics in discussions of nuclear models. The calculation of y 2
from the differential cross sections of the (d, p) reaction is carried out by means of the Born approximation
formula for the stripping mechanism [2, 3]
da _ 2J-I-l kp Mi M
dS2 2/ [ 1 kd M f j Mn Mi } Md 4nNa
1 1 _ 1 1
Z Z (a2+ff2 02+K2
I
(1e-Pale) 2 hz We yz
If the stripping mechanism were unique, the Born approximation with distorted waves [3] would be a good one,
since the deformation of the internal wave function of the deuteron is evidently not very great [4]. Meanwhile,
the experimental results connected with the investigation of forbidden stripping [5] indicate that in the (d, p)
reaction with a deuteron energy of the order of several million electron-volts,and knock-on of a proton by a
deuteron with the capture of the deuteron is an important process [the (d, p) reaction is of this type]. In those
cases in which ordinary stripping is allowed, the angular distribution of the reaction products can differ little
from the Butler distribution, since for a knock-on process it is characterized, as in the case of stripping, by a
maximum at small angles. The cross section of the knock-on effect, however, is described by a formula of a
type different from that in the case of stripping and with a different energy dependence [6, 7]. In the analysis
of the energy distribution of the (d, p) reaction cross section, it is convenient to use the Born approximation
formula for stripping [2] and to investigate the dependence of the quantity y2 appearing in this formula in the
form of a multiplier of the energy. If the stripping mechanism were basic, then the dependence of y 2 on Ed
-obtained by means of the Butler formula would have the form of a monotonic, smoothly increasing curve tending
to a limit for Ed greater than the barrier height for deuterons. The height of the barrier depends, of course,. on
the charge of the nucleus and the orbital angular momentum of the captured neutron. ` In the presence of the
competitive processes indicated above, the energy dependence will be different and will not be connected so
simply with the reduced widths.
For the analysis of the experimental data, we have chosen cases with a sufficiently large deuteron energy
(Ed > 4 Mev) , in order to decrease the role of the, various corrections to the Butler formula, for example, Coulomb
corrections.
The results of experiments for various reactions are presented in the table.
The effect of the distortion of the proton and deuteron waves due to the nuclear interaction between the proton
and the deuteron can be neglected if Bowcock's method is used.
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The dependence of the relative reduced widths y 2 on Ed are of the order 5056, and the errors for the relative
values are much less (about 10%).
As can be seen, both the relative and absolute values of the reduced widths, in a number of cases, vary
with Ed. The change in y 2 with Ed was noted previously [18], but it was not analyzed, owing to insufficient
data.
Experimental Results for Different Reactions
f
l
Reduced widths *
o
Leve
final nucleus
r- >,
~
Transition
Ed, Mev
Reaction
ev
Mev
'
characteristic
8 [2]
8,9 [91
9 [10]
14,8 [11]
19[12]
Q) Cd
ground state
1/2-
p8 > p9
2,2
1,3
1,2
1,9
0,9
3,09
1/2'
p8 -' P6s1/2
:3,0
3,8
6,5
8,3
1,6
C'2 (d, p) C'3
3,684
3/2-
p8 =-' p9
0,16
0,48
0,38
0,28
-
3,855
5/2'
p8 > p8d5/2
3,5
2,6
1,4
5,7
1,1
3,6 [2]
9 [101
14 [13]
s to
3
ground state
0'
P5 > p8
1,2
22 (relative)
0 , 093 * *
-
-
Be (d, p) 1
e
3,37
2'
p5 ' p6
0,19
8 (relative)
0, 012 *
4,11 [14]
7,7 [2]
7,73 [15]
9 [10] 19,1 [15]
11
ground state
5/2'
p12 , p12d5/
0,66
1,0
2,5
3( relative)
1,8
016(d p) 0
0,875
1/2k
2
p12 - p12si/2
1,1
2,6
7,0
9( relative)
3,7
8 [2] 9 [10] 14,8 [17]
ground state 1/2- p1o -> pu 2,1 - 1,0 - -
N14 (d p) N15 6,33 3/2- plo ' p" - 0,23 0,18
8,32 1/2 P1a ' pij2 11,5 4,5 5,4
*All results except those denoted by (**) are in units of 10-19erg/ cm
** The results are in units of 3fi2/(2ma)
it is interesting that, in the examples considered, there are cases in which 1) the captured neutron is
outside a filled shell (capture of s and d neutrons by C12 and 016), and 2) the captured' nucleon is simply
added to the unfilled shell (in all cases of neutron capture considered above, 1 = 1). These cases have the
following differences: 1) in the first case, the remaining direct processes (proton knock- on by a deuteron and
stripping of a heavy particle) are difficult (in comparison with ordinary stripping); the cause for this lies not
only in the binding energy of the core nucleons, but also in the fact that when a deuteron is captured, one of
its nucleons should lie in the ip orbit and the other in the id or 2s orbit; 2) the ratio of the two reduced widths
in the Born approximation for processes in the second case will be practically constant, and for processes in the
first case (where the widths are compared for the capture of a neutron with different 1) it will change slightly
with the energy. We see that the data in the table apparently reflect both these circumstances. In fact, for the
case of the capture of p nucleons, the change in the reduced widths is sharper, which, perhaps, indicates the role
of other direct processes. It should, however, be noted that the data used in our work was obtained at different
times and by different authors; this reduces their accuracy. But in any case, analysis shows that, first, the use
of the reduced widths in Butler's formulas for quantitative spectroscopic purposes is not entirely reliable, and
second, it is of great interest to carry out a series of experiments in which the dependence of the absolute
differential cross sections on the energy is studied. It will be very important to increase the accuracy of the
experimental data, which will make it possible to obtain information on the mechanism of the (d, p) reaction,
supplementing the results of the investigations of forbidden stripping and stripping of heavy particles.
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1. T. Averbach and J. Freuch, Phys. Rev. 98, 1276 (1955); K. Standing, Phys. Rev. 101, 152 (1956);
B. Flauers, Proceedings of the All-Union Conference on Nuclear Reactions at Low and Medium Energies
[in Russian] (Moscow, Acad. Sci. USSR Press, 1958) p. 536.
2. S. Yoshida, M. Fujimoto, and H. Kikuchi, Progr. Theor. Phys. 11, 264 (1954).
3. W. Tobocman, Phys. Rev.. 94, 1655 (1954); W. Tobocman and M. Kalos, Phys. Rev. 97, 132 (1955).
4. W. Tobocman, Phys. Rev. 108, 74 (1957); K. A Ter-Martirosyan, Zhur. Eksp. Teor. Fiz. 29, 713 (1955).
5. N. Evans and W. Parkinson, Proc. Phys. Soc. 67A, 684 (1954).
6. A. French, Phys. Rev. 107, 1655 (1957); N. Evans and A. French, Phys. Rev. 109, 1272 (1958).
7. V. G. Neudachin, Zhur. Eksp. Teor. Fiz. 35, 1166 (1958).
8. J. Bowcock, Proc. Phys. Soc., 68, 512 (1955).
9. F. El Bedew, Proc. Phys. Soc. 69A, 3, 221 (1956).
10. T. Green and R. Middleton, Proc. Phys. Soc. 69A, 1, 28 (1956).
11. R. Freemantle, W. Gibson, and J Rotblat, Phil. Mag. 45, 370, 1200 (1954).
12. J. McGruer, E. Warburton, and R. Bender, Phys. Rev. 100, 1, 235 (1955).
13. J. French and A. Fujii, Phys. Rev. 105, 2, 652 (1957).
14. E. Baumgartner and H. Fulbreit, Phys. Rev. 107, 1, 219 (1957).
15. W. Fairbairn, Proc. Phys. Soc. 67, 564 (1954).
16. E. Warburton and J. McGruer, Phys. Rev. 105, 2, 639 (1957).
17. G. Owen and L Madansky, Phys. Rev. 105, 1766 (1957).
18. T. Fulton and G. Owen, Phys. Rev. 108, 789 (1957).
HIGH ENERGY y-RAY BEAMS
V. S. Barashenkov and Hsien Ting-ch'ang
Translated from Atomnaya Energiya, Vol. 9, No.- 4, pp. 300-301, October, 1960
Original article submitted April 4, 1960
It is known that experiments with high-energy y quanta are of great interest for the investigation of the
structure of elementary particles and. for verifying electrodynamics at small distances [1].
In order to obtain hard y quanta, one ordinarily uses electron accelerators. The maximum energy of the
y quanta obtained in this way is 1.06 Bev (the Stanford linear accelerator). Intense beams of still harder y
quanta can be obtained from the large proton accelerators: the joint Institute of Nuclear Studies' proton syn-
chrotron at Dubna and the recently finished accelerator in Geneva. The main source of y quanta in these
accelerators is the decay of 7r ? mesons generated in collisions between the accelerated protons and nucleons of
a target.
Statistical calculations show that the number of y quanta produced in a single nucleon-nucleon collision
is, on the average, equal to the number of charged 7r mesons produced in the collision. From Fig. 1, in which
the results of these calculations are shown, it is seen that the intensity of y - quanta beams can be very great.
If the 7r *-meson distribution is described by the function W7r (p; 0 ), where .2 is the momentum and 8
is the angle of emission of the 7r? mesons produced in nucleon- nucleon collisions, then the corresponding
distribution of the y quanta is expressed by the function
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Pmax Pmax
l~Vv (k; 0) = 2 W., (P; 0) dp - 2 41'n (P; 0) dP
1~n2_A_[L2 P
where K = j/ k2 (1 } k~/k2) - ?2; k and 0 are the energy and angle of emission, respectively, of the y quanta
in the laboratory system; k 0 is the energy of the y quanta in the n ? meson rest system; ? is the 7r ? meson mass.
5 10 15 20 25 30E,Bev
Fig. 1. Mean number of y quanta produced in the Fig. 2. Energy spectrum of y quanta integrated
decay of 7r? mesons generated in a single nucleon- over all angles [ WY WY (k)J . At high energies
nucleon collision (n ). (E is the kinetic energy of the values of log [106 Wy (k)] are given (k is the
the accelerated protons in the laboratory system.) y -quanta energy in the laboratory system.)
This formula is applicable for k Z 0.5 Bev, but it can also be used for estimations at energies k =
0.2 Bev [2].
Since the experimental values of the function W7r (p; 0) for the joint Institute of Nuclear Studies' proton
synchrotron are not known at the present time, we calculated this function by means of the statistical theory of
multiple production of particles (accelerated-proton kinetic energy E = 10 Bev).
0 5 10 15 20 25 30
Fig. 3. Angular distribution of y quanta integrated
over all energies [Wy = WY (0 )]. (0 is the
scattering angle in the laboratory system.)
Fig. 4. Energy spectrum of y quanta for different
angles [Wy = WY (k; 0 A. (k is the y-quanta
energy in the laboratory system.)
Figures 2 and 3, respectively, show the calculated spectrum and angular distribution of the y quanta:
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n
Wv (k) = 21t WV (k; 0) sin 0 d0,
n
kmax
WY (0) = W (k; 0) dk.
0,2
Figure 4 shows the calculated spectrum of y quanta WY (k; 0) for various angles 0. (In all cases the angle 0
is the angle of emission of the y quanta with respect to the direction of the accelerated proton beam.)
For the function WY (k; 0), we carry out the normalization
n k max
2n sin 0 d0 WY (k; 0) dk=1.
0 0
As can be seen, the basic part of the y radiation is concentrated in the small-angle region. With an increase
in the y - quanta energy, the intensity of the radiation drops rapidly.
In conclusion, we consider it our duty to thank M. A. Markov, on whose initiative we carried out the
calculations, for discussion and advice, and V. I. Veksler for valuable comments in the discussion. of the results.
1. I. M. Zlatev and P. S. Isaev, Nuovo cimento 13, 1 (1959); Zhur. Eksp. Teoret. Fiz. 37, 1161 (1959);
V. A Petukhov, A. A. Komar, and M. 1. Yakimenko, Joint Institute of Nuclear Studies, Preprint R-283
(1959); D. I. Blokhintsev, V. S. Barashenkov, and B. M. Barbashov, Uspekhi Fiz. Nauk 68, 417 (1959).
2. R. Sternheimer, Phys. Rev. 99, 277 (1955).
The intensity of the y radiation also drops rapidly as k --> 0. This part of the spectra is not shown in
the figures.
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ON THE DOUBLE FOCUSING OF A BEAM
Translated from Atomnaya fnergiya, Vol. 9, No. 4, pp. 301-303, October, 1960
Original article submitted May 4, 1960
In a number of articles [1-3] devoted to the double focusing of a beam of charged particles by a sectional
magnetic field, it is assumed in the derivation of the horizontal focusing conditions that there is no leakage field
(the magnetic field at the edge of the magnet suddenly drops to zero from the homogeneous field intensity). In
derivation of the vertical focusing conditions, the leakage field region is considered to be so narrow that the
vertical coordinate and approach angle of the particle trajectory in it do not have time to change.
In the present work, the influence of the leakage field on the vertical focusing of charged particles is in-
vestigated,and a method of calculating a sectional magnet with double focusing with allowance for the leakage
field is presented.
The influence of the leakage field on the horizontal focusing of sectional magnets is considered in [4].
It is readily shown that the variation of the angle of approach of the particle trajectory, as the particle travels
through the leakage field, can be determined from the expression
ve
sine sinle--- 6 S I"d(y
e eo Ho 8
where eb is the angle between the normal to the edge of the magnet and the tangent to the trajectory at the
point of entry into the leakage field (the angle of approach at the beginning of a leakage field); e.is the
angle of approach at the end of the leakage field; p s is the radius of curvature of the particle trajectory in a
homogeneous field; Yb and ye are the coordinates of the trajectory at the beginning and end of the leakage
field, respectively; H 0 is the homogeneous field intensity; H is the field intensity at an arbitrary point of the
leakage field; 6 is the air gap between the poles; Y is the coordinate of the particle.
If the law according to which the field intensity drops at the edge of the magnet is known, then
the integral C if d (y) can be calculated. It is simpler, however, to replace the decreasing field by a
Ho S
11b
homogeneous one with intensity H U, The width of the homogeneous field b /6 is chosen so that
Ve
S ho d [ S ] = g . The boundary of the resulting field is then at a distance r / 6 from the edges of the
vb
pole-faces (Fig. 1). The axis of ordinates on this figure coincides with the boundary of the pole. In practice,
K / 6 is determined as the difference between the areas F, and F2, and, for the usual magnet designs, it is equal
to 0.6-0.7.
In the region of the leakage field, the particles move along a complicated curve. As a result, the real
trajectory is displaced with respect to the calculated one by a distance a. In general form
a = Yet g e e+ eo (cos Eb?5 se )
v
sin Eo+ dy
Q
0
V
-(sinEO+ dy
\ el
0
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The quantity a was calculated by a graphical-analytical method for a given range of values of 6 / p
and approach angles of the trajectories Ee (or Eb). The results of the calculations are shown in Fig. 2. From
this figure it is seen that for large 6 /p0 and Ee,the displacement essentially depends on Ee, which leads
to the increasing of the focal distance, and to additional spreading of the focus. For small 6 /pU and Ee,
this effect only increases the focal distance. The calculations show that for p /6 o s 0.1 and Ee < 10
the increase in the focal distance does not exceed 110.
5
0 5 10 15 20 25 30
Fig. 1. Variation of the magnetic-field intensity
at the edge of the magnet.
-10
Fig. 3. Dependence of the average approach
angle on the initial approach angle for
various values of the parameter 6 / p o:
1) 0; 2) 0.02; 3) 0.08; 4) 0.14;
5) 0.22; 6) 0.32.
. 6 e, deg
Fig. 2. Dependence of the size of
the "drift" on the approach angle
for different values of the parameter
6/pa: 1) 0.32; 2) 0.22; -3) 0.14;
4) 0.08; 5) 0.04.
1 ~
J 52=-E,-?2
Fig. 4. Horizontal focusing by a sectional magnetic
field.
Influence of the leakage field on the vertical
focusing. As shown by the calculations, the con-
ditions of vertical focusing given in [1] and [21 are
valid only in the case of a long-focus system (6 /l . 0.15 and a large angle of beam divergence is calculated,
then one should introduce corrections for the displacement of the trajectories and the increase in the focal distance
due to this displacement. The calculation of the system of parameters should then be carried out, according to
expressions (4) and (5), with this new value of the focal distance.
In conclusion, the author expresses his gratitude to E. S. Mironov for suggesting the topic of this work, for
his constant interest, and valuable advice.
1. M. Camac, Rev. Sci. Intr. 22, 197 (1951).
2. W. Cross, Rev. Sci. Instr. 22, 717 (1951).
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 Declassified and Approved For Release 2013/02/14: CIA-RDP10-02196R000100060004-4
3. H. Hintenberger, Rev. Sci. Intr. 20, 743 (1949).
4. Experimental Nuclear Physics, edited by E. Segre [Russian translation] (Moscow, IL, 1955) Vol. I,
p. 513.
BEAM LOSS AT THE LIMITING RADIUS IN A PHASOTRON
V. P. Dmitrievskii, B. I. Zamolodchikov, and V. V. Kol'ga
Translated from Atomnaya Energiya, Vol. 9, No. 4, pp. 303-305, October, 1960
Original article submitted March 28, 1960
The resonance interaction of the radial and vertical oscillations close to n = 0.25 is considered. It has
been shown that this resonance is considerably more harmful than the parametric excitation of the vertical
oscillation, which is caused by the first harmonic in the magnetic field structure.
In actual high-energy phasotrons (400-700 Mev), the limiting energy to which the particles are accelerated
corresponds to the radius at which the exponent of the magnetic-field fall-off lies in the limits 0.25" >
0,25>n>0,2 n- - H dr / The coupled oscillations in the zone n = 0.2 cannot lead to the complete loss of
the beam [1]. The region of parametric excitation of vertical oscillations at a frequency Qz = 0.5 (n = 0.25)
penetrates directly into this zone. In actual phasotrons, parametric excitation cannot cause an essential increase
in the amplitude, since the width of the resonance zone for a first harmonic of el '- 0.001 is usually less than,
100 revolutions [2].
We shall consider the effects which.may be evoked by an increase in the vertical oscillation amplitude.in
the presence of an azimuthal inhomogeneity of the magnetic field.
In this case, the betatron oscillations (r, z, (p) are described by the equations
.2 f
z' +[n+eln cos (w+(p0)] z- r (d-n) ze--zI 0i
R dH
where n = - H (R) dr
e t'+(1-n)t+ 1_R+d e2 2R 2R(2d-n)z2-} 2Rz;2=.-ejRcos(cp+To), UI
2 z
d= R2 d2 H . e =r-R; Hz =H (r) [14- sl cos (p-{- qp)] (the prime denotes
r=R' Z H (R) dr 2 r-R'
differentiation with respect to the azimuthal angle).
We shall consider the coupled oscillations which result from the distortion of the closed orbit of the
azimuthal magnetic field with an inhomogeneity in the structure. The equation for the vertical oscillations in
this case can be represented in the form
z~'-~- {n+nz~ld2 n)+ tln
2(d-n)1 1
2 J cos ((P+(po) J z
n =0.
+
* In the system (1), only the resonance terms containing el are retained.
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y I `~ z n l >> I for n = 0.25, i.e., the beam loss at the limiting
For all actual phasotrons, the quantity n
radii due to the coupling of the oscillations and not from the parametric excitation of the vertical oscillations
of the first magnetic field harmonic. From Eq. (2), it follows that the oscillation amplitude is 2.7 times as great
for v revolutions, where
n.
V= .
L:tel l d-n i
Taking into account the fact that d =- r dn. we readily obtain the radial width of the resonance zone:
Z dr ,
Ar=Ei R.
n
If the average increase in' energy in the accelerator is equal to eV, then the number of revolutions for the width
of the resonance zone is
1--n..If,k2Eo-+ -Isk
ve -- - ,
n nl L0-j-1;k
(5)
where Ek is the kinetic energy of the field. Thus, under conditions of quasi-static operation, one can estimate
from (3) and (5) the increase in the amplitude of an ion in the resonance zone:
amax 1--n E k2Eo?E k
lit ao =2~Fid n2 el' Eo?Ek
Using the method of averaging [31, we can estimate the increase in the amplitude for dynamical operation:
In amax- 41'r 3-1 y1
where X is the rate of change of the characteristic frequency in the resonance zone n = 0.25. This rate is deter-
mined from the expression
4d eV Eo-j- E j
X-3nTk 2Eo+Gk
The results were checked on an EMU-8 electronic simulator. The simulator made it possible to solve the
problem of the dynamic passage of particles through the resonance of the coupling with forced radial oscillations
for n = 0.25. Equation (2), which could be represented in the system of two equations
s"-'n ((p) z-f 2 1d Uz=O, (9)
U"?U =0,
where n (q)=no -1-Xc; U ((p0)=1; U' (q)o)=0; z (?o)=no; z' (qo)=0. was integrated on the EMU-8.
Since the time was used as the variable in the solution of the system (9) on the electronic simulator, it
was possible to make an arbitrary choice of the initial phase go.
In the integration, cpo was chosen in such a way so as to obtain the maximum increase in amplitude of
the vertical oscillations in the process of the particle passing through the resonance of the coupling for no = 0.25.
The maximum oscillation amplitude was observed directly on the screen of the indicator and was deter-
mined from the voltage at the output of the integrator. The accuracy of the solution is 1 %.
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a
In max
a0
2,0
Fig. 1. Variation of the maximum amplitude with Fig. 2. Variation of the maximum amplitude
the size of the first harmonic of the magnetic field with the size of the first harmonic of the magnetic
for a phasotron with Ek = 680 Mev. field for a phasotron with Ek = 57 Mev.
Fig. 3. Process of passing through the resonance zone for a phasotron with Ek = 57 Mev for E1 =
=4.10-4.
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The system of equations (9) was solved for the following values of the parameters corresponding to two real
phasotrons (the phasotron of the joint Institute of Nuclear Studies at Dubna and the phasotron of the Institute of
Nuclear Research in Tokyo):
Ek=680 Mev;
V=10 kv
d=-10;
1,k=57 Mev;
V=7 kv
d7,5.
The obtained variations of in as ax = f (el), after passage through the resonance of the coupling
0
for n = 0.25 are shown in Figs. 1 and 2 (dotted curves). Shown in the same figures are curves calculated from
formula (6) by means of the quasi-static method (curve a) and from formula (7) by the method of averaging
(curve b).
Fig. 4. Process of passing through the resonant zone for a phasotron with Ek = 57 Mev for e1 =
=3.10-4.
From the theoretical consideration of the excitation of oscillations resulting from the coupling with forced
oscillations in the zone of nonlinear resonances of higher order, it is seen that for mirror-symmetry of the
magnetic field structures the resonance Qz = 1/3 does not appear, and resonances beginning with Qz =1/4 are
not harmful. The resonance excitation of the vertical oscillations caused by the second and higher harmonics
in the magnetic-field structure are, also practically not harmful.
The investigations of the process of the passage of ions through the resonance of the oscillation coupling
in the zone n _ 0.25 shows that this resonance determines the limiting radius to which ions are accelerated in
large phasotrons. In phasotrons of small radius (R < 100 cm), the passage of particles through the zone of this
resonance is possible, since the rate of:change of the characteristic frequency of the oscillations close to n
0.25 in such phasotrons is considerably higher.
The authors express their gratitude to V. P. Dzhelepov for valuable comments during the discussion of the
results of this work.
LITERATURE CITED
1. D. Hamilton and H. Lipkin, Rev. Sci. Instr. 22, 783 (1951).
2. L. Henrich, D. Sewell, and J. Vale, Rev. Sci Instr. 20, 887 (1949).
3. N. N. Bogolyubov and Yu. A. Mitropol'skii, Asymptotic Methods in the Theory of Nonlinear Oscillations
[in Russian] (Moscow, Physical and Mathematical Press, 1958).
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COMPARISON OF CASCADE CIRCUITS FOR PRODUCING
LARGE CURRENTS WITH LITTLE RIPPLE
A. A. Vorob'ev and S. F. Pokrovskii
Translated from Atomnaya Energiya, Vol. 9, No. 4, pp. 305-308, October, 1960
Original letter submitted March 4, 1960
In order to carry out accurate measurements on large targets in the 2-3 Mev energy region, it is necessary
that the particle energy oscillate no more than 0.1-0.5% for a beam current of the order of 5-10 ma. It is
interesting to consider the possibility of using for this purpose an electrostatic accelerator employing a cascade
generator with a power up to 30 kw.
The operation of cascade generator circuits under various conditions has been considered by the authors of
[1-3], who proposed various computational formulas for describing the action of these circuits under identical
conditions. In the present work, we compare the results of calculations of the operating conditions of cascade
generator circuits with various theories and confront them with the experimental data.
Type of circuit
Cockcroft-Walton
with a de-
creasing stage
symmetrical
three ..phase
capacity
Characteristics
after
after
_
after
after
after
after
Vorob'ev and
Bouwers
Vorob'ev and
Novikovskii
Pokrovskii
Pokrovskii
Melikhov
Melikhov
(n+i) n
n iL
X
12
Ripple voltage
(n+1)n iL
(n+1)n. iL
n iL
n il,
_
4 fC
rl,
n
4 IC
4 fC
2 T c-
fC
n! L
-
X
fC 6 X
_
t
4C I
X tl
iL 2+
n
Voltage drop
cj, (4n3 {
6
C
IL (8n3 }
12
C
fC
n
L (2n3-{
12
C
i L (2n3 {
cL (4n3
8
C
1
/
f
121C
1
1
+3n2
-F9n2-Fn)
+3n2+4n)
+3n2?4n)
+3n2+2n)
Optimum num-
f
b
fC
v Umax
LL
LC
Umax
z
fC
-Umax
ZL
fC
'L
2V- Umax
I fC
2 Uma
LL
fC
3-Umax
`L
stages
er o
L
Note. iL is the mean load current; f is the source-voltage frequency; n is the number of stages; C is the
capacitance of one condenser; Umax is the voltage to which the capacitance of each stage is charged; t1 is
the time during which the condenser of one column charges the condenser of the second column.
Computational formulas for four cascade generator circuits recommended by various authors are listed in
Table 1.
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Figure 1 shows the variation of the output ripple voltage SU with the number of stages n for a cascade
generator with selenium rectifiers for a load current iL = 1 ma. The number of stages was varied from two to
seven for a supply frequency of 50 cps. The capacitance of each cascade was 1 p f.
In calculating the ripple for the three-phase circuit, we neglected the term niL 6C tj.
Figure 2 shows the variation of the voltage drop at the output DUmin with the number of stages n for a
load current iL = 2 ma for the same cascade generator.
From a comparison of the experimental and theoretical results, we conclude the following:
1. For the Cockraft-Walton circuit, better agreement between the experimental data and the calculations
is obtained with the use of Vorob'ev and Melikhov's formula.
2. For the symmetrical circuit, Novikovskii's formula for the calculation of the ripple voltage gives an
increased value. More accurate results are obtained from Pokrovskii's formula; for the calculation of the voltage
drop, the formulas of Novikovskii and Pokrovskii are recommended.
3. The calculation of the voltage drop.from Pokrovskii's formula for the three-phase circuit gives an
increased value.
Comparison of the Properties of Cascade Generators for Various Circuits
Type of circuit
Characteristics
Cockroft-
Walton
with decreasing
stage capacitances
Symmetrical
Pokrovskii
( )
Three-phase
kii
k
(Vorob'ev and
(Vorob'ev and
rovs
)
(Po
Melikhov)
Melikhov)
Ripple voltage, kv
4,0
0,313
0,154
0,25
Voltage drop, kv
267
15,7
69,3
89
Total capacitance pf
5,0
65,0
7,50
10,0
Maximum allowable average load
current I, rel. units
1
1
2
3
Optimum number of stages, n
31
31
62
55
Output voltage, kv
2233
2484,3
2430,7
2411
Voltage drop ratio
AU,
1
17
3,85
3
DU
Voltage drop for total capacitance
of 5?f, %
100
130
258
150
In Table 2, the properties of the four types of cascade generator circuits, determined from the recommended
formulas, are compared for iL = 5 ma, f = 2 kc, Umax = 50 kv, n 52, C = 0.1 11 f, 2nUmax = 2500 kv.
Comparing the data of Table 2, we conclude the following:
1. The Cockroft-Walton circuit is the simplest, but it has the largest ripple voltage and voltage drop;
the maximum allowable mean load current is not greater than the rectifier current.
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2. The circuit of the cascade generator with capacitances linearly decreasing with, the distance from the
voltage source has values of the ripple voltage and the voltage drop 1/13 and 1/17, respectively, of the values
for the Cockroft-Walton circuit. The circuit with increasing capacitances is not more complicated than the
Cockroft- Walton circuit, but the nonstandard stages of condensers require the construction of condensers with
calculated characteristics. The total capacity of the circuit with decreasing capacitances for the same number
of stages is thirteen times the total capacitance of the Cockroft- Walton circuit.
/
2
E
l:
3'
Fig. 1. Variation of the output ripple voltage
with the number of stages. Cockroft-Walton
circuit: 1) experiment [2]; 1') calculation
according to formula of Vorob'ev and
Melikhov. Circuit with decreasing stage
capacity: 2) calculated according to the
formula of Vorob'ev and Melikhov. Sym-
metrical circuit: 3) experiment [2]; 3')
calculated according to the formula of
Novikovskii. Three-phase circuit: 4)
experiment [2]; 4') calculated according
to the formula of Pokrovskii.
3. With the symmetrical circuit, the values of the
ripple voltage and voltage drop are 1/86 and almost '/4,
respectively, of those of the Cockroft-Walton circuit.
A shortcoming of the symmetrical circuit in comparison
with the Cockroft-Walton circuit is the fact that it
requires twice the number of rectifiers and one and a half
times the number of condensers,
Fig. 2. Dependence of the voltage drop
at the generator output on the number of
stages. Cockroft-Walton circuit: 1)
experiment [21; 1') calculated according
to Vorob'ev and Melikhov's formula; 1?)
calculated according to Bouwers' formula.
Circuit with decreasing stage capacitances;
2) calculated according to Vorob'ev and .
Melikhov's formula. Symmetrical circuits:
3) experiment [2]; 3') calculated accord-
ing to Novikovskii's formula; 3") cal-
culated according to Pokrovskii's formula.
Three-phase circuit: 4) experiment [2];
4') calculated according to ?Pokrovskii's
formula.
4. The three-phase circuit is more complex than the symmetrical circuit, and with a smaller voltage
stability has a greater number of condensers and rectifiers. The basic advantage of the three-phase circuit is
the possibility of obtaining a maximum average load current three times that of the Cockroft-Walton circuit.
1. A. A. Voiob'ev and V. S. Melikhov, Izvest. TPI 70, 2, 139 (1951).
2. B. S. Novikovskii, Atomnaya Energiya 4, 2, 175 (1958).?
3. A. Bouwers, Elektrische Hochstspannungen (Berlin, 1939).
? Original Russian pagination. See CB translation.
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 HYDRAULIC
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RESISTANCE TO THE FLOW OF A LIQUID
Subbotin, P. A. Ushakov, and B. N. Gabrianovich
Translated from Atomnaya Energiya, Vol. 9, No. 4, pp. 308-310, October, 1960
Original article submitted March 25, 1960
In an attempt to extend the ideas on the hydrodynamics of channels of complex shape, the authors in-
vestigated the hydraulic resistance to the flow of a liquid along staggered bundles of densely (s/d = 1.0) and
semidensely (s/d = 1.13) packed rods.
The bundles consisted of 7 rods and 12 spacers. The spacers ensured that the liquid flowed with a uniformly
distributed velocity past the ,bundle cross section. The experiment was carried out for water whose flow was
measured by graduated Venturi tubes. Mercury differential manometers and water pressure gauges were used to
measure the pressure drop. The water temperature was controlled by means of a thermocouple. The resistance
coefficients were calculated from the formula
2 Q&
where dh is the hydraulic diameter of the bundle, equal to the ratio of four times the cross section past which the
liquid flowed to the total wetted perimeter.
In Fig. 1, the results of the present work are compared with the results of the study of staggered bundles
obtained by various authors. Also shown are curves corresponding to the formulas for round tubes (Blazius'
formula [1] and Frenkel's formula [2]):
0,3164
1, = 0 s
-2 Ig L 3 7 d h+ \ Rc ) l .
Our experimental points for the densely packed bundle lie approximately 40% above curves 1 and 2, and
the results of the experiment on the bundle with s/d = 1.13 lie 10-15% above these curves. The experimental
data of [3] and [4] agree with one another and lie considerably above the curves for circular tubes. The data
of [5] lie close to curves 1 and 2. The experimental results are in satisfactory agreement with the approximate
analytical calculations of Buleev and Pyshin. They solved the problem by a numerical method. In the cal-
culations, they used the following approximation for the turbulent viscosity coefficient: in the laminar sublayer,
uT = 0, and in the turbulent nucleus,
y x(1-( r0+x -r)2]u
1 V l 2 v c2
where r o is the rod radius; x is the distance from the center of the mesh of the bundle to the rod surface; c =
= 7.1 is an empirical coefficient found by comparing the calculated velocity profile in a circular tube with the
experimental data of Nikuradze.
In our experiments, we found a weaker dependence of a on Re for the bundle with s/d = 1.13 than in the
experiments for a smooth circular tube. If the cause for this is the roughness, then the absolute value of the
roughness found from formula (2) should fall within the limits 0.005-0.01 mm. However, in complex channels,
the influence of roughness may be different from that in a tube.
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It should be noted that the resistance coefficient increases with the relative spacing of the rods in the
bundle. This is not unexpected. In fact, with the use of a hydraulic diameter as a 'universal" dimension,
it is assumed that the mean dimensionless tangential stress T ~- depends only on the Re number and the
pw
roughness of the channel. It is difficult to expect that this assumption is valid for any channel. - In narrow
parts of the mesh of a densely packed bundle, a stagnant zone may be created. As a result, the basic flow of
the liquid will take place only through parts of the mesh cross section. From this it follows that the friction
resistance coefficient calculated from dh will be smaller than for a circular tube.
9,8
0,6
10 2 4 6 8 104 2 4 6' 8 105 2 4 6 8 106
Re
Fig. 1. Friction resistance coefficient of staggered bundles (the hydraulic diameter is taken as the
characteristic dimension). 0, 0) Data of the present work for d = 1.0 and d = 1.13,
respectively; 0) data of [3] for
=1.76 - 2.37; A ) data of [4] for .. d =1.46; A )
data of [5] for d = 1.12; 0, ?) data found from the calculations of Buleev and Pyshin for
=1.0 and s d = 1.2, respectively. 1) Data calculated from formula (1); 2-4) data calculated
from formula (2) for dh = 4.8 mm with A equal to 0, 0.005, 0.010 mm, respectively.
In some channels, the resistance coefficient can prove to be larger than for a circular tube, owing to
secondary currents arising in the liquid (secondary currents are observed in the experiments of Nikuradze).
The data obtained for a densely packed bundle were recalculated under the assumption of the existence
of stagnant zones. As the characteristic dimension, we used the diameter of a circle inscribed in the bundle mesh.
For such an analysis, these data are in agreement with formula (2) for a circular tube (Fig. 2). Apparently, such
good agreement is accidental, since by means of the inscribed circle one can only characterize qualitatively the
stagnant zones.
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With an increase in the velocity of the liquid, the stagnant zones in a densely packed bundle should
probably decrease. Therefore, it is logical to assume that at very large Re numbers the characteristic dimension
of the densely packed bundle will practically coincide with the equivalent hydraulic diameter.
Fig. 2. Data for a densely packed bundle analyzed from the diameter of the circle inscribed in the
bundle mesh. 0) experimental data. 1) Data calculated from formula (1); 2) data calculated
from formula (2) for A = 0.
On the basis of the experiments and analysis of the data of other authors, it was established that the use of
the hydraulic diameter as a characteristic dimension in the generalization of experimental data from the hydraulic
resistance of bundles does not lead automatically to the taking into account of the spacing of the rods in the
bundle.
The obtained data are recommended for practical calculations of the resistance of bundles in the axial
flow of a liquid.
LITERATURE CITED
1. H. Schlichting, Boundary Layer Theory [Russian translation] (Moscow, IL, 1956) p. 394.
2. I. I. Agroskin, G. T. Dmitriev, and F. I. Pikalov, Hydraulics [in Russian) (Moscow, State Energetics
Press, 1954).
3. A. P. Salikov, Ya. L. Polynovskii, and K. I. Belyakov, Teploenergetika 8, 48 (1954).
4. P. Miller, J. Byrnes, and D. Benforado, Am. J. Chem. Eng. 2, 226 (June, 1956).
5. B. LeTourneau, R. Grimble, and J. Zerbe, Trans. Am. Soc. Mech. Engr. 79, 8, 1751 (1957).
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V. I. Subbotin, P. A. Ushakov, and I. P. Sviridenko
Translated from Atomnaya Energiya, Vol. 9, No. 4, pp. 310-312, October, 1960
Original article submitted October 28, 1959
For the purpose of obtaining data that has not appeared heretofore in the literature and for studying the
effect of two-way heat elimination on heat transmission, in 1958, the authors of the present paper investigated
the heat exchange involved in a turbulent flow of mercury in narrow annular ducts.
The experimental ducts, which were made of carbon steel, had the following dimensions. First duct:
I = 1000 mm, d2/dl = 1.05, width of the annular space 1 mm; second duct: l = 400 mm, d2/dt = 1.09, width
of the annulus 2 mm. Uniform heat elimination was realized with electrical heating units. There were no
sections for preliminary hydrodynamic stabilization. The mercury was filtered beforehand through chamois.
To protect the mercury against oxidation, the free space of the sections was filled with argon.
Fig. 1. Temperature variation along the
annulus of the second duct for q =
= 50 ?.108 kcal/rr?. hr: a) two-way
heating; b) one-way heating; 0, 4 )
temperature of the wall and mercury,
respectively.
The temperature of the heat-exchange surface was measured
by twelve thermocouples having diameters of 0.2 mm; these
were embedded at uniform spacing along the length of the outer
tube. The effective temperature correction for depth of insertion
of the thermocouples was estimated from the results of measure-
ments in water of the heat transfer in a circular tube in which the
thermocouples had been inserted in the same manner as in the
annular ducts, as well as by electrosimulation of the thermocouple
weld. This correction was insignificant (approximately 0.050 C
for a flow of approximately 5.104 Cal/ m2 ? hr).
In the tests the heat transfer coefficients, stabilized along
the length of the ducts, were determined. The hydraulic diameter
dh, equal to twice the width of the annulus, was taken as they
characteristic dimension in the similarity criteria. The physical
parameters were calculated according to the mean temperature
of the mercury. The tests were performed in a mercury flow
velocity interval of 0.3-3.8 m/sec at a temperature of 30-40? C
and heat flow q ranging from 25.10'3 to 75 ? log kcal/m2?hr.
The limiting analytic errors in measuring heat transfer in
the second duct was no greater than 16 and 30% for one-way and
two-way heating, respectively. The measurement accuracy on
the first duct was somewhat lower.
Experimental Results
In carrying out the preliminary preparations and adjustments
for the first duct, bending of the inner tube was noted to exert
considerable influence on the temperature field of the heat-exchange surface. With a crimp in the tube of
approximately 0.3 mm and q w 40.10'3 kcal/m2. hr, the temperature nonuniformity in the heat-transferring
wall reached 50-60? C. This indicates that in designing for the heat-exchange properties,it is important to take
into account the effect of deformation of the ducts.
The heat transfer was measured on the first duct only after careful checking of the duct. The main results
were obtained on the second experimental duct, where deformation of the channel was almost entirely avoided.
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Nu
40
10
8
6
40
8
6
10
. ~e
so
1
00
Pe
2 4 6 8 f0 2 4.904
Fig. 2. Comparison of results of the present paper with those calculated
on the basis of serniempirical equations: a, ?) Experimental data,
obtained with two-way heating on the first and second ducts, respectively;
O, ?) experimental data, obtained with one-way heating; )
calculation according to Eqs. (1) and (2); ----- ) calculation according
to Eqs. (3) and (4).
Fig. 3. Comparison of data of the present paper with the data of other authors (one-way heating):
for mercury: (D, ? ) data of present paper for the first and second ducts, respectively; ?, 0,
iV, o) data from [4-7] for d2/dl equal to 1.54; 1.71; 1.85; 1.17 - 1.75, respectively; for
Pb-Bi alloy: ?, A) data from [8] and [9] for d2/di equal to 1.74 and 1.25, respectively;
) calculation according to Eq. (2).
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The temperature distribution over length of the duct is shown in Fig. 1. The heat stabilization length was
approximately 30 dh, which corresponds with the data for a circular tube [1].. The results of the tests, .processed
on the basis of the dimensionless criteria, are shown in Fig. 2. The slight disparity in points is explained by
errors of measurement. The solid lines in the figure correspond to the semiempirical equations of Harrison and
Menke for a plane-parallel duct [2]:
A'uz= 10- --0,025Pe 0,",
Nu, -4,9 --0,0175 Pe?,91'.
(1)
(2)
The dashed lines are constructed from equations given by Buleev for one-way heat elimination outward from the
outer wall of the duct, d2/di < 2:
and for symmetrically heated plane-parallel ducts:
N'12= 9d-(1,41 -Oj2 18 100 Pr) (Nu, -5). (4)
Equations (3) and (4) were. derived on the basis of the methods outlined in [3].
The satisfactory agreement between the experimental data and those calculated from semiempirical
equations indicates that either the contact thermal resistance on the heat-exchange surface is small,or that the
assertions of the polyempirical theories are not precisely correct.
From Fig. 2, it is evident that heat transfer with two-way heating is about twice as great as with, one-way
heating.
In Fig. 3 the data of various authors on heat transfer in gaps for mercury [4-7] and Pb-Bi alloy [8, 9J are
compared. Such a comparison is valid, since the contact resistances of the mercury and of the Pb-Bi alloy are
of the same order [10].
The effect of the curvature of the surfaces generating the annulus are often accounted for with the
correction (d2/dl) 0.8 But the introduction of this factor only aggravates the disparity of the experimental
points.
Consequently, it has been experimentally established that:
1. With two-way heating of an annular duct having d2/di < 1.09, and under the condition that the heat
fluxes on both surfaces, the heat transfer to the mercury is about twice as great as with one- way heating. The
application of the hydraulic diameter as a characteristic dimension does not therefore imply automatic allowance
for the characteristics of heat transfer to liquid metals with one-way and two-way. heating of the annular duct.
2. In a narrow annular duct, with no preliminary hydrodynamic stabilization, the thermal stabilization
length for mercury comes to approximately 30 dh.
3. Disturbing the geometry of the annular duct has a substantial effect on heat transfer to liquid metals.
In addition to the authors themselves, B. N. Gabrianovich and A. V. Zhukov took part in the testing.
1.
H. Johnson, W. Clabaugh, and J. Hartnett, Trans. ASME 76, 4,
505 (1954).
2.
W. Harrison and J. Menke, Trans. ASME 71, 797 (1949i.-
3.
N. I. Buleev, Collection: Heat-Exchange Problems [in Russian]
(Izd-vo AN SSSR, Moscow, 1959),
p. 208.
? Nul corresponds to one-way, Nut to two-way heat elimination.
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 Declassified and Approved For Release 2013/02/14: CIA-RDP10-02196R000100060004-4
4. S. S. Kutateladze et al., Liquid Metal Heat-Transfer Media (Supplement No. 2 to the journal Atomnaya
Energiya) [in Russian] (Atomizdat, Moscow, 1958) p. 57."
5. M. A. Styrikovich, I. E. Semenovker, and A. R. Sorin, Sovetskoe Kotloturbostroenie 9, 316 (1940);
Zhur. Tekhn. Fiz. 16, 10, 1324 (1940).
6. M. I. Korneev, Teploenergetika 7, 30 (1955).
7. L. Trefethen, General Discussion on Heat Transfer (1951) p. 436.
8. R. Seban and D. Casey. Trans. ASME 79, 7, 1541 (1957).
9. B. Lubarsky and S. Kaufman, Rept. Nat. Advis. Comm. Aeronaut. 1270 (1956).
10. V. I. Subbotin, M. Kh. Ibragimov, and P. A. Ushakov, Atomnaya Energiya 8, 1, 54 (1960).
B. P. Kiselev
Translated from Atomnaya Energiya, Vol. 9, No. 4, pp. 312-313, October, 1960
Original article submitted May 16, 1960
A method of rectification is applied for the concentrated isotope P. BF3, BC13, and BBrs were employed
as the working liquids.
From 'approximate calculations [1] the separation factor a for the system B10 Cl3 - BU Cl3 should be equal
to 1.013. Experiments [2] have shown that the separation factor for boron isotopes with rectification of BC13 is
less than the calculated value by nearly one order of magnitude, and the light isotope of boron is concentrated
in the liquid phase. In the literature [31,investigations
of the temperature dependence of the separation
Data on Separation of the Boron Isotopes factor are described. At temperatures above 61.7 ? C
Stage of enrichment Bit/Bin the isotope Btt is highly volatile, and at temperatures
I I a
below 61.7 ? C this is the case for B10; the separation
factor for the system B10C13 - B 1 Cl3 varies, depend-
Initial . . . . . 4,49?0,02
ing on the temperature (from - 85 to + 13 ? C)
First deg. 4,35?0,02 1,031?0,008
4,45?0,02
Initial from 0.998 to 1.003 [3].
First de . . . . 4,33?0,02 1,0'28?0,008
Initial g 4,60?0,03
In addition to the rectification method for
Second deg... ? ? . . 4,39?0,04 1,025?0,01
separating the boron isotopes, the reaction of
Initial . . 4,63?0,03
Second deg .? ? ? ? 4,38?0,02 1,028?0,008 chemical isotope interchange is of considerable
interest. It has been shown [4, 5] that in chemical
Average 1,028?0,008 interchange between gaseous BFg and a liquid coin-
, - 1
plex compound of anisol with boron fluoride the
light isotope B10 is concentrated in the liquid phase.
The separation factor for this reaction is equal to
1.013 t 0.005.
The interchange reaction between the gas molecule BF3 and ion BF; has also been suggested. The inter-
change between hydrofluoroboric acidt and gaseous BFs can be written as follows:
B10Fa+B11F3 F-' B11F.~+B'?F3,
? Original Russian pagination. See CB translation.
t Saturated 18.5 N aqueous solution of hydrofluoroboric acid.
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Experiments have shown that in bringing about.the given reaction, separation of the boron isotopes is ob-
served, where the light isotope B10 is concentrated in the gaseous phase. In the table are shown data from some
tests for one and two degrees of interchange. Isotope analysis was performed on the MS-2 M mass spectrometer.
The separation factor for interchange between hydrofluoroboric acid and gaseous BF3 is almost one order of
magnitude greater than in the case of distillation of BF3 and is equal to 1.028 ? 0.008. These tests were per-
formed at a temperature of 20- 30 ? C.
The author expresses his indebtedness to B. P. Konstantinov for suggestion of the topic and advice during
the work.
Taking part in the experiments and measurements were Yu. P. Batakov, O. N. Shuvalov, and Yu. G.
Basargin.
LITERATURE CITED
1.
H. Urey, Chemistry of Isotopes, Collection I [Russian translation] (IL, 1948) p. 86.
2.
M. Green and G. Martin, Trans. Faraday Soc. 48, 5 (1952).
3.
N, N. Sevryugova,
O. V. Uvarov, and N. M, Zhavoronkov, Atomnaya Energiya 4, 113 (1956).'
4.
G. M. Panchenkov,
V. D. Moiseev, and A. V. Makarov, Zhur. Fiz. Khim. 31, 1851 (1957).
5.
G. M. Panchenkov,
V. D. Moiseev, and A. V. Makarov, Doklady Akad. Nauk SSSR 112, 4, 659
(1957).
DEMARCATION OF PETROLEUM-BEARING AND WATER-BEARING
STRATA WITH THE APPLICATION OF ELECTRON AND
PHOTON BEAMS
V. I. Gomonai, I. Yu. Krivskii, N. V. Ryzhkina,
V. A. Shkoda-Ul'yanov, and A. M. Parlag
Translated?from Atomnaya Energiya, Vol. 9, No. 4, pp. 313-315, October, 1960
Original article submitted February 10, 1960
The demarcation of petroleum- and water- bearing strata in oil wells is a very complex problem because
of the similarity of certain properties of water and petroleum. The solution to this problem is currently based
on the presence in the water- and petroleum- bearing strata of various concomitant chemical elements (1, 2].
A direct method for solving this problem can be based on the application of various nuclear properties of
the isotopes intering into the composition of the water- and petroleum-bearing strata. For example, these
isotopes are distinguished by energy thresholds of the (y. n) reaction (see table).
From the table, it is evident that in the water and petroleum there exist isotopes with different (y, n)
thresholds. If the water and petroleum could be irradiated with electrons or photons with energies of 2.23 Mev,
as a result of the (y, n) reaction on deuterium, photoneutrons would be formed in the body of the water and
petroleum; with higher energies photoneutrons are formed as the result of isotopes whose (y, n) reaction threshold
is equal to or less than the energy of the incident electrons.
The number of photoelectrons formed in an infinite layer of water or petroleum upon irradiation by
electrons can be determined with a high degree of precision, as shown in [11], by means of the equilibrium
? Original Russian pagination. See CB translation.
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photon spectrum. In the case of irradiation by electrons this number of photoneutrons, as the result of fully
developed photon showers, can be calculated from the following equation [11, 12]:
_ FO Fyn E
f) (Ep) - 1 E] APP (EO. E) A.
a tot (
E
Here Q(0) is the number of photoneutrons formed by a single electron with energy E0; `p, (e o, e) _
F6
= eE S x2 dx; E = 2, is ; E(,= 2, -- (E is the photon energy in the interval where oy,, (E) 0; 8 is the
critical energy) ; o y it (e) is the (y , n) reaction cross section; atot (e) is the total photon absorption cross
section, which was calculated from the equation [121
atot (e)= acomp (e)+ opar (e).
In the calculation of ocomp( E) for water,the fraction of electrons borne by oxygen and hydrogen atoms
was considered; in the calculation of acomp ( e) for petroleum,those borne by carbon and hydrogen were con-
sidered. In the determination of apar (e) the fraction
(y, n) Reaction Thresholds for Various Isotopes of nuclei of each element in the corresponding mole-
Occurring in the Composition of Water and Oil cules was considered.
Content of isotope, mol.%
y, n-reac-
Isotope
in water
in petro-
tion thres-
I
leum
hold, Mev
0,0147 [3]
0,0147 131
2,2314, 5]
C"
-
98,892 161
18,75[7]
C19
-
1,108 [6]
4,95[8,9]
01(
99,760 1101
--
15,63[7]
017
0,042 1101
--
4,1/,181
018
0,198 [10]
--
8,0 [81
C"+D!
/
Cf3
3 4 5 6
7 8
9 10 E. M
ev
Fig. 1. Photoneutron emissions from water and
petroleum due to the (y, n) reaction on
deuterium and the isotope C13.
The value of the critical energy for water is
equal to 66.92 Mev, for petroleum 84.03 Mev [12].
A calculation of 8 and dtot (e) was made for
petroleum having a density p = 0.9 g/cm3, molecular
weight of 256, and containing 87.0156 carbon and
12.15% hydrogen, which corresponds to 31.4 atoms of
hydrogen and 18.7 atoms of carbon belonging to a
single average petroleum molecule [13]. The density
of electrons in the petroleum was 3.05 per cm3.
22 E, Mev
Fig. 2. Photoneutron emissions from water and
petroleum due to the (y, n) reaction on the
isotopes 01fi and C12.
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In order to explain the reason for using the (y, n) reaction on deuterium for the separation of petroleum
from water in wells, calculations of Q( e) for petroleum and water were made; it was assumed in'so doing that
photoneutrons were formed exclusively as the result of photofission of the deuterium, the content of which in the
petroleum and water was assumed to be the same.
The results of calculating Q(E), as obtained by numerical integration with the.aid of Simpson's equations,
show (Fig. 1) that the number of photoneutrons formed in water and petroleum having the same energy is almost
the same. However, in the literature [14-16] there are indications of an increased deuterium content in the
hydrogen of the petroleum by 50% or more. Consequently, taking into account the true deuterium content in
petroleum, one can obtain a marked difference in the number of photoelectrons formed in water and petroleum.
From the table it is clear that in the petroleum, in addition to deuterium, there is found the isotope Cis
with a ,rather low (y, n) -reaction threshold (4.95 Mev). From the corresponding calculations of Q (E) given in
Fig. 1, it follows that the number of photoneutrons formed in the petroleum as the result of Cls when the energy
of the incident electrons is 7.6 Mev, is 105 neutrons/ p a of the electron current. The total number of photo-
electrons formed due to C13 and deuterium for this same energy is 2.75 neutrons /p a. The total number of
photoneutrons of the isotope C13 and deuterium in the petroleum is shown in Fig. 1 by a broken line. In the water,
with an electron energy of 7.6 Mev,the photoneutrons are formed only due to deuterium.
The water also contains the isotope 017, whose (y , n) reaction threshold is equal to 4.14 Mev. In the
literature there are no data giving information on the (y , n) reaction for this isotope, so that an accurate deter-
mination of its contribution to the formation of photoneutrons in water is impossible. If we assume that the cross
section oy n (E) of the isotope 017 is equal to the cross section for the isotope C13 and take into account the
low content of the isotope 017 in water (0.042% [10]), then it is possible to obtain the number of photoneutrons
in water (due to 017), which is approximately 1/40 times the number of photoneutrons formed in petroleum due
to C13.
As evident from Fig. 2, in the case when the principal isotopes of carbon (C12) and oxygen (016) are used
with higher energies of the incident electron beam - of the order 19 Mev [higher than the (y, n) reaction
threshold for 016] - in water, about 3' 107 neutrons/pa are formed, whereas in petroleum considerably fewer
neutrons are formed for this energy. The photoneutrons formed due to C13 and deuterium may be neglected
because of the low percentage content of these isotopes in comparison with the percentage content of 016 and
C12. Consequently, in the energy interval 17- 21 Mev, favorable conditions are created for solution of the problem
in question.
Therefore, resting on the formation of different numbers of photoneutrons in water and in petroleum, the
problem of segregating petroleum- and water-bearing strata can be solved by application of electron beams with
energies of the order 8 Mev and higher. For this purpose, it will be necessary to develop physically small electron
accelerators with high intensities.
The authors express their appreciation to G. Z. Borukhovich and V. M. Vorobeichik for assisting in the
numerical calculations.
1. A. I. Kholin, Session of the Acad. of Sci. of the USSR on the Peaceful Uses of Atomic Energy (meetings of
the technical science divisions) [in Russian] (Izd-vo AN SSSR, Moscow, 1955) p. 267.
2. N. K. Kukharenko, Yu. S. Shimilevich, D. F. Bespalov, and V. A. Odinokov, Heft. Kh-vo 3, 43 (1956).
3. A. M. Brodskii, Chemistry of Isotopes [in Russian] (Izd-vo AN SSSR, Moscow, 1957).
4. R. Sher, J. Halpern, and A. Mann. Phys. Rev. 84, 387 (1951).
5. J. Blatt and V. Weisskopf, Theoretical Nuclear Physics [Russian translation] (IL, Moscow, 1954).
6. D. Strominger and J. Hollander, Rev. Mod. Phys, 30 (2) 2 (1958).
7. R. Montalbetti, L. Katz, and J. Goldemberg, Phys. Rev. 91, 659 (1953).
8. N. A. Vlasov, Neutrons [in Russian] (Gostekhteorizdat, Moscow, 1955).
9. B. Cook, Phys. Rev. 106, 300 (1957).
10. I. Kirschenbaum, Heavy Water [Russian translation] (IL, Moscow, 1953).
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11. V. I. Gol'danskii and V. A. Shkoda-Ul'yanov, Zhur. fksp. i Teoret. Fiz. 28, 629 (1955).
12. C. Z. Belen'kii, Shower Processes in Cosmic Rays [in Russian] (Gostekhteorizdat, Moscow, 1948).
13. M. M. Kusakov, Methods for Determining the Physical-Chemical Characteristics of Petroleum Products
[in Russian] (ONTI, Moscow, 1936).
14. N. S. Filipova, Dokiady Akad. Nauk SSSR 3, 29 (1935).
15. V. I. Vernadskii, A. P. Vinogradov, and R. V. Teis, Dokiady Akad. Nauk SSSR 31, 573 (1941).
16. I. V. Grinberg and M. E..Petrikovskaya, Geologicheskii Zhurnal 17, 4 (1957).
INVESTIGATION OF ATTENUATION FUNCTIONS IN WATER
FOR NEUTRONS FROM ISOTROPIC AND ONE-DIRECTIONAL
FISSION SOURCES
V. A. Dulin, Yu. A. Kazanskii, V. P. Mashkovich,
E. A. Panov, and S. G. Tsypin
Translated from Atomnaya fnergiya, Vol. 9, No. 4, pp. 315-317, October, 1960
Original article submitted April 27, 1960
Q
In the literature, functions are given for the attenuation in water of neutrons from isotropic sources; for
example, point [1] and disk' [2] sources. However, in a number of instances it is required to know the attenuation
function for neutrons from plane one-directional sources; for instance, when the source is located at a large
distance from the shield, and when it may be assumed that the radiation is incident normal to its surface.
In the present paper an experimental investigation is
made on the spatial distribution in water of fission neutrons.
As a neutron source we used a BR-5 reactor [3, 4]. The neutrons
entered the concrete shield through a channel with diameter of
Fig. 1. Geometry of the experiment.
approximately 250 mm and impinged on a tank filled with bi-
distilled water (dimensions of the tank 137 x 139 x -217 cm).
A neutron beam with total angular divergence of approxi-
mately 5 ? entered the tank through the middle of a 137 x 139 cm
wall. For neutron detectors we used proportional boron counters.
The applied apparatus permitted us to run measurements at any
point in the tank, fixing the position of the detector with an
accuracy of 1 mm in the horizontal and vertical directions.
The proportional counter was situated at various distances
r from the source and was moved in the direction h perpen-
dicular to the beam (Fig. D.
The measured neutron distribution is shown in Figs. 2
and 3.
In Fig. 4 is shown the attenuation function for neutron from an isotropic point source, the function being
multiplied by r2 (curve a), along with the attenuation function for neutrons from a plane one- directional
source (curve b). These functions were obtained on the basis of the experimental data with the aid of the
transformations for an isotropic point source
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n/2
G,(r)=Cr ( N(r, O)sinOdO,
and for a plane one-directional source
Gp1 (r)=C2 .1 N (r, h) ii A.
0
In these equations, N (r, 0) and N (r, h) are the distributions given in Figs. 2 and 3; C1 and C2 are con-
stants. In calculating GT (r) a correction for the finiteness of the source is introduced.
ly pulse/min
1
4
6
-
7
-8
\\
20 40 60 80 100 120 1404cm
Fig. 2. Measured neutron distribution for the
following values of r (cm): 1) 33.4; 2) 43.4;
3) 53.4;. 4) 63.4; 5) 73.4; 6) 83.4; 7) 93.4;
8) 103.4; 9) 113.4; 10) 123.4; 11) 133.4;
12) 143.4; 13) 153.4.
The maximum error for curve a occurs for small
r attaining a value of approximately 201o for r =
= 40 cm. The minimum error (approximately 5%) is
observed.for large r . The error for curve b varies from
approximately 5% for r = 40 cm to approximately 20%
for r = 140 cm.
In examining the curves a and b (Fig. 4), we
see that the attenuation functions represented by these
curves (for an attenuation of approximately 106) differ
by approximately 20%, which lies practically within
,the limits of probable error.
10
2
3
4
5
6
7
9
20 40 50 80 100 120 140 160r, cn
Fig. 3. Measured neutron-distribution for the following
values of h (cm): 1) 0; 2) 20; 3) 30; 4) 40;
5) 50; 6) 60; 7) 70; 8) 80; 9) 90.
Consequently, for practical problems in cal-
culating the shielding for water thicknesses greater
than 40 cm,these attenuation functions may be assumed identical. This conclusion confirms the assumption
expressed by Goldstein [1].
For the purpose of comparison with the attenuation functions determined experimentally in (2) for neutrons
from an isotropic disk source 71.2 cm in diameter (curve b), in Fig. 5 is shown the attenuation function for
neutrons from an isotropic disk source of the same diameter (curve a), but now calculated according to the
equation
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10s
90 5
90 4
90 3
102
10'
100L
0
Fig. 4. Attenuation function for neutrons from an isotropic point source, multiplied by r2
(curve a), and a plane one-directional source (curve b).
106
104
102
101
100
10-IL
0
20 40 60
140 r, cm
Fig. 5. Attenuation function for neutrons from an isotropic disk source 71.2 cm in diameter.
a) Results of the present paper; b) experimental curve from [2].
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G. disk (r, a).-;= 2n
where Gp (R) is the attenuation function for neutrons from an isotropic point source; a is the radius of the disk.
Comparison of these curves for r > 40, cm for an attenuation of 3.105 exhibits maximum divergence by a
factor of two.
The authors express their gratitude to O. I. Leipunskii and V. V. Orlov for valuable remarks in a discussion
of the work.
LITERATURE CITED
1. H. Goldstein, The Attenuation of Gamma-Rays and Neutrons in Reactor Shields. U. S, AEC, Washington,
1957.
2. G. Chapman and C. Storrs, Effective, Neutron Removal Cross Sections for Shielding U. S. AEC, Report
AECD-3978 (1958).
3. A. I, Leipunskii et al., Atomnaya Energiya 5, 3, 277 (1958).*
4. Atomnaya Energiya 7, 2, 192 (1959).*
V. A. Dulin, V. P. Mashkovich, E. A. Panov, and S. G. Tsypin
Translated from Atomnaya Energiya, Vol. 9, No. 4, pp. 318-319, October, 1960
Original article submitted April 27, 1960
In [1] the method of moments was used to obtain the energy spectrum of neutrons in water at various
distances from an isotropic point source of fission neutrons. The experimental determination of the neutron
spectrum in water at depths down to 30 cm from the reactor was the subject of [2]. However, in view of a
difference in geometries, it is impossible to compare directly the results of these papers.
Characteristics of the Threshold Indicators, in Order of Increasing Threshold Energy
Reaction
Decay
half life
Energy
threshold,
Mev
Indicator material
Indicator height, mm
P91 (n p) Si31
155 min
-,-1,5
Solution of chemicallypure
3,5
phosphor in paraffin
S32 (n, p) P32
14,3 days
-1,5
Melted or pressed chemical
6,0
ly pure sulfur
AI27(n, p) Mg27
9,8 min
-2;1
Metallic aluminum
4
0
A127 (n, a) Na24
15 hr
---6',0
The same
,
4,0
* Original Russian pagination. See CB translation.
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In the present investigation, the energy distribution in water of fission neutrons was studied experimentally.
As a neutron source we used a BR-5 reactor [3, 4]. A description of the experimental setup is given in [5].
0
600
U
2 400
0 9 18 27 36 45 8, deg
Fig. 1. Activity of threshold indicators as a
function of the angle 0 for r = 30 cm: )
P3 (n, p) Si"; - - - - ? ) SM (n, p) P sZ
-..-..-..-) Al27 (n,p)Mg"; ---.- )
Al" (n, a) Na 2
For detectors of the fast neutrons we used threshold
indicators in the form of disks 35 mm in diameter and of
different heights. The data for these indicators are shown
in the table.
The threshold indicators (in the form of three-
indicator assemblies) were placed in the water at various
angles 0 with the direction of the incoming neutron beam
and at various distances h perpendicular to the beam (see
[5], Fig. 1). All threshold indicators in the assembly
were activated simultaneously. The absence of an effect
was established experimentally by the mutual shielding of
the detectors in the assembly within the limits of experi-
mental error. The induced activity of the indicators was
determined by constructing the decay curves, which made
it possible to exclude the influence of foreign impurities.
The activities of the indicators, normalized to 0 =
= 9?, are shown in Fig. 1 as a function of 9 for r = 30 cm.
The activities of the phosphor indicators, normalized
to h = 0, are given in Fig. 2 as a function of h for r =
=30 cm and r=60 cm.
The energy distribution of neutrons in water at dis-
tances of 30 and 60 cm, recalculated from the geometry
of the experiment for a point source, is shown in Fig. 3.
Fig. 2. Activity of phosphor threshold in-
dicators as a function of h for r = 30 cm
and r = 60 cm.
0 2 4 6 8 10 12
E, Mev
Fig. 3. Spectral distribution in water of neutrons from
an isotropic point fission source: ) analytic
spectrum [1]; - - - - ) spectrum obtained in the
present paper.
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The neutron spectral distribution was determined in the solution of the following set of equations by the.
method of successive approximations:
Nt(r)=c? (1-e-siT)eI it ` a) (r, E) Qt (E)dE=cei(1-e_~i'
Eti
-kit ` (Dj (r. E) oij (k) AE, j, (1)
? j=i
where Ni (r) is the activity of the ith threshold indicator, located at a distance r from the source, after irradiation
for a period of time T and waiting for a period of time t ; ei is the efficiency of recording of the indicator
activity, including corrections for absorption and scattering in the sample, air and the. inlet window of the counter,
the,geometry, resolving time, etc.; oi(E) is the cross section of the given reaction for an energy E [6]; (r, E)
is the differential flux of neutrons with energy E at a distance r from the source; c is a constant; i is the
indicator index (i =1, 2, ... , n); j is the energy interval index. The quantities Ni (r) are determin7from
the equation
n/2
Ni (r)=c1 S Ni (r, 0) sin A dB,
0
(2)
where Ni (r, 0) is the activity of the ith threshold indicator, situated at a distance r from the source at an angle
0 with the direction of the neutron beam; c1 is a constant.
The relative value of the quantity ei for each indicator was determined experimentally by activating the
threshold detectors with a beam of monoenergetic 14.1 Mev neutrons and a flow of fission neutrons.
The set of equations (1) was solved by the matrix method.
In Fig. 3,the data obtained are compared with an analytic neutron spectrum [1]. The experimental spectral
histograms are normalized to the analytic spectrum at r = 30 cm. The discrepancy, determined relative to the
area beneath the corresponding portions of the curves, is 30-50016, which is almost entirely within the limits of
probable error of the spectral distributions obtained, which is estimated at 30 ,. The error due to finiteness of
the source is included in the experimental error.
The authors express their deep appreciation to O. I. Leipunskii and V. V. Orlov for valuable remarks in a
discussion of the work.
1. R. Aronson et al., Penetration of Neutrons from a Point Isotropic Fission Source in Water, U. S. AEC,
Report NYO-6267 (1954).
2. R. Cochran and K. Henry, Fast Neutron Spectra of the BSF Reactor, U. S. AEC, Report AECD-3720 (1953).
3. A. I. Leipunskii. et al., Atomnaya Energiya 5, 3, 277 (1958).?
4. Atomnaya Energiya 7, 2, 193 (1959).
5. V. A. Dulin et al., Atomnaya Energiya 9, 4, 315 (1960).
6. D. Hughes and R. Schwartz, Neutron Cross Sections, Second edition (New York, 1958).
? Original Russian pagination. See CB translation.
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CALCULATION OF THE DOSE CREATED IN AN IRRADIATED
OBJECT MOVING IN THE RADIATION FIELD OF A
LINE SOURCE
U. Ya. Margulis, S. M. Stepanov, and V. G. Khrushchev
Translated from Atomnaya Energiya, Vol. 9; No. 4, p. 320, October, 1960
Original article submitted March 18, 1960
The existing methods for calculating the dose field generated by.sources of different. configurations are
referred, as a rule, to the case when the objectJs fixed relative to the source [1-4]. However, in industrial
y-ray equipment it would be more sensible to use a continuous method of irradiation. In this connection, it
would seem desirable to come up with a method for calculating the total radiation dose created inside the
irradiated object during its motion in the radiation field of the source.
The most functional form of radiator in industrial y -ray equipment is an activated rod or plane, which can
be considered as a system of line sources. Thus, under definite conditions, the analysis can be reduced to a deter-
mination of the dose created by a line source.
A line source of length L is placed on the Z axis, one of its ends being located at the origin. The irradiated
object is moved with constant velocity y (em/min)` along a straight line parallel to the Y axis (see the figure).
We determine the radiation dose created at some point A inside the irradiated object during the time of its
motion along a linear segment of length So:
*oipo
D_2kvm"[ 1 (' 1' e-?h(al-f 1)secIsec q) I secs dJdw +A2 ~0 1oe-?h(a2-+ -1)sec1)secTSeelpdipdgj,
V L ) 1
0 0
To = arctg li `c 1)o =arctg Lso
Y
Diagram for calculating the dose of an object while
moving in the field of a line source: AB = H is the
shortest distance between the point A and the rod;
Aa = h is the thickness of the absorbent layer along
the X-axis; AO=H'; Ab = h'; H' =H sec 0; h' = h sec
m is the linear activity of the source (p C /cm); p is
the linear attenuation factor of a narrow y-ray beam
in th e irradiated object; Ar, A2, al, a2 are constants
appearing in the analytic expression for the accumulation
factor, which takes into account the role of multiple
scattering. The value of the dose is given for a point
located on the perpendicular erected from the end point
of the rod (in our case the origin). The radiation dose
at any other point of the object will be equal to the sum
or difference of doses obtained from the corresponding
line sources.
The formula given is valid if the irradiated objects
are placed on a conveyor belt near enough to one another,
so that the distances between them may be neglected and
so that we may assume the conveyor belt to be filled in
If the velocity is expressed in cm/min, the y constant
of the source ky will be expressed in r/min.
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solidly with the substance to be, irradiated. This same equation is yalid for calculating the dose generated by a
source in the form of a rectangular plane, which may be regarded as a system consisting of n line sources. If
So is large enough, so that the dose created at some point A at, the end of the path of each line source becomes
sufficiently small, then the total dose acquired by the object upon moving in the field of the plane source will
be equal on the .path So to nD, where D is the dose generated by one line.source, n is the number of line sources
composing the active plane.
In those instances when it is necessary to determine the dose power generated by an active plane with
stationary irradiation, the equation given may also be used, assuming half the path length So/2 to be equal to
the linear dimension of the plane in the direction of the Y axis.. In this case the dose is determined at a point
located on the perpendicular erected from a corner of the rectangular plane.
1. G. V. Gorshkov, Gamma Radiation of Radioactive Bodies [in Russian] (Izd LGU, 1956).
2. ' Nuclear Reactor Shields (edited by T. Rockwell) [Russian translation] (IL, Moscow, 1958).
3. D. P. Osanov and E. E. Kovalev, Atomnaya Energiya 6, 6. 670 '(1959).*
4. A., V. Bibergal', Atomnaya Energiya 7, 3, 244 (1959). `
SIMPLE CALORIMETRIC METHOD OF MEASURING THE ABSOLUTE
ENERGY DOSE RECEIVED FROM POWERFUL SOURCES OF
IONIZING RADIATION
M. B. Fiveiskii, Yu. S. Lazurkin, and M. A. Mokul'skii'
Translated from Atomnaya Energiya, Vol. 9, No. 4, pp. 321-323, October, 1960
Original article submitted April 14, 1960
The energy dose is an important characteristic condition of irradiation in all radiochemical, radiobiological,
and material-testing studies involving nuclear reactors and other sources of powerful penetrating radiation.
In order to determine the energy dose, one usually uses a calorimetric method which gives directly the
amount of radiation energy absorbed by the sample under the condition that the release or absorption of energy
as a result of other processes taking place in the substance can be neglected. t In this procedure, a stationary
calorimetric method is used. For the cases of intensive irradiation (high dose rate) this method is not convenient,
since it has basic shortcomings, the main one being the considerable time for the establishment of thermal
equilibrium. The dosimetric sample can then heat up considerably, even to the temperature of melting or de-
composition; moreover, it receives a large integral dose, which can essentially change its structure and properties.
In this connection, we developed, in 1957, a simple nonstationary calorimetric method that is suitable for a
nuclear reactor and other strong sources of radiation. The practical application of this method during three years
has established its suitability, reliability, and sufficient accuracy. .
Original Russian pagination. See CB translation.
tThis is valid in many cases (to an accuracy of several percent). An exception is the triggering of chemical
chain reactions, intensive luminescence, phase transitions, etc. by radiation.
865
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The method is based on the following. If a dosimetric sample, at the instant t = 0, is placed in a constant
radiation field which is homogeneous within the boundaries of the sample, then, if shielding and side effects are
neglected, the temperature at the center of the sample will increase for a certain time r in accordance with a
linear law, independently of the temperature of the surrounding medium. This can be shown by solving the
equation of thermal conduction [1] for any sample of simpler shape with uniformly distributed heat sources. The
time r proves to be proportional to the square of the characteristic dimension of the sample d and inversely pro-
portional to the diffusivity X (this can also be obtained from dimensional analysis). The heating rate
dT
dt of a linear segment is determined only by the capacity of the heat sources (dose rate) and the specific heat
of the material of the dosimetric sample. The dose rate can be calculated from the.formula
P=0,417cCdT\ ~
d tt '
where c' is the specific heat of the substance
(cal/g deg); KdT~
0
is measured in deg/hr and P in Mrad/hr.
In the practical application of this method, it is necessary to decide on the material and size of the sample,
the method of measuring the temperature at the center of the sample, and the design of the entire apparatus.
The method was applied to the determination of the dose rate in vertical experimental channels of a reactor
operating on ordinary water and enriched uranium M.
'p > 2 z 3
t min
Fig. 1. Time variation of the temperature at
the center of the dosimetric sample (poly-
ethylene) in the reactor channel (dT/dt) _
= 420 deg/hr; r = 2.5 min.
With regard to choice of material for the dosimet-
Fig. 2. Dosimetric apparatus. 1) Sample;
2) thermal junction of thermocouple; 3)
aluminum foil; 4) flexible support- insu-
lator; 5) cold junction of thermocouple;
6) leads to the measuring instrument.
ric samples, they were chosen to simulate (in chemical composition) the substances subject to irradiation. In our
dosimetric experiments we used polystyrene, polyethylene, slicone rubber, teflon, fused quartz, and others.
The basic samples simulating substances containing hydrogen are polystyrene and polyethylene; for
simulating glass (apart from that containing borium), we used quartz. All three substances are convenient,
owing to their high radiation stability and small activation.
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The samples consisted of cylinders 30 mm in diameter and 50 mm in height. Such a sample size is
sufficiently large so that the time constant r was 2-3 min, which is necessary for a reliable measurement of
the heating rate (Fig. 1). At the same time, for such dimensions, the shielding of one part of the sample by
another can still be neglected, and the dose field could be investigated over the channel height and diameter.
For the measurement of the temperature, we used a copper-constantan thermocouple-introduced between
two halves of the cylinder, which was split vertically in two. Special control experiments. indicated that for a
sufficiently small wire thickness (0.15 mm), the temperature at the thermal junction did not differ, in practice,
from the temperature of the sample. The thermocouple's emf was measured on a portable potentiometer (for
dose rates, of 30-200 Mrad/hr) or on a galvanometer (for dose rates of 0.5-30 Mrad/hr). In the latter case, an
auxiliary scale was used. Telemetric measurements were made.
The design of the dosimetric apparatus is extremely simple, and is shown in Fig. 2. Except for the thermo-
couples, all parts were made from nonactive materials. Sometimes, instead of a flexible hose, we used a flexible
aluminum tube. When the dosimetric sample was a liquid, it was placed in a thin-walled container (for example,
one made from polystyrene) and an additional cylindrical shield was introduced to prevent convective mixing.
In order to increase the time T for samples of high thermal conductivity, a thermal-insulating jacket was pro-
vided, which somewhat increased the size of the apparatus . When it was necessary to measure the dose in the
channel filled with water, the entire dosimetric sample was placed in a thin, water-tight case. The position of
the sample in the channel was accurately established by means of centering and measuring instruments.
The following procedure was used for the measurements. Before the measurement, the sample was con-
ditioned in the channel outside the irradiation zone, and then lowered to the proper point. The temperature was
read for three to six minutes, after which the sample was again raised. We determined the dose rate in this way
for various substances in a number of the channels of the VVR reactor. The error in the measurements was 5-101o.'
As an example, Fig. 3 gives the curves for the dose-rate
distribution over the height along the axis of one of the
channels of the VVR reactor for two materials, poly-
ethylene and quartz glass. The difference in the size
9
2
40 60 B0
Dose rate, Mrad/hr
Fig. 3. Distribution of the absorbed-dose rate along the
height of channel "65" for polyethylene (1) and quartz
(2). The arrow marks the bottom of the channel. The
ordinate 480 corresponds to the center of the active zone.
The reactor power is 2000 kw.
T, ?C
70r-
10 0
3 4
t, min
Fig. 4. Heating curve of polytetro-
fluoroethylene under irradiation.
As a rule, we used the tabulated values of the specific heat of the materials; in some cases we made additional
measurements of the specific heat.
867
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of the dose between these two substances results from the fact that the former receives a dose not only from y
rays, but also from the moderating of fast neutrons, while the latter receives a dose almost exclusively from y
rays. The values of the dose for other substances containing smaller amounts of hydrogen than polyethylene lie
between curves 1 and 2 of Fig. 3.
When the character of the neutron spectrum in the irradiation zone was known, even if only approximately,
it was possible to divide the absorbed energy into parts resulting from the absorption of y radiation and from the
slowing down of fast neutrons. To do this, one can use the method of variation of the sample composition (mainly
the hydrogen content) [3]. We carried out such a separation for a number of substances.
In conclusion, it should be noted that the nonstationary calorimetric method makes it possible to discover
certain additional phenomena caused by irradiation if it is accompanied by considerable thermal effects. Thus;
for example, with the use of polytetrofluoroethylene as a dosimetric sample, we observed on the heating curve
an abrupt change in slope corresponding to a sharp acceleration in the heating of the sample (Fig. 4). It was
suggested that this was the result of additional crystallization of the material facilitated by the destruction of
the macromolecules under irradiation. This hypothesis was confirmed by. measurements carried out by us, by
the authors of [4], and by' calculation of the thermal effect of the phase transition.
It follows from the above that the nonstationary calorimetric method can be suitable for determining the
absolute energy dose in studies connected with the irradiation of organic substances, polymers, water and aqueous
solutions, glass, biological objects, etc. in reactors and by powerful y sources and x-ray tubes.
LITERATURE CITED
1. A. N. Tikhonov and A. A. Samarskii, Equations of Mathematical Physics [in Russian] (Moscow, State
Technical and Theoretical Press, 1953).
2. Yu. G. Nikolaev, Reactor Design and Reactor Theory [in Russian], Reports of the Soviet Delegation to the
International Conference on the Peaceful Uses of Atomic Energy (Geneva, 1955) [in Russian] (Moscow,
Acad. Sci. USSR Press, 1955) p. 265.
3. D. M. Richardson, A. O. Allen, and G. W. Boyle, Dosimetry of Ionizing Radiation, Reports of foreign
scientists at the International Conference on the Peaceful Uses of Atomic Energy (Geneva, 1955) [Russian
translation] (Moscow, Acad. Sci. USSR Press, 1955) p. 265.
4. A. Nishioka,M. Tajima, and M. Owaki, J. Polymer Sci. 28, 118, 617 (1958).
THE DOSAGE OF OUTDOOR y RADIATION FROM RADIOACTIVE
V. P. Shvedov, G. V. Yakovleva, and M. I. Zhilkina
Translated from Atomnaya Energiya, Vol. 9, No. 4, pp. 323-324, October, 1960
Original article submitted March 15, 1960
Radioactive fallout, gathered monthly during 1958-1959 in the vicinity of Leningrad, was subjected to
analysis on a scintillation y -ray spectrometer. The method of investigation was similar to the method used in
[1]. The figure presents the results of the y -spectrometric analysis of radioactive fallout. Two factors affect
the amount of radioactive fallout: contamination level of the troposphere by radioactive fission products and
the amount of precipitation. The effect of contamination level of the troposphere must be regarded as the main
factor. In fact, despite relatively slight fluctuations in the quantity of precipitation, the amount of radioactive
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fallout during the period of greatest contamination of the troposphere has characteristic, well-defined maxima
as the result of recent nuclear weapons tests or the magnified interchange of the radioactive masses of the
stratosphere and troposphere. The maximum fallout in April - June of 1959 was caused precisely by the effect
of such an interchange between masses of the stratosphere and troposphere. Despite heavy rains, the fallout in
the second half of 1959 had a very low activity due to the absence of any appreciable contamination of the
troposphere.
15107
0 5.107
6.107
14. 0 2.107
9
c 141
10107
z 2107
> 6000
2000
U 1.23455789f0fff~
? =L 1958 .
12 2345678 fflff12
34567
1959
30-Year Dosage of Outdoor y Radiation from
Radioactive Fallout in 1959
Isotope Dosage, mr
Nb95
1,34
(Zr+ Nb)"5
5,26
]IIt103
0,49
Rh'?
1,76
00,11
0,0'1
Cs137
10,25
Total ........
19,2
The data given in the figure provides us with
the opportunity of determining the 30-year dosage
of outdoor y radiation from radioactive fallout -
during 1959. The method of analysis was similar
to the method used in [1]. The results of this deter-
mination are shown in the table. -
The 30-year dosage of fallout from 1954 to
1956, 1957, and 1958, according to [1,72], comes
to 16, 18, and 40.1 mr, respectively. Hence, it is
clear that the 30-year dosage caused by fallout in
1959 will be considerably less than the fallout
dosage for 1958. This fact is due to the absence
of nuclear explosions in the year 1959. -
LITERATURE CITED
1. V. P. Shvedov, G. V. Yakovleva, M. I. Zhilkina, and T. P. Makarova, Atomnaya Energiya 7, 6, 544 (1959).`
2. L. I. Gedeonov, V. P. Shvedov, and G. V. Yakovleva, Atomnaya Energiya 7, 6, 545 (1959).'
* Original Russian pagination. See CB translation.
?Q''f0?f07
R!L 103
RU1os
C,S 137
21'95
-
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THE INCREASE IN RADIOACTIVE FALLOUT
IN GRADETS KRALOVE (CZECHOSLOVAKIA) AS THE
CONSEQUENCE OF NUCLEAR TESTS IN SAKHAR'
V. Santgolzer
Physics Department, Medical Faculty, Karlov University, Gradets Kralove,
Czechoslovakia
Translated from Atomnaya Energiya, Vol. 9, No. 4, pp. 324,326, October, 1960
Original article submitted May 17, 1960
As the result of a temporary cessation of nuclear testing from November 1 of 1958 until the end of 1959,
a considerable reduction in the amount of radioactive fallout was observed [1]. In October of 1959, the average
daily activity was 0.07 p C/km2 . In November it dropped to 0.03 ? C/km2, and this level was maintained
until the end of February, 1960. The increase in fallout activity due to rains in these months was relatively
mild (maximum of 0.06 ? C/km2 per day). It may be assumed that during this period, tropospheric fallout
was curtailed almost altogether and only the stratospheric variety remained, its source being the stratospheric
reservoir [2].
As a result of the first nuclear test in Sakhar (February 13, 1960), an increase in fallout activity was
observed in Czechoslovakia from March 1 to 21, 1960.
The results of systematic measurements of rainfall activity from November 1, 1956 until March 31, 1960
are shown in Fig. 1 (in m pC/liter). The strongest radioactivity of rain was observed on March 1, 1960 and was
25.9 in p C/liter. The drop in this activity with passage of time is evidence that it originated on February 13,
1960:
As shown in Fig. 1, even before June of 1959, radioactive rains were noted. The strongest radioactive
rain during this time was noted on April 6, 1959, i.e., half a year after the temporary curtailment of nuclear
testing. The rain activity on March 1, 1960 was 2.5 times greater than on April 6.+ One reason for this appears
to be that the distance from Czechoslovakia to the test site in Sakhar (4000 km) is less in comparison with the
distance from the Nevada desert (11,000 km). Prior to the nuclear tests in Sakhar, Nevada was considered the
prime source of radioactive aerosols for Central Europe [3].
The activity that accumulated during March of 1960 was 23.38 mC/km2, as contrasted with an activity
of 0.91 mC/km2 in February and 0.55 mC/km2 in January. The accumulated radioactivity was determined
by measurements of fallout samples for one month, gathered in a special container, and by calculation of the
daily quantities of radioactivity with allowance for the drop in its activity.
Figure 2 shows the activity of daily fallout (in mC/km2) from February 1 until April 28, 1960. In
comparison with the mean fallout of 0.03 mC/km2 per day that was observed in February, the fallout by
March 1 had stepped up to 17.65 mC/km2 per day, i.e., almost 600 times. After three weeks, the quantity
of radioactive fallout diminished and stabilized to an average level of 0.04 mC/km2 per day. The source
of this fallout is also the stratospheric reservoir. About a week later this situation was once again upset by a
second nuclear test in Sakhar, taking place on April 1, 1960. An increase in activity was noted in a specimen
taken on April 9, when the activity was 0.70 mC/km2 per day, i.e., approximately 20 times greater than
the reading for the end of March and beginning of April, 1960.
* The paper was written for our journal.
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C)
1-~
~JIY.O~:144-YS3r ~\
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Fig.. 2. Increase in radioactive fallout
activity in March and April, 1960.
29/11 31/111 30/IV
Date
Extrapolation of the activity decay curves for the samples
gathered at the beginning of March and in April confirms the
dates of nuclear testing on February 13 and April 1, 1960.
The behavior of 7the activity curves can be described by the
expression A = at - n, where n =1, 2.
It should be pointed out that the activity decay for rain
samples followed the law A = at - n, where n = 1, 2, whereas
for active dry fallout n,= 1.3.
V. Santgol'zer, Atomnaya Energiya 7, 5, 480 (1959)* ;
9, 1, 60 (1960)*.
As a standard preparation for measuring the activity, we
adopted Sr90 + Y90.
We extend our gratitude to Academician F. Begounek
for so kindly furnishing us with this preparation and for all the
counsel he gave us in this work. We express our appreciation
also to I. Matsek for designing and building the automatic
apparatus for measuring the radioactivity with a record of the
number of pulses.
V. P. Shvedov and L. I. Gedeonov, Collection: Soviet
Scientists on the Danger of Nuclear Weapons Testing
[in Russian] (Atomizdat, Moscow, 1959) p. 45.
3. M Hinzpeter, F. Becker, and H. Refferscheid, Atomtechnisches Aerosol and atmospharische Radioaktivitat
(Strahlenschutz Vol. 7) " Braunschweig, 1959.
* Original Russian pagination. See CB translation.
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NEWS OF SCIENCE AND TECHNOLOGY .
THE PM-2A NUCLEAR PACKAGE POWER FACILITY
Development plans in power reactor design in the USA have been concentrated in recent years more and
more on low-power reactors, i.e., reactors producing electric power up to 10 Mw. The results of the development.
plan heading in this direction include a project of a transportable packaged power plant named PM-24, designed
for installation in Greenland [2,_.3].
The PM-2A system utilizes a pressurized-water reactor designed by the Alco Products firm. In addition
to generation of electric power, the plant will provide steam for space heating.
The basic data of the PM- 2A facility are the following:
Reactor thermal power, Mw .
10
Minimum energy output for a single loading
of core, Mw/yr
8
Available electric power, kw
1560
Amount of heat made available for space
heating, kcal/hr
250,000
Operating pressure, atmos
123
Coolant temperature, ? C:
at reactor inlet
260
at reactor exit
270
-Number of fuel assemblies
32
Core:
diameter, cm '51.3
height, cm $5.2
composition:
U =, kg 19.5
B10 g 1.7
stainless steel, kg
172
H20, kg
91.5
Number of fuel elements in each assembly
18
Number of control rod bundles
5
Steam generator:
operating pressure, atmos
38
steam temperature, ? C
240
The fuel elements contain enriched uranium oxide dispersed in stainless steel.
The entire PM- 24 package plant will be housed in a tunnel dug out in the ice or snow (Fig. 1). It will be
delivered to the assembly site in 27 package shipments totaling 300 tons in weight, the heaviest shipment weighing
13.6 tons. A 10-girder support will facilitate assembly and servicing of the entire installation; the reactor and
steam generating unit will be resting on the first girder or supporting skid, while the equipment for feedwater
heating will rest on the second. The third skid will support the heat exchanger, and the next three will accomodate
air blowers. The turbogenerator unit will rest on the seventh skid, switchgear on the eighth, motors on the ninth,
and the steam condenser on the last. Each of the skids consists of two H-frames linked by crossties and matting
(Fig. 2). The rigid girder-skids on which the operating units will be mounted provide an adequate substitute for
a rigid foundation.
It takes a team of 36 engineers and technicians six weeks to assemble the installation on the operating site.
Seventeen men are sufficient for running the plant.
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The PM-2A is the third packaged power plant designed by Alco Products. Its first job was the SM-1, now ,
known under the designation APPR-1 [41, which was built in 1957 at Fort Belvoir (Virginia, USA); the second was
the SM-1A, now known as APPR-1A, now under construction at Fort Keeley (Alaska). It has double the power
rating of the SM-1 for the same core dimensions. The experience accumulated by Alco Products has paid off in
plant weight reduction from 2,500 tons (SM-1 and SM-1A) to 300 tons (PM-2A), with the assembly and run-in
and testing time telescoped from 18 to 3 months.
The reactors used in all three plants feature the same design, but the heat removal systems are of different
design: the PM- 2A uses air and ethylene glycol to cool the condenser, while the SM-1 and SM-lA use water for
the same purpose.
The experience acquired in operating the SM-1 facility was fully utilized in the design of the PM-2A; the
only additional feature incorporated was a study of the distribution of coolant flow through the core. To compute
the physical, thermal, and transient processes, the problem was run on a computer. The core of the PM- 2A
reactor uses eight less fuel assemblies than the SM-1 to achieve the same thermal power output. This means
26 cm less in the girth of the reactor pressure vessel.
Fig. 1. Driving and lining the reactor tunnel, in Fig. 2. A support skid with the reactor and steam
Greenland. A similar tunnel will be prepared for generator in place.
the PM- 2A facility.
The cost of the entire PM-2A plant, including the assembly job, is 4.1 million dollars, and operational
costs are expected to run in the neighborhood of 800 thousand dollars yearly. By way of comparison, a diesel
power plant yielding the same power output costs 400 thousand dollars, but the operational costs (not counting
fuel transportation costs) are 1.1 million dollars yearly. The nuclear power plant displays an indisputable
superiority over its diesel competitor in this application.
1.
Atomwirtschaft 5, 5, 242 (1960).
2.
Nuclear Energy 14, 145, 257 (1960).
3.
Nucleonics 18, 6, 120 (1960).
4.
Atomnaya Energiya 2, 6, 574 (1957):`
* Original Russian pagination. See CB translation.
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BERYLLIUM
(Present Status of Beryllium Technology and Research)
N. Mironov and S. Kostogarov
Beryllium has a small thermal absorption cross section and high moderating power. It is therefore found
very useful as material for reflectors, displacement rods, and fuel element jackets (in gas-cooled reactors).
Beryllium boasts among its exclusive advantages: high modulus of elasticity ranging'(3.1-3.2),104 kg/cm2,
high specific heat and thermal conductivity (when heat-absorbing structures are used), and high ratio of tensile
strength to specific weight [1].
However, beryllium's low plasticity at room temperature is a hindrance to its use on a broad scale. The
defect in its plastic behavior, affecting beryllium of even the highest grade purity, is explained either by the
unfavorable geometry of the metal's hexagonal close-packed lattice which impedes slip flow, or by the possible
effect of trace impurities. If it were possible to produce the metal free from any impurities, it would obviously
display a cert?in degree of plasticity, as is the case with impurity- free titanium and zirconium, which have
hexagonal lattices with constants close to those of beryllium.
Production of higher-purity beryllium is, therefore, a task meriting close attention. The work is proceeding
in the following directions:
1) superhigh-vacuum distillation (residual pressure down to 10-T to 10-9 mm Hg) in a vacuum furnace with
high-frequency heating [2] and a condenser in the form of a flat rotatable plate. The vacuum system is enclosed
in a chamber filled with purified argon. To avoid contamination of the metal by carbon, the system is outgassed
by a mercury diffusion pump, and provided with a liquid-nitrogen cold trap. The chamber walls are outgassed by
means of a coronal discharge;
2) zone melting in vacuum (residual pressure down to 10-5 mm Hg) or in the protective atmosphere of an
induction furnace, vertically in a suspended state [3];
3) reduction of beryllium halides by gaseous alkali metals at a temperature of approximately 1000 ? C.
Silver is used for the lines bringing in beryllium chloride into the system; the reaction chamber is made of
molybdenum [4].
The purest beryllium metal was made [4] from high-purity beryllia converted to beryllium chloride; the
metal was further purified by subsequent distillation. Beryllium buttons obtained by electrolysis of beryllium
chloride had the same degree of purity prior to vacuum remelting as found in ordinary beryllium prior to re-
melting.
The electrolytic process results in beryllium much purer with respect to metallic impurities than that
obtained by the magnesium thermal-reduction process, but the electrolytic metal contains a high quantity of
sodium and chloride [5], which is diminished in subsequent vacuum remelting.
The plasticity of the most brittle cast beryllium at room temperature is practically, null; beryllium
obtained by powder metallurgy techniques shows somewhat greater plasticity; cast beryllium has been virtually
completely displaced by metal fabricated by powder metallurgical techniques, since 1950.
The beryllium pebbles are remelted in a vacuum, and the resulting ingots are then converted to beryllium
flakes. The flakes are comminuted in beryllium- lined pulverizers in a nitrogen or helium atmosphere and passed
through a 200 mesh sieve. In a more sophisticated method for fabricating beryllium powder [61, process control
is much tighter and the flakes are pulverized in high-speed centrifugal mills,and the powder charge is inserted
into dies with all contact with air avoided. The standard powder has the following granulometric make-up [4]s
12% > 100 p; 15% for both 80-100 p and 60-80 p ; 22% for 40-60 p ; 14% for 30-40 p 10% for
20-30 p ; about 12% for < 20 p (beryllium oxide content ranging from 0.5 to 1.5%). The finer powders
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are more liable to contamination, by oxygen for the most part. Oxygen content may be minimized by carrying
out the pulverization step in dry helium at a temperature of - 64? C.
Beryllium powder items are fabricated by cold pressing (specific pressure 8-25 tons/cm2) and subsequent
sintering in vacuum at a temperature of 1100-1200 ? C, hot pressing in vacuum (residual pressure of 10-1 to
10-2 mm Hg) at a specific pressure of 4-20 kg/cm2 and temperature of 1050-1100? C, warm and hot ex-
trusion of cold-pressed and hot-pressed items, rolling over a temperature range of 455-1100? C for items ob-
tained by various methods.
Vacuum hot pressing is also resorted to in industry. Beryllium ingots weighing as much as 147 kg and
measuring 100 by 610 by 1300 mm in size have been fabricated by such methods, while cylindrical, conical,
and spherical ingots surpassing 1220 mm in diameter and weighing up to 910 kg are also a matter of record [7].
Recent years have seen more and more frequent recourse to the extrusion of small ingots (up to 60 mm)
without cladding at temperatures of 400-450? C (known as warm extrusion) [7]. In this case, the reduction was
4:1, the specific pressure ranged from 6.5 to 16.0 tons/cm2, the extrusion speed seldom exceeded 2.5 meter/hr
[5]. The lubricant used was graphite backed up by molybdenum disulfide [5, 7].
Hot extrusion is carried out in the 900-1150 ? C range in mild steel jackets covered with copper foil. The
specific pressure at 1150 ? C temperature is 2.2 tons/cm2. 40 :1 reduction is used. Hot extrusion imparts optimum
mechanical properties to the metal [5].
A 'bare' extrusion practiced in an intermediate temperature range (730-820? C) is also resorted to. The
beryllium in this case is covered by a glass or salt lubricant. The reduction used is 9 :1.
Extrusion involving the use of a floating core [5] is employed to fabricate tubing of diameter ranging from
63.5 (at a wall thickness of 19 mm) to 12.7 mm (at a wall thickness of 1.5 mm). According to recent data [8],
tubes of wall thickness down to 0.25 mm have been fabricated for the AGR reactor, and the tubes are capable of
taking a radial flexure of 125 mm in the cold, as well as tubing 8 mm in diameter and 610-680 mm long at a
wall thickness of 0.1 mm.
There are some reports in the literature on the use of silver cladding as a lubricant in extrusion, and as pro-
tection against oxidation [8].
One report [9] provides a description of hydrostatic pressing of beryllium powder tubing at a temperature
of approximately 800 ? C and pressures up to 31 kg/mm2. The beryllium powder was charged into a space between
a segmental die of Nimonic-80 alloy and a concentrically placed thin-walled steel tube. After being warmed up,
the beryllium powder is compacted under radially applied compression (transmitted via an extensible steel tube
through molten lead).
Rolling of beryllium on an industrial scale [7] (rolled sheet up to 750 by 1520 mm) is carried out over the
range of recrystallization temperatures (455-845 ? C) [5, 7]. To avert a rupture in the direction of the base plane,
to minimize oxidation and level out the temperature distribution, the metal is inserted,within heavy steel jackets.
After rolling to a thickness, of 3.2 mm, the jackets are removed, and the metal sheets, baled into packs, are given
an additional rolling.
Reports are available [10] on beryllium rolling studies carried out at temperatures 1040 ? and 870 ? C. At
1040 ? C, the reduction of extruded specimens was 5 :1 to 15 :1 in transverse rolling (reduction including ex-
trusion to 108:1). The sheet metal obtained showed an elongation of about 30-40%, tensile strength of 40-45
kg/mm2, and true ultimate strength of 63-70 kg/mm2. Lowering of the rolling temperature to 870? C resulted
in a metal showing a wide spread in elongation and strength data. Temperature effect was negligible over the
870-1040 ? C for hot rolling i n two directions, and the properties of the beryllium transverse to and in the rolling
direction were about the same. Sheet rolled with 14: 1 reduction (3.75 :1 in each direction) showed approxi-
mately 25% elongation, tensile strength of 45-49 kg/mm2 with a true ultimate strength of 60-65 kg/mm2. The
absence of plasticity in the third direction in rolling imposes important limitations on the material, from an
engineering standpoint. The property of two-dimensional plasticity is unsuitable for combined stresses, and does
not favor the use of rolled beryllium as a really plastic material.
Beryllium is forged over the 730-845? C temperature range [7], but beryllium-forging technology.is still
at a rudimentary stage of development. It has been reported [11] that vacuum forging of beryllium enclosed
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in a crucible has been successfully carried out. It is assumed that this technology will lead to an isotropic metal
of much greater plasticity, and equal in this respect to hard aluminum alloys,while retaining the desirable pro-
perties inherent in beryllium.
A beryllium ingot 1572 mm in diameter and jacketed in steel was pressed to a diameter of 2030 mm and
a height of 76.2 mm at a temperature 1094? C, on a 50,000 ton press [12]. The quality of the ingot was
monitored by ultrasonic flaw detection techniques.
Rotary swaging of beryllium and beryllium wire drawing are rarely resorted to, but may be performed at
400 ? C on beryllium rods [5] extruded from a jacket.
A method has been developed for sintering beryllium blanks, in pressure working of beryllium powder
compacts to densities of 1.8 - 1.83 g/cm3, without applying pressure [6]. Powders of 200 mesh were comminuted
in an atmosphere of purified argon, and charged into molds. Sintering proceeded for 6 hours in vacuum induction
furnaces at temperatures of 1200-1220? C, in graphite molds. This method was also used on a laboratory scale to
produce beryllium cladding for uranium oxide fuel elements [6]. Among the drawbacks encountered in this
technique, we might mention the need for strict maintenance of the granulometric make-up of the powders, which
must be kept to a minimum of oxidation, and the impossibility of controlling sinter size.
A technique for argon vapor arc welding of beryllium has been developed to overcome the formation of
welding cracks, oxidation, and gas formation. The general disadvantage inherent in welding techniques, where a
liquid phase is present, is the coarse-grained structure of the cast metal seam and the deterioration in mechanical
properties. To reduce thermal shock, it is recommended to preheat the metal prior to welding to 540 ? C or
higher. In diffusion bonding of beryllium, the mechanical properties of the seam are almost indistinguishable
from the mechanical properties of the bulk metal. Diffusion bonding is possible over a broad temperature and
pressure range. Good results are had by welding for two hours at a temperature of 1200 ? C [5, 13].
Brazing of beryllium parts to beryllium or to other metals yields satisfactory results. The best brazing
alloy for work in an inert atmosphere is an alloy of aluminum and silicon (12%). For furnace-brazing parts of
simple geometry, aluminum and silver brazing alloys and aluminum-silver and silver-copper alloys are also
used. Highly encouraging results have been obtained from electron-beam welding in a high vacuum [5, 7].
Machining of beryllium is a difficult proposition in view of its brittleness and hardness, and also because
of the formation of toxic beryllium dust in the process. Equipment and tools for beryllium machining must be
kept in excellent condition and, as a rule, must be housed within suctioned dust- tight hoods.
The following operating conditions and parameters are recommended [7].
In lathe work, cutting speed is recommended at 50-85 meter/min, rate of speed not greater than 0.4
mm/rev, nose angle about 0% side clearance about 7 ?; attacking angle of the cutting-tool edge should be kept
within 7-10 ? and should not exceed 10 ? in any case (to avoid chipping); radius of curvature of the cutting edges
must be kept within 0.4-0.8 mm.
The recommended cutting speed for milling operations is 30-50 meters/ min, for rough milling and
50-70 meters/min for finish work, with rate of feed ranging from 75 to 150 mm/min.
The recommended drilling speed is 20-30 meters/min, with rate of feed 0.025-0.05 mm/rev (for drills
of 0 3.2 mm or less), 0.05 - 0.1 mm/rev (for drills of 0 6.35 mm), and 0.125 mm/rev (for drills of 0
12.7 mm); the angle of the drill-bit cutting faces should be 90?.
Cutting tolerances can be brought to 0.1 mm. A cutting speed of about 6 cmg/min is recommended.
Oxidation of beryllium exposed to air is negligible at temperatures below 600 ? C. The corrosion resistance
in water is not constant. Standard powder compacts conforming to USAEC regulations were subjected to corrosive
attack tests for 96 hours at 250 ? C [5].
Beryllium presents a considerable health hazard, and this calls for strict attention to safety procedures in
handling the material. A new British beryllium mill (of 7 tons annual capacity, staffed by a crew of 200) is
being built in a single-storey building with continuous ventilation of warm filtered air provided for; 3000 air
samples will be tested weekly in an in-plant laboratory [8].
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USAEC rules which have remained in vigor since 1950 stipulate that the beryllium content in air is not to
exceed 2 pg/ms. Beryllium content in the air outside the building concerned should not exceed 0.01 p g/ms.
New methods for treating of beryllium poisoning are being sought, to supplement the usual work safety rules.
Reports have come out on positive results in the introduction of HAu (COD)3 into the organisms of experimental
animals. This substance inactivates the beryllium present in the organism [41.
Because of the difficulties encountered in its handling, beryllium products are still quite expensive. The
price of sheet rolled from extruded products is $ 250 per kg [4]. The cost of complex shaped products, e.g.,
thin-walled tubing (0.25 annulus) has gone up to 400 ? sterling [8].
LITERATURE CITED
1. Metal Progr. 74, 4, 96 (1958).
2. A, Martin, Vacuum VII and VIII, 38 (1957-1958).
3. G. Ellis, Metallurgia 58, 348, 349, 350 (1958).
4. H. Weik, Metall. 3, . 202 (1959).
5. Y. Williams, Metallurgical Reviews 3, 9, 1 (1958).
6. T. Barrett, G. Ellis, and R. Knight, (Geneva, 1958) Conference on Peaceful Uses of Atomic Energy, P/320.
7. K. Wikle, Metal. Ind. 93, 26, 529 (1958).
8. Nuclear Engn. 5, 44, 31 (1960)..
9. A. Blayney, Metal. Progr. 74, 3, 95, 184 (1958).
10. J. Greenspan, Trans. of the Metallurg. Society of ASME, 215 (February, 1959).
11. J. Metals 11, . 5, 307 (1959).
12. Metal Progr. 76, 3, 128 (1959).
13. N. Weare and It. Monroe, Light Metal Age (August, 1959) p. 10.
THE SECOND AZERBAIDJAN REPUBLIC-WIDE CONFERENCE
ON THE USES OF RADIOACTIVE ISOTOPES AND
NUCLEAR RADIATIONS
At the Second Republic- wide Conference on the Uses of Radioactive Isotopes and Nuclear Radiations held
at Baku during March, 1960, 33 papers were presented by various scientific institutions of the Academy of Sciences
of the Azerbaidjan SSR, the Council of the National Economy for the Republic, the Health Ministry of the
Azerbaidjan SSR, and also by the Academy of Agricultural Sciences of the Azerbaidjan SSR and the N. Narimanov
Azerbaidjan State Medical Institute.
The reports presented shed light on some of the results obtained from problems resolved in biology, medicine,
agriculture, particularly the petroleum industry, on the study of properties of materials (semiconductors), and other
fields.
A discussion of the papers showed that during the period stretching from the first conference (in 1957) to the
present one, the number of local scientific research, industrial, agricultural, and other organizations which have
used radioactive isotopes and nuclear radiations to advantage for scientific and practical aims has increased
measurably.
The use of radioactive isotopes, gamma and neutron sources, has resulted in the successful solution of such
problems as the effect of irradiation of seeds prior to sowing on the growth, development, and crop yield of cotton,
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the effect of prolonged exposure to a radioactive factor on the state of blood and on several biochemical indices,
the effect of phosphorus isotopes on the nature and spread of aseptic inflammation, determination of saturated vapor
pressure, the behavior of impurities in the distillation of semiconductors, determination of the diffusion coefficient,
determination of water content in petroleum blends, circulation of catalyzer in industrial systems, shielding pro-
perties of concrete, etc. Radioactive tracer methods have resulted in notable economic savings in several branches
of industry.
The ?Azneftegeofizika? (Azerbaidjan geophysical oil exploration) trust has made extensive use of radio-
active methods in geophysical drill hole studies.
In order to discover aquifers and gas-bearing strata in cases where electric logging methods fail to yield
the desired results, and in old boreholes, successful use has been made of gamma-gamma and neutron-gamma
logging techniques. Radioactive isotopes have also been used to determine the position of thief zones, break
points of casing strings, flow patterns of water outside the casing string in production oil wells. For a quantitative `
estimate of the distribution. of injection fluid in a well zone, a method of ascertaining the injectivity profile
with the aid of an activated slurry is used. A method of quality control of drill-hole cement completions which
presents a picture of the distribution of cement over vertical and horizontal sections of the wells has been de-
veloped.
New methods for holding back the oil-water contact surface in the case of completed wells have recently
been developed and put into practice. A new two-channel set of equipment for radioactive logging designed for
deep-hole work has also been developed and introduced into field work.
A line of instruments to protect the safety of field personnel working with radioactive preparations has been
developed for insertion into the well -during checking of hydraulicking performance, cementing, and for locating
thief zones; a special field laboratory-on-wheels has been developed.
At the Petrochemical Process Institute of the Academy of Sciences of the Azerbaidjan SSR, plans have
been developed for studying the mechanism of transformation of hydrocarbons, the mechanism of the action of
oil additives, and for elaborating isotope control techniques. Labeled-atom techniques have been used in a
many-pronged attack on the problem of the catalytic cracking method for alkyl aromatic and paraffinic hydro-
carbons usually encountered in crude petroleum. The reaction mechanisms of primary and secondary cracking
reactions have been studied. The active role of the alkyl aromatic side chain in the formation of coke has been
established. Data have been obtained which are of interest in making a correct choice of catalyst for cracking
various types of crude.
The labeled- atom tracer technique was applied to the study of the performance of oil additives AzNII-7
and TsIATIM-339. A new instrument was designed to study the qualities and properties of oil additives, and
their performance mechanisms; the instrument may be used either with or without labeled atoms. A comparative
estimate of the properties of additives AzNII- 7 and TsIATIM- 339 has been made. The kinetics of film formation,
solubility and stability of additives, anticorrosive properties, conditions of formation of a protective film, the
degree of participation of the individual components of the additive in the formation of a film layer, are among
the topics investigated.
Moreover, the Institute has elaborated a technique of accurate and rapid determination of water content
in oil, and a method of exactly estimating the rate of circulation of alum-silica catalyst in the catalytic cracking
process using a catalyst pellet labeled with Co".
Facilities for studying saturation vapor pressures have been built in the isotopes laboratory of the Physics
Institute of the Academy of Sciences of the Azerbaidjan SSR. These facilities are highly accurate in the
measurement of vapor pressure, and use radioactive isotopes. The laboratory has carried out measurements of
the vapor pressure at saturation of selenides, thallium sulfate, and selenium vapors. It has been found that T1s.Ses
decomposes. upon vaporization from the solid phase, and is converted to TI2Se.. Experiments were conducted with
the proper isotope to investigate the behavior of several impurities in the vacuum distillation of selenium. It
was found that mercury impurities present in the vacuum distillation of selenium volatilize simultaneously with
the selenium. The isotope Se7' was used to study the effect of bismuth impurities on self-diffusion in selenium;
The finding was that the activation energy of self-diffusion is dependent on the content of bismuth impurity.
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At the Institute of Soil Science and Agrochemistry of the Academy of Sciences of the Azerbaidjan SSR,
experiments were conducted with radioactive phosphorus to study P32 uptake in the cotton plant and its effect
on the yield of raw cotton both under field conditions and ,greenhouse conditions (1958-1959). On the basis of
these investigations, conclusions were drawn that manuring of the soil with superphosphate fertilizer (dosage of
2.5 and 5 millicuries per square meter) was accompanied by accumulation of the greatest quantity of P32 in the
leaves, with the least propensity to accumulation shown in the stalks and roots of the plant (the cotton yield
increased 10-14%); and that when the plants are sprayed with a radioactive phosphorus solution during tite budding
stage, flowering stage, and cotton boll formation stage (three times in all), the yield is increased by 9-18%.
In the Plant Physiology department of the Institute of Botany of the Academy of Sciences of the
Azerbaidjan SSR, the tracer method was used to study the effect of microfertilizers on the rate of photosynthesis
and the outflow of assimilates, as well as on the distribution and translocation of radiophosphorus in the white
mulberry.
In addition to these research efforts, the tracer isotope method has been applied to the study of the effect
of water supply irrigation and vegetation irrigation on the rate of photosynthesis in the mulberry, honey locust,
eldar pine, and the olive tree. These investigations showed that the rate of photosynthesis is measurably favored
in these species, along with increased outflow of assimilates, by optimized dosages of copper, zinc, and cobalt.
Excess cobalt dressing will have an adverse effect on the efficiency of photosynthesis, if performed during the
vegetation perio since this will have a toxic effect on the plants. The build-up, translocation, and distribution
of radiophosphorus show improvement in response to optimized dosage with these trace elements. Under drought
conditions, water supply irrigation techniques increase productivity in various species of trees by improving the
rate of photosynthesis.
At the Institute of Genetics and Selective Breeding of the Academy of Sciences of the Azerbaidjan SSR,
research has been pursued on a study of the effect of radioactive radiations on the variability of several strains
of agricultural plants raised in Azerbaidjan. The acceptable range of critical doses for various crops were arrived
at. Application of large doses (30,000-40,000 r) brought about pronounced changes in the external features of
several varieties of cotton and wheat plants. These changes were accompanied, in some cases, by chromosomal
rearrangements in the first mitoses of the rootlets.
In the physiology department of the Academy of Sciences of the Azerbaidjani SSR, the effect of phosphorus
isotope on the nature and spread pattern of aseptic inflammation was investigated. A radioactive phosphorus
isotope (P32) introduced 20 hours before and 2 hours following the onset of a burn trauma showed a considerable
effectiveness in modifying the spreading pattern of the inflammatory process. It was shown that the magnitude
and character of interoceptive exchange reflexes are appreciably altered by x-ray treatment. While no appreciable
shifts take place in the blood (at the sugar level) in irradiated animals, significant modifications were detected
in the character of the interoceptive exchange reflexes. This is of much practical importance.
During the Conference, an exhibit demonstrating the latest radioactive logging equipment was open to the
attending public.
DECONTAMINATING ENCLOSURE
G. N. Lokhanin and V. I. Sinitsyn
The widespread use of radioactive materials in the various branches of the national economy, science, and
medicine, has posed the task of developing special shielding equipment and appurtenances.
In working with radioactive materials, contamination of equipment, of various types of laboratory ware,
instruments, and other objects occurs, and decontamination is a task which is not always possible under ordinary
conditions.
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In order to create the conditions for cleaning laboratory ware from radioactive contaminations, special
washing hoods had to be developed and fabricated. The ShM washing and decontaminating enclosure is now
being manufactured to service Soviet industry.
The hood is designed to facilitate washing laboratory ware, instruments, and other equipment contaminated
by a-, 8 -, and y -active materials.
The hood (Fig. 1, 2) consists of three separate glove boxes interconnected by coupling flanges. The
length of the enclosure is 3580 mm, width 825 mm, height 2320 mm, weight (of the whole assembly) 860 kg.
The internal volume of each box is 0.4 m3.
Each box is a gastight frameless enclosure resting on a supporting base. The boxes are fitted with viewing
windows, removable fluorescent lamps, rubber gloves,. ventilation ducts (intake and exhaust filters, gate valves),
a liquid waste drain, liquid waste receptacle, hot and cold water taps, water pressure gages, etc.
Transfer compartments are built on the left and right hand sides of the glove- box line to facilitate placing
and removal of laboratory ware and instruments. The transfer chamber is provided with two access doors (an
inner and an outer one). The glove boxes are washed down and deactivated by spraying showers installed in each
box. The first box is also provided with leads for bringing special deactivating reagents (acids, lyes, etc.) into
play, while the second box has a cold water tap for rinsing laboratory ware.
Fig. 1. Decontaminating and washing enclosure, stainless steel fabrication (operator's
view): 1) base; 2) hot and cold water and gas valves, etc.; 3) rubber gloves; 4)
transfer chambers; 5) viewing windows; 6) body of enclosure; 7) water pressure
gauge; 8) gate valves; 9) exhaust filters; 10) intake filters.
All of the ducting is controlled by valves with the valve controls accessible in the front panel. The valve
panels are properly labeled to show the function of each valve. The piping arrangement is designed to facilitate
unencumbered drainage of the solutions.
A rectangular viewing. aperture is built into the enclosure for observation of the insides of each enclosure.
One pair of special rubber gloves is installed in the front of each enclosure.
Three baths each (two round and one rectangular) are placed in the first and second enclosures for washing
laboratory ware and instruments. The round bath has a capacity of 8 liters, 244 mm diameter, height 190 mm.
The rectangular bath has a capacity of 12 liters, height 190 mm, length 390 mm, width 190 mm. Each bath is
provided with a small metal basket to hold laboratory ware and instruments during washing. Electric heating is
used on the rectangular bath and one of the round baths.
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Washing of laboratory ware, instruments, and other objects is attended to in the baths of the first enclosure,
using deactivating solutions. After washing, the laboratory ware and instruments are passed through the passage
Fig. 2. Stainless steel decontaminating and washing
enclosure (view from maintenance access side): 1)
transfer compartment; 2) door for placing and re-
moval of equipment and instruments, and for access
to inside of glove box; 3) air ducts; 4) rubber hose
connecting drain with liquid wastes receptacle; 5)
10KZhO liquid wastes collector, 10 liter capacity.
Fig. 3. 10KZhO liquid wastes receptacle:
1) welded gas-tight cylindrical tank; 2)
grasping handles; 3) discharge connection;
4) filling and blowdown connection; 5)
connection for annunciator device; 6)
filter.
from the first enclosure into the second, where the objects are rinsed in water baths. On leaving the second
enclosure, the laboratory ware is admitted into the third enclosure, where it is dried off. The degree of radio-
active contamination of the laboratory ware is checked with a 'Tiss' dosimeter following the drying step. Clean
laboratory ware leaves the third enclosure via the transfer chamber at the end of the enclosure line.
To provide removal of contaminated air, the enclosure assembly is outfitted with intake and exhaust
ventilation facilities (25 volumes of enclosed air are removed every hour). A rarefaction of 20 mm H2O is
maintained inside each separate glove box; the performance of the ventilation system may be monitored by
consulting a TIM-890 water pressure gauge with a 0 ? 125 mm H,ZO reading scale. The gate valves are adjusted
to control the rarefaction within each glove box separately. Air removed from the glove-box line is subjected to
decontamination in a two-stage exhaust filter system prior to being discharged to the stacks; clean air is admitted
to the chambers through an intake filter.
The intake and exhaust filters are similarly designed and of similar size. The first and second stage of
each filter are included in a common housing and are mounted on the top of each enclosure. The filter system
presents a surface of 0.11 m2. The first filter stage is filled with fiberglass to a layer thickness of 100 mm. The
first and second filter stages are equal in cross section. The second filter stage is filled with the high-efficiency
FPP filter fabric.
Depending on the extent to which the filters are contaminated, which may be established with a dosimeter,
the filters must be replaced at some point. An air duct is positioned on the rear of each enclosure to facilitate
removal of contaminated air.
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Maintenance access doors 600 X 500 mm are present
in the rear of each glove box, to facilitate maintenance
work and for placing of some types of material to be
cleansed.
The removal of the li uid wastes formed furing the
washing operation takes place through the drainage system
linkec up to each chamber, or directly into the 10-liter-
capacity 10KZhO liquid wastes receptable (Fig. 3). Th
receptable measures 206 mm in diameter, stands 415 mm
high, and weighs 8 kg ; the filter is made from FPP fabric.
A 10-liter-capacity KTO solids wastes receptable
is provided for the accumulation of solids (Fig.. 4) such as
broken laboratory ware, inaccurate instruments, rags, etc.
A ten-liter receptable mounted on a 2T dolly may also
be used to hold radioactive waste solids.
Fig. 4. KTO solid wastes container: 1)
foot pedal for raising and lowering
container lid; 2) welded cylindrical
receptacle body; 3) clamps for leak-
tight sealing of lid (hold-down clamps);
4) lid.
NEW LEAK-TIGHT GLOVE BOXES FOR HANDLING
ALPHA- AND BETA-EMITTING MATERIALS
G. N. Lokhanin and V. I. Sinitsyn
The use of radioactive materials in the various branches of science and the national economy has necessitated
research and development work on special-purpose shielding equipment and accessories.
The 1KNZh model glove box, designed for a single operator, has been developed by Soviet industry and is
being put on a mass fabrication basis, to service handling of alpha- and beta-active materials (Fig. 1, 2). The
box is built for work at high temperature and humidity levels in a corrosive alkaline or acid environment. The
box provides adequate shielding of nearby workers from alphas and betas, and prevents contamination of the air
in surrounding rooms by aerosols and gases or radioactive and toxic materials.
The dimensions of this glove box model are: height 2320 mm, length with one transfer chamber 1270 mm,
width 875 mm. All leads coupled into the glove box enclosure are sealed (with acid-resistant soft rubber packing)
.and held fast with adhesive. The frameless body of the glove box is welded with stainless steel up to 3 mm thick;
the glove-box tables are also welded stainless, to 10 mm thickness. The outer surface of the box is given a prime
coating after cleaning from grime and scale, and is then finished with a cream-colored acid-proof enamel. The
inner surface of the box frame has a smooth streamlined surface. The leak-tight volume of the box comprises
0.4 m3. The support base (see Fig. 1) for the 1KNZh box is welded carbon steel.
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The 1KNZh glove box is made with either one or two transfer chambers. The function of the transfer
chambers is to allow insertion of radioactive materials, withdrawal of finished products, and introduction of
clean equipment and other accessories and instruments required inside the box for work. Transfer compartments
come with two ports. The outer port (see Fig. 2) is for introducing radioactive materials, laboratory ware,
instruments, etc., from the operator's side or the maintenance zone into the transfer chamber. The inner port
serves for admitting radioactive materials and other objects inside the box proper. The inner port has a sector
which opens up simultaneously with the port. Three-lamp fluorescent fixtures, SDS-45 at 45 watts power, are
installed at the top of the box for illumination.
A rectangular viewing window is built into the front of the glove box to facilitate observation of the work
(Fig. 1).
A special ventilation arrangement is provided in all rooms where radioactive materials are handled in the
open, to protect the air environment of occupied rooms and the atmosphere from contamination by radioactive
aerosols.
Radioactive aerosols and other harmful substances and fumes are removed by means of a system of ven-
tilating air ducts (Fig. 2). The air removed from the glove boxes is cleaned prior to venting by passing through
a two-stage exhaust filter system. Air entering the glove box passes through a two-stage intake filter.
Fig. 1. The 1KNZh glove box (operator's view):
1) glove box frame; 2) support; 3) hot and cold
water and gas valves, etc.; 4) transfer compart-
ment; 5) rubber gloves; 6) viewing window; 7)
gate valve; 8) water pressure gauge; 9) intake
filter; 10) exhaust filter.
Fig. 2. The 1KNZh glove box (view from mainten-
ance access side): 1) air duct; 2) transfer-chamber
port; 3) liquid wastes drain; 4) liquid wastes receptacle;
5) dolly with solid wastes receptacle; 6) solid wastes
receptacle; 7) access door to box interior for placing
and removing equipment, materials, and instrunlients,
and for maintenance and other work inside the box
proper; 8) luminaire.
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Fig, 3 Solids receptacle on a
2T dolly: 1) dolly;' 2) ten
liter-capacity solids container;
3) lid; 4) meter-lehgth
steering handle. - .
Fig. 4. 'Two-place stainless steel welded 2KNZh
glove box (operator's view):' 1) glove-box frame;"
2) transfer. compartment; 3) base 4) hot and
cold.water and gas valves, etc.; 5) rubber gloves;
6) , viewing window; . 7) gate valve; 8) exhaust
filter; 9) ,water pressure gauge; 1-0) intake
filter.
A, rarefaction not less' than 20 mm H20'is maintained
inside the.glove box while, work is in progress.
The intake and exhaust filter systems are similar in
dimensions and design. The first and second stage of each
filter are mounted integrally on the glove box enclosure
frame, at the top of the unit. When a dosimetry check
indicates that the filters have reached a certain degree of
contamination, they are replaced.: The spent'filter is
discarded in a special waste receptable. Throttle valves
controlled from the front panel of the box straddle the
intake and exhaust passages. "
Radioactive waste solids may form in the course of
working with radioactive substances inside the glove box;
viz, rags, filter paper, damaged glassware, metal objects,
etc. A. waste receptable (see Fig. 2) containing a plastic
Fig. 5. Two-place stainless"steel 2KNZh-glove box,? ' radioactive'solids. 'This pouch is held by a 'special device
seen from behind (maintenance zone view): 1) air to the discard aperture of the glove-box table; and receives
duct; 2)' transfer compartment vacuum door; 3) radioactive solid wastes through 'that hole The discard
access hole to interior of box, for placing and re- aperture is fitted with a lid to.seal it hermetically, the lid.
moving equipment and instruments, and also for' being opened only inside the box. A 'pouch completely
maintenance and other work in''the interior of'the filled with solids is tied tightinside the box and is later low-
box; 4) liquid wastes drain; 5) liquid wastes ered into a valved metal receptacle fastened to the glove -
receptacle; 6) solid wastes dolly; 7). solid wastes ? box table. The lid is closed after the pouch is discarded, the
receptacle; 8) luminaires, .
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receptacle control valve is opened and the filled pouch is dropped into the solids receptacle resting on the dolly
beneath the glove-box table. This procedure for removing radioactive wastes from the box interior eliminates
any hazard of hot spills in the area occupied by personnel.
The ten- liter- capacity wastes receptacle (Fig. 3) is designed for temporary storage and hauling of alpha-,
beta-, and gamma-active waste solids. The receptacle is carted off on a dolly steerable with a meter-length
handle, and constitutes a cylindrical tank with carbon steel lid.
Drainage piping underneath the box table level receives liquid discard, which may also be funneled into
the liquid wastes receptacle.
A ten- liter- capacity 10KZhO receptacle (see Fig. 3 in the preceding article) is designed for storage and
carrying of liquid alpha-, beta-, and gamma-active wastes. The receptacle is a welded leak-tight vessel made
of stainless steel. An air filter and three pipe connections are contained in the receptacle lid. The air filter is
a connection sleeve with FPP filter fabric. The connection sleeve is capped by a plug, which is removed when
the receptacle is full. The waste receptacle may be moved either by means of the dolly or manually, using the
grip handles.
A two- place 2KNZh glove box, similar -in design to the 1KNZh, has been developed and is now being
manufactured, with a box interior volume of 0.8 ms, air exchange volume of 25 box volumes per hour, exhaust
filter area of 0.25 m2 and intake filter area of 0.25 m2 (see Fig. 4, 5).
This two- place glove box is equipped with waste receptacles, two pairs of special rubber gloves, and other
accessories. A maintenance access door 600 x 500 mm is built into the rear wall of the box. The entire. glove
box unit with accessories in place weighs 600 kg.
The 1KNZh and 2KNZh glove boxes are supplemented by viewing windows, rubber gloves, filter systems,
fluorescent lighting, plastic waste transfer pouches, ball- joint manipulator (optional), 10KZhO liquid waste
receptacles, solid waste containers, rating plate, and instructions for maintenance and operation.
BRIEF COMMUNICATIONS
USSR. A scientific colloquium on heat stress in rods, plates, and shells used in turbomachinery design was
held at the Kiev House of Scientists, during June,1960. The conference was organized by the permanent Commission
on Turbomachinery of the State Science and Engineering Committee of the Ukrainian SSR jointly with the
Academy of Sciences of the Ukrainian SSR, the Commission on the Strength of Gas Turbines of the Academy of
Sciences of the Ukrainian SSR. and was chaired by the Director of the Institute of the Institute of Mechanics of
the Academy of Sciences of the Ukrainian SSR, A. D. Kovalenko, a Corresponding Member of that body-
A paper presented by V. I. Danilovskii (Institute of Mechanics of the Academy of Sciences of the USSR)
was devoted to applications of the theory of a complex variable in determining temperature fields and thermal
stresses in two connected plane regions. A report by S. Ya. Yarema (Institute of Machine Design and Auto-
mation of the Academy of Sciences of the Ukrainian SSR) dealt with an investigation of the fundamental solution
of the temperature problem for cylindrical shells, based on the familiar solution for a shell subjected to lumped
forces. In his report, R. A. Adadurov (Central Aerohydrodynamics Institute, Moscow) discussed flexural and
torsional modes of loss of stability of unevenly heated thin, long strips. Lowering of the torsional rigidity of a
plate in response to uneven heating was pointed out. A. G. Kostyuk (Moscow Power Engineering Institute) dealt
in his report with a method for determining the temperature field and thermal stresses in a gas turbine wheel by
replacing the real wheel with a model convention.
A paper submitted by V. I. Savchenko (Kiev University) was based on the use of polarization optical
techniques applied to the determination of thermal stresses in planar models with convective heat transfer over.
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the lateral surfaces by way of a predetermined temperature dependence of the variation of optical and elastic
constants of the material comprising the model. D. A. Gokhfel'd (Chelyabinsk Polytechnical Institute) dis-
cussed the adaptability of designs to iterated thermal effects. Up to a certain temperature, the adaptability
temperature, plastic deformation is caused solely by the first thermal stress cycle and, at the maximum tem-
perature of a cycle which is higher than the adaptability point, plastic flow is caused by each thermal cycle.
Calculations take into account yield limit as a function of temperature.
A report delivered by Ya. S. Podstrigach (Institute of Machine Design and Automation of the Academy
of Sciences.of the Ukrainian SSR) was devoted to a determination of a steady-state temperature field in bounded
plates and shells. In a paper by A. D. Kovalenko, the problem of thermoelastic stresses in sloping shells of
revolution was discussed, with variability of shell thickness and elastic modulus taken into account.
S. V. Fol'kovskii adduced a solution of the problem of temperature fields and thermoelastic stresses in
steam piping A. F. Pronkin suggested a method for calculating the strength of contoured turbine wheels in
terms of a limiting state.
In her report, E. D. Pletnikovaya (Central Aerohydrodynamics Institute, Moscow) discussed the problem
of steady-state temperature fields in a system of rods converging to a single point, with convective heat transfer
over the lateral surfaces.
The conference took note of the need to develop experimental techniques for determining temperature
fields and thermal stresses to match performance data, the problem of evolving efficient methods for calculating
thermal stresses in elastic and plastic regions affected by the dynamics of a process.
Yugoslavia. An agreement was signed with Indonesia for collaboration in the cause of the peaceful uses
of atomic energy. The accord envisages in particular, collaborative efforts in the field of theoretical research,
geology, the technology of winning and processing nuclear raw materials, and the study of problems related to
the production of nuclear power and shielding against nuclear radiations. Collaboration between the two countries
will be developed by an exchange of specialists, documentation, and equipment.
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BIBLIOGRAPHY
Books and Symposia
B. V. Petukhin. Heat Power Engineering in Nuclear Facilities. Moscow, Atomizdat, 1960, 323 pages,
8 rubles, 80 kopeks.
The book outlines available information on nuclear reactors; presents the basic data on design and cal-
culations. of nuclear facilities working on steam and gas turbine power cycles; discusses the characteristics of
commonly used reactor coolants; describes in detail the flow schemes of the most typical nuclear electric power
stations now operating or in construction; cites data needed for the calculation and design of heat exchange
equipment and steam raising units for nuclear electric power stations.
The book may be used as a text for students in physics and power engineering departments, and may also
prove useful for scientific workers, engineers, and graduate students interested in problems of nuclear power and
the design of heat exchange equipment.
G. E. Kaplan, T. A. Uspenskaya, Yu. I. Zarembo, and I. V. Chirkov. Thorium: its Raw Material
Resources, Chemistry, and Technology. Moscow, Atomzidat, 1960, 226 pages, 8 rubles, 60 kopeks.
This book presents data on thorium which have appeared in the Soviet or foreign literature during the past
15-20 years. The basic information on the geochemistry and mineralogy of thorium is given, and light is shed
on the present state of the raw materials base of thorium abroad. The physicochemical, corrosive, and radioactive
properties of thorium, its fields of applications, the processes used in producing thorium metal (both technical-
grade and high-purity thorium) are offered, as well as the basic thorium compounds and alloys. Concise informa-
tion is presented on the analytical chemistry of thorium. The book is well illustrated. Each chapter ends with a
pertinent literature reference list.
The book is written for scientific and engineering workers and technicians; it may also be useful to
students in chemical and metallurgical advanced vocational schools.
A. M. Rozen. Theory of Isotope Separation in Columns. Moscow, Atomizdat, 1960, 438 pages,
16 rubles, 50 kopeks.
The first section of the book (chapters 1 to 6) deals with the peculiarities of individual methods of counter-
current separation: distillation, isotope exchange, thermodiffusion, mass diffusion, and centrifugation. The
second section (chapters 7 to 10) describes the general methods for designing columns and cascades. The theory
and methods of column technology are developed. The general theory of separation is applied to the optimization
of fundamental column parameters. Attention is focused on optimization of.separation conditions using two-phase
methods, when a fraction of the losses are proportional not to the work of separation, but to the process stream.
Chapter 11 singles out some comparatively simple formulas for calculating the rate of approximation to
equilibrium in various types of columns.
Chapter 12 contains examples of the design of separation facilities and, to a certain extent, compares
different techniques of heavy water production and other isotope processes.
The book is written for scientific workers, chemical engineers, and students interested in problems of
separation of mixtures, and may also be found useful in designing separation facilities.
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A. V. Lebedinskii and Z. N. Nakhil'nitskaya. The Effect of Ionizing Radiations on the Nervous System.
Moscow, Atomizdat, 1960, 187 pages, 7 rubles, 60 kopeks.
Material from Soviet and foreign literature on the effect of ionizing radiations on analyzers and on the
nervous system. Particular attention is focused on the work of Soviet researchers. On the basis of a generaliza-
tion of data culled from the literature and their own data, the authors develop a concept of the mechanism
involved in the disturbance of functions of the nervous system in response to exposure to ionizing radiations.
The book is written for a broad readership of scientific workers in the field of radiobiology, biophysics,
physiology, pathophysiology, and also for practicing physicians.
V, A. Sokolov. The Radiosulfur Isotope SsS, Moscow, Atomizdat, 1960,.16 pages, 60 kopeks.
This brochure presents the basic information required on the radioactive sulfur isotope S*. Methods for
producing the isotope are described, as well as the properties and field of applications of the isotope; some data
are offered on the chemistry and technology of S35-tagged compounds, as well,as information on safety techniques.
The book is written for a broad audience.
Neutron and reactor physics. Physics of hot plasmas and controlled fusion.
Physics of the acceleration of charged particles.
Zhur. eksptl. i teoret. fiz. 38, 6 (1960)
V. N. Nefedov, pp. 1657-1662. On the mechanism of emission of prompt fission neutrons.
S. E. Grebenshchikov and M. D. Raizer, pp. 1665-1667. The skin effect and shock waves in an inductive
gas discharge.
K. A. Petrzhak et al., pp. 1723-1728. Spread of ranges and kinetic energy of fission fragments of U gas.
A. B. Kitsenko and K. N. Stepanov, pp. 1840-1846. On the instability of a plasma with an anisotropic
distribution of ion and electron velocities.
Pribory i tekhnika eksper. No. 3 (1960)
Yu. G. Abov and D. F. Litvin, pp. 3-15. Experimental neutron diffraction methods (survey article).
N. P. Glazkov, pp. 16-19. Cylindrical fast-neutron spectroscopy camera.
J. Appl. Phys. 31, 6 (1960)
P. Auer and H. Hurwitz, pp. 1007-1009. Space charge neutralization by positive ions in diodes.
Nucleonics 18, 7 (1960)
G. Johnson, pp. 49-53. Status and promise of peaceful nuclear explosions.
D. Hughes, pp. 54-58. Neutron spectroscopy with nuclear explosions.
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Nukleonik 2, 4 (1960)
K. Beckurts, pp. 129-131. A simple time-amplitude converter for slow neutron time-of-flight experiments.
A. Kraut, pp. 149-174. Results of physical research on fission of nuclei.
Nukleonika V, 4 (1960)
T. Rzeszot, pp. 191-194. Measurement of the angular distribution of neutrons scattered from a reactor
experimental channel.
Progr. Theor. Phys. 23, 3 (1960)
Y. Ichikawa, pp. 512-518. Bremsstrahlung in a high-temperature plasma.
Reactor Science 12, 1-2 (1960)
D. Stupegia, pp. 16- 20. Thermal neutron cross sections for the Cs37 (n, y) Cst3s reaction.
R. Tattersall et al., pp. 32-46. Measurement of resonance absorption integrals using pile oscillators.
II. NUCLEAR POWER ENGINEERING
Nuclear reactor theory and calculations. Reactor design.
Performance of nuclear reactors and nuclear power stations.
Morskoi flot 7 (1960)
1. Bykhovskii, pp. 15-16. On nuclear-propulsion merchant ships.
Atompraxis 6, 6 (1960)
- - - pp. 231-238. The Hannover industrial fair.
Atomwirtschaft V, 6 (1960)
F. Hoffman et al., pp. 250- 253. The U. S. HTGR high-temperature reactor project.
R. Nass, pp. 257- 264. Materials and welding operations in reactor design.
- - - pp. 273- 276. Nuclear technology at the Hannover fair.
D. Weissbarth, pp. 276-281. On the uses of various research reactors.
Energia Nucl. 7, 6 (1960)
G. Calabria, pp. 385- 396. The Latina nuclear electric power station.
L. Biondi, pp. 397-406. On the further development of organic-moderated reactor designs.
A. Cicchitti et al., pp. 407-425. Experimental investigations of cooling with a two-phase fluid coolant.
Measurements of pressure drop, heat transfer, and burn-out heat loads.
F. Faldini et a]., pp. 426-434. Study of the dynamics of a nuclear power plant steam generator, by
analog simulation.
Energia Nucl. 7, 7 (1960)
S. Battaglia and U. Cassinary, pp. 457-462. Applications for organic fluids in nuclear reactors.
G. Casini et al., pp. 488-495. Nuclear composition of natural uranium fuel after prolonged irradiation
in a thermal reactor.
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Kernenergie 3, 5 (1960)
W. Mai, pp. 407-413. Calculations of temperature field and thermal stresses in a pressurized-water
reactor vessel.
Kerntechnik II, 6 (1960)
0. Werner, pp. 192- 200. New trends in fuel element development.
K. Kimbel, pp. 208-211. Report on the seventh conference on hot laboratories and associated equipment
(Cleveland).
G. R6bert, pp. 211- 212. The French nuclear industry represented at the Hannover fair.
Nucl. Energy No. 146 (1960)
- - - - pp. 305-307. The Windscale AGR reactor, as the prototype for a new type of nuclear power
stations.
C. Wheihel and C. Robbins, pp. 321-322. The use of rarefactions in the containment shells of nuclear
power reactors.
Nucl. Power 6, 51 (1960)
R. Campbell, pp. 68-72. Power reactor control. 1.
- - - - pp. 73-75. Economic problems in atomic energy.
F. Theunissen, pp. 76- 79. Economical uses of conventional and nuclear fuels.
D. Iggulden, pp. 84-87. Design of gas-cooled reactors.
P. Garay, pp, 96-99. Use of hydrogen for reactor cooling.
--- - p. 101. Fuel elements for the DRAGON reactor.
Nucl. Power 5, 52 (1960)
- - - - p. 66. The Dungeness nuclear power station.
J. Collier and P. Lacey, pp. 68-73. Two-phase liquid coolant for nuclear reactors.
R. Campbell, pp. 74-77. Control of power reactors. 2.
J. Burkett, pp. 78-82. Nuclear-powered merchant ships.
G. Deleuze and G. Petillat, pp. 98-99. Decladding fuel elements at Marcoule.
G. Wallis, pp. 99-101. Gas-fluid model for bulk boiling.
Nucleonics 18, 7 (1960)
R. Lightle, p. 59. Criticality of MTR-type fuel elements.
W. Brown and C. Bergmann, pp. 60-63. Reducing radioactivity buildup in pressurized-water reactors.
D. Bradley, pp. 84-88. Hydrogen as a power reactor coolant.
W. Lewis and R. Goin, p. 91 and p. 93. Safer packages for shipping fuel.
Nukleonik 2, 4, (1960)
M. Kuchle, pp. 131-139. Measurement of the temperature dependence of neutron diffusion length in water
and diphenyl, by the pulsed method.
M. Brose and K. Beckurts, pp. 139-141. Measurements of neutron absorption cross section in aluminum.
G. Blasser, pp. 141-144. Slowing down of neutrons in a heterogeneous system.
T. Springer, pp. 144-149. Spatial dependence of neutron temperature in a moderator.
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Nukleonika V, 3 (1960)
A. Kostyrko, pp. 133-142. Studies of radiolytic decomposition of water in the primary coolant of the
Polish EVA reactor (WWR-S).
Reactor Science 12, 1-2 (1960)
A. Takahashi, pp. 1-15. Probability of the 'first impact' in periodic systems.
M. Angelopoulos, pp. 21- 25. Calculation of attenuation of neutron flux in thin absorbing rods.
H. Takahashi, pp. 26-31. Probability of resonance escape in annular fuel elements.
Nuclear geology and primary ore technology.
Nuclear metallurgy and secondary ore technology.
Chemistry of nuclear materials.
Radiokhimiya II, 3 (1960)
V. I. Grebenshchikova and R. V. Bryzgalova, pp. 265- 273. Study of coprecipitation of Pu (IV) with
lanthanum oxalate.
E. I. Moiseenko and A. M. Rozen, pp. 274-280. Distribution of plutonium in extraction with tri-n -butyl
phosphate.
V. B. Shevchenko et al., pp. 281- 290. On the effect of hydrocarbons of the aliphatic and aromatic series
on extraction of U (IV), Pu(IV), Zr(IV), and Ce(III) from nitrate solutions, using tri- n-butyl phosphate.
V. M. Vdovenko and E. A. Smirnova, pp. 291- 295. On the hydration of uranyl nitrate in organic solvents
in extraction from salt solutions.
V. M. Vdovenko et al., pp. 296- 300. Infrared spectra of organic solutions of uranyl nitrate hydrates in the
region of bending vibration frequencies of water.
V. M. Vdovenko et al., pp. 301-306. Spectrophotometric investigation of the formation of nitrato com-
plexes of plutonyl compounds in acetone.
V. M. Vdovenko et al., pp. 307-311. Spectrophotometric investigation of the formation of nitrato com-
plexes of plutonyl compounds in aqueous solutions, and extraction of Pu (IV) with dibutyl ether.
V. M. Vdovenko et al., pp. 312-314. On the formation of a complex trinitrate neptunyl compound.
P. I. Kondratov and A. D. Gel'man, pp. 315-319. Oxalate compounds of tetravalent neptunium.
V. D. Nikol'skii et al., pp. 320-329. Properties of plutonyl nitrate solutions. III. Stability of plutonyl
in nitrate solutions.
V. P. Nikol'skii et al., pp. 330-338. On the existence of a monoacetate uranyl complex in solution.
A. A. Zaitsev et al., pp. 339-347. Disproportionation of americium M.
A. A. Zaitsev et al., pp. 348-350. Kinetics of the reduction of americium (V) with hydrogen peroxide.
I. A. Lebedev et al., pp. 351-356. Study of complexing of Am with oxalate ions.
V. 1. Kuznetsov and T. G. Akimova, pp. 357-363. Organic coprecipitants. XIII. Coprecipitation of
tetravalent plutonium.
V. N. Bobrova, pp. 364-368. Determination of solubility of double sulfates of zirconium and plutonium in
saturated potassium sulfate solutions, using radioactive tracers.
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T. A. Slepyan and S. M. Karpacheva, pp. 369-376. Physicochemical properties of nitrate solutions of
uranyl nitrate, and determination of the properties of those solutions (specific weight, electric conductivity,
refractive index).
L. E. Drabkina, pp, 377-378. Determination of the solubility of ammonium plutonyl carbonate in various
aqueous solutions.
Atomwirtschaft V. 6 (1960)
H. Krauch, pp. 265-268. Radiochemistry from the engineering and economics points of view.
Energia Nuci. 7, 7 (1960)
L. Damiani et al., pp. 463-469. A pulsed column for liquid-liquid extraction. II. Column design.
G. Imarisio, pp. 470-476. Preparation of a UO2 sinter.
E. Cerrai and C. Testa, pp. 477-487. Use of quercetine-hydrogen peroxide in the calorimetric determination
of hafnium in zirconium.
J. Inorg. and Nucl. Chem. 13, 1-2 (1960)
L. Asprey and F. Kruse, pp. 32-35. Bivalent thulium. Thulium di-iodide.
N. Brett et al., pp. 44-53. Investigation of the uranium-thorium-carbon system.
J. Danon, pp. 112-118. Determination of the stability constants of thorium nitrate complex, using anion-
exchange resins.
W. Brown et al., pp. 119-124. Extraction of the lanthanides using acetylacetone.
F. Habashi, pp. 125-137. Processes for recovering uranium from phosphoric acid.
D. Peppard et al., pp. 138-150. Extraction of thorium (IV) with diethers of orthophosphoric acid
(GO)2PO(OH).
T. Siddall, pp. 151-155. Effect of various alkyl substituents in trialkylphosphates on extraction of actinides.
E. Hesford and H. McKay, pp. 156-164. Extraction of mineral acids with tri-n-butyl phosphate.
E. Hesford and H. McKay, pp. 165-173. Extraction of uranyl perchlorate with tri-n-butyl phosphate.
C. Hardy and D. Scargill, pp. 174-180. Extraction of niobium (V) from nitrate solution with tri-n-butyl
phosphate.
J. Inorg. and Nucl. Chem. 13, 3-4 (1960)
R. Panzer and J.,Suttle, pp. 244- 247. A new uranium complex compound 5UC15 - CC12 = CCl - COC1.
D. Mathews et al., pp. 298-309. Investigation of transference and solvation phenomena. 1. Uranium
chloroxide in water, ethanol, and water-alcohol mixtures.
J. Kooi et al., 310-312. Solubility of Cs2UCI6, Cs2UOC14, and Cs2PuC16 in hydrochloric acid.
G. Duyckaerts et al., 332-333. Extraction of lanthanides in the form of their di-n-butylorthophosphates.
Kernenergie 3, No. 5 (1960)
D. Naumann and R. Ross, 425-428. Anion exchange of cerium, thorium, and uranium in concentrated elec-
trolytes. F
R. MUnze, 429-434. Removal of trace amounts of fluorine compounds by coprecipitation with barium sulfate.
Kernenergie 3, No. 6 (1960)
R. MUnze, 518-521. A new method for continuous production of carrier-free I"32 and Tent.
Nucl, Energy No. 146 (1960)
- - - , 317 ff. Fluorine derivatives and atomic energy. II. Properties and fields of application of fluorine.
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Nucleonics 18, No. 7 (1960)
R. Koch and G. Grandy, 76 ff. Xenon-krypton separation by gas chromatography.
Nukleonika V, No. 3 (1960)
J. Minczewski et al., 115-122. 'Some methods used in the separation and determination of uranium.
W. Korpak and Sz. Deptula, 123-132. Effect of chlorides and sulfates on solvent extraction (using TBP) of
uranium and iron.
Nukleonika V, No. 4 (1960)
E. Herczyr ska, 195-204. A mechanism effecting precipitation of anions and cations from aqueous solutions.
S. Siekerski and R. GWO5,Pdg, 205-217. Investigation of the TBP-HC1O4-H2O system.
Z. Zagurski and W. Nej, 219-226. Novel arrangement used for direct investigation of systems in a gamma-
radiation field, by physical-chemistry means.
VDI-Zeitschrift 102, No. 11 (1960)
K. Frank, 409-424. Reactor metals. Some metals used in reactor'design: thorium, beryllium, zirconium,
hafnium, lithium, and others.
IV. NUCLEAR RADIATION SHIELDING
Radiobiology and radiation hygiene. Shielding theory
and techniques. Instrumentation
Biofizika 5, No. 3 (1960)
S. A. Valeva, 362-365. On effects of gammas and neutrons on seeds of agricultural plants.
Gigiena i sanitariya No. 5 (1960)
0. S. Andreeva, 77-82. Some data on the effects of exposure to uranium and uranium compounds in the human
organism (review paper).
Gigiena truda i prof. zabolevaniya No. 6 (1960)
N. Yu. Tarasenko, 21-27. Note on thorium toxicity.
Izvestiya akad. nauk SSSR, seriya biol. No. 3 (1960)
A. M. Kuzin, 355-363. Modern problems in radiobiology.
Med. radiologiya 5, No. 4 (1960)
I. K. Sokolova, 68-72. Investigation of chloroform in aqueous solution as a possible dosimeter for x radiation
and gamma radiation.
Med. radiologiya 5, No. 5 (1960)
V. V. Dmokhovskii, 78-84. Shielding calculations for a 25 Mev betatron.
Pribory i tekhnika eksperimenta No. 3 (1960)
Yu. V. Dukarevich and A. N. Dyumin, 48-50. Effective fast-neutron detector with little sensitivity to gamma
rays. -
M. Yu. Tissen, 51-53. Note on one possibility of achieving absolute measurements of the activity of C14-
labeled and S35-labeled gases.
Teplodnergetika 7, No. 7 (1960)
A. V. Surnov, 90-91. On calculations of the size of a gamma source by transmission of a pencil beam of
gamma photons.
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Atompraxis 6, No. 6 (1960)
W. FUchtbauer and W. Simonis, 217-219. Equipment for beta irradiation of suspended biological objects.
D. Sauerbeck, 221-225. C14 labeling of plants.
K. Scharrer and S. Heilenz, 226-227. A simplified technique for using cation exchange to determine strontium
in plants.
E. Welte and U. Marckwordt, 228-229. Simplified method of Sr90 determinations in plants.
Energia Nucleate 7, No. 7 (1960)
R. Somigliana, 495-496. Determination of spurious counts in coincidence measurements.
Kernenergie 3, No. 5 (1960)
H. Abel and V. Bredel, 414-421. Preparation and study of neutron scintillators.
K. Oertel et al., 422-424. Dissociation of isopentane in Geiger-Muller, counters.
Kerntechnik II, No. 6 (1960)
M. Pollermann, 185-191. Low-pressure ionization chamber filled with pure water vapor.
T. Nielsen, 201-202. A device for continuous monitoring of beta and gamma activity in air and gases.
T. Springer and M. Oberhofer, 203-205. Neutron shielding and gammas released in the shielding.
Nucl. Power 5, No. 51 (1960)
D. Taylor, 88-92. Dosimetric instrumentation. 4.
R. Brooks, 94-96. Personnel shielding problems in beryllium reactor applications.
Nucl. Power 5, No. 52 (1960)
W. Loosemore, 84-89. Ionization radiation detectors.
D. Taylor, 91-94. Dosimetric instrumentation. 5.
Nucleonics 18, 7 (1960)
J. Waugh and R. Nicholson, pp. 70-74. Transistor amplifier for fast proportional counting.
Nukleonika V, 3 (1960)
J. Domanus and B. Osuchows.ki, pp. 143-148. On the concept of the "Gram-equivalent of radium' for
nonexact isotopic gamma sources.
A. Malik and J. Malesa, pp. 149-155. Relative measurements of organic scintillators.
Nukleonika V, 4 (1960)
K. Mikke et al., pp. 181-189. ZnS (Ag) paraffin type scintillators for detecting fast neutrons.
J. Domanus and L. Halski, pp. 227- 238. Film dosimeter measures gamma radiation doses of radioactive
isotopes.
V. RADIOACTIVE AND STABLE ISOTOPES
Labeled-atom techniques. Uses of radioactive radiations.
Direct conversion of nuclear energy to electrical energy.
Voprosy ekonomiki No. 5 (1960)
G. Mikheev, pp. 67-76. Economic advantages in the use of radioactive isotopes in industry.
Izvestiya akad. nauk SSSR. Otdel tekhn. nauk. Metallurgiya i toplivo No. 2 (1960)
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S. P. Zaitseva et al., pp. 120-132. Applications of radioactive isotopes and nuclear radiations in research
on the flotation process.
Izvestiya vyssh. ucheb. zaved. Tsvet. metallurgiya No. 3 (1960)
A. I. Levin and V. I. Falicheva, pp. 62-69. Studies of cathodic processes in electrodeposition of zinc,
using tracer techniques.
Koks i khimiya No. 5 (1960)
A. Z. Kulishenko, pp. 17-19. Automation of the flotation shop at the Yasinovka coke and chemicals plant,
using radioactive density gauges.
Konserv. i ovoshchesushil'n. prom. No. 7 (1960)
L. V. Metlitskii, pp. 19- 22. Uses of nuclear radiations in the food processing industry and in agriculture.
Prom. stroitel'stvo No. 5 (1960)
V. A. Volokhov and Yu. D. Markov, pp. 47-50. Quality control of concrete placing operations, using
tracer quality control techniques.
Atomwirtschaft V, 6 (1960).
L. Wiesner, pp. 253- 256. Applications of nuclear physics techniques in oil prospecting and exploitation. II.
Nucl. Energy No. 146 (1960)
- - - - pp. 333-337. Radioactive isotope applications in railroad transportation in the USA.
Nucl. Power 5, 51 (1960).
B. Lindley, pp. 80-83. Direct conversion. 2.
Nucleonics 18, 7 (1960).
W. Mayer et al., pp. 64-68. Radioactive gauges for air-cooled turbine blades. Liquid-source-gauge for
hollow blades and vanes. Scanning beta gauge for cooling-channel walls.
C. Clayton, pp. 96-100. Precise tracer measurements of liquid and gas flows.
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Declassified and Approved For Release 2013/02/14: CIA-RDP10-02196R000100060004-4
 Declassified and Approved For Release 2013/02/14: CIA-RDP10-02196R000100060004-4
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Declassified and Approved For Release 2013/02/14: CIA-RDP10-02196R000100060004-4
 Declassified and Approved For Release 2013/02/14: CIA-RDP10-02196R000100060004-4
META.nAyPrH.5q
n0J1ynPOBOJHHKOB'
THE METALLURGY OF SEMICONDUCTORS'
by U.N. Shashkov.
Translated from Russian
(Original published by the State Scientific-Technical Press for Literature on
Ferrous and Non-Ferrotts Metallurgy, Moscow)
In this up-to-date, survey of semiconductor metallurgy, the two main semiconductor
materials - germanium and silicon - and their physical and chemical properties,are
discussed. A detailed review of chemical methods of purifying these elements, is 'fol-
lowed by a description of metallurgical methods of purification. Partition coefficients
are given for many impurities and apparatus used in purification is, surveyed.
Methods 'of growing single crystals are treated in an excellent chapter, with the method
related to the type of defect to be expected in the finished crystal. Of particular interest.
is a discussion of alloying and doping.
The last section of the volume deals with heat treatment and diffusion of impurities,
production of rectifiers and transistors, and methods 'of etching crystals and finished
devices.
Technologists. concerned with industrial processes of established value, as well as grad-
uate students, will?find this book of great value. The 293 references provide a good selec-
tive guide, to the existing literature on the subject.
PARTIAL CONTENTS
Physicochemical,' Electrical, and Optical Prop-
erties of Germanium and Silicon
Physicochemical Properties of Germanium and
Silicon
Electrical Properties of Germanium and Silicon.
'Optical Properties of Germanium and Silicon
Chemical Methods of Purifying Germanium and
I Silicon
Production of Pure Germanium
Production of Pure Silicon
Acid Washing of Silicon
Reduction of Silicon Tetrachloride with Zinc
(Beketov's I Method)
Iodide Method of Producing Silicon
Si lane Method of Producing Silicon
Trichlorosilane Method of Producing' Silicon
Reduction of Silicon Tetrachloride with Hy-
drogen
Other Methods of Producing High-Purity Silicon
Metallurgical Methods of Purifying Germanium
and Silicon
Zone Melting, Pulling, and Controlled Crystal-
lization
Theory of Distribution of Impurities in Zone
Melting, Pulling, and Controlled,,Crystal-
lization
Distribution of Impurities in Zone Melting and
Controlled Crystallization
Distribution of Impurities in Pulling and Con-
trolled Crystallization'
Distribution of Impurities in Zone Melting '
Evaporation of Impurities During Fusion and
Pulling Under vacuum
The Partition Coefficient
Apparatus' Used in Metallurgical Methods of
Purifying Germanium and Silicon
Practical Details of Metallurgical' Methods 'of
Purifying Germanium and Silicon
CONSULTANTS BURFrAU
ENTERPRISES, INC.
227 W. 17 ST., NEW YORK 11, N. Y.
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 Declassified and Approved For Release 2013/02/14: CIA-RDP10-02196R000100060004-4
VOLUME I
VACUUM MICROBALANCE TECHNIQUES
Proceedings of, the 1960 Conference Sponsored by
The Institute for Exploratory Research
U. S. Army Signal Research and Development Laboratory
Edited. by
M. J. KATZ'
U. S. Army Signal Research and Development Laboratory
Fort Monmouth, New Jersey
Introduction by
Thor N. Rodin
Cornell University
The proceedings 'of this conference provide
an authoritative introduction to the rapidly
widening scope of microbalance methods
which is not available elsewhere in a single
publication. -
The usefulness of microbalance techniques in
the study of the properties of materials lies
in their extreme sensitivity and versatility.
This renders them particularly important in
studies of properties of condensed systems.
In'addition to the historical use of microbal-
ance techniques as a tool of microchemistry,
they have, in recent years, found extensive ap-
plication in the fields of metallurgy, physics,
and chemistry. The uniqueness of the method
results from the facility it provides in making
a series of precise measurements of high sen-
sitivity under carefully controlled conditions
over a wide range of temperature and
pressure. 170 pages $6.50
N T P L
E N U M PRESS, INC. 227, West 17th St., New York, 11, N. Y.
This significant new volume contains papers
in three major categories. The first group of
reports deals with the general structural
features'-and measuring capabilities of micro-
balances. In.the second group, a sophisti-
cated consideration and much needed evalua-
tion of sources of spurious mass changes
associated with 'microbalances is presented.
The third group describes some of the most
recent extensions in microbalance work to
new research areas such as semiconductors,
ultra-high vacuum, and high temperatures.
These papers.provide an interesting account
of advances in the application of the micro-
gravimetric method to three new and impor-
tant fields of research on the behavior of
materials.
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