THE SOVIET JOURNAL OF ATOMIC ENERGY VOL. 8 NO. 6

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• Collection: General CIA Records
• Document Number (FOIA) /ESDN (CREST): CIA-RDP10-02196R000100050006-3
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• Original Classification: K
• Document Page Count: 111
• Document Creation Date: December 27, 2016
• Document Release Date: February 19, 2013
• Sequence Number: 6
• Publication Date: July 1, 1961
• Content Type: REPORT

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111
Volume 8, No. 6
THE SOVIET JOURNAL OF
July, 1961
OMIC ENERGY
7.?is,NS
CONSULTANTS BUREAU
A
1-1TOMHa51
311qpr1151
RoLk auzarAN
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Molecular Structure and Reactivity in Catalysis. A. A. Balandin
The Role of the Electron Factor in Catalysis. S. Z. Roginskii
The Principles of the Electron Theory of Catalysis on Semiconductors.
F. F. Vorkenshtein
- The Use of Electron Paramagnetic Resonance in Chemistry.
V. V. Voevodskii
The Study of Chain and Molecular Reactions of Intermediate Sub-
stances in Oxidation of e-Decatte. Z. K. Maizus, I. P. Skibida,
N. M. Emanuel' and V. N. Yakovleva
The Mechanism of Oxidative Catalysis by Metal Oxides. V. A. Roiter
The Mechanism of Hydrogen-Isotope Exchange on Platinum Films.
G. K. Boreskov and A. A. Vasilevich
Nature of the Change of Heat and Activation Energy of Adsorption with
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Catalytic Function of Metal Ions in a Homogeneous Medium.
L. A. Nikolaev
Determination of Adsorption Coefficient by Kinetic Method. I. Adsorp-
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Contents of the first issue include:
Electron-Diffraction Investigation of the Structure of Nitric Acid and
Anhydride Molecules in Vapors. P. A. Akishin, L. V., Vilkov and
V. Ya. Rosolovskii
Effects of Ions on the Structure of Water. I. G. Mikhailov and Yu P.
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Proton Relaxation in Aqueous Solutions of. Diamagnetic Salts. I. Solu-
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Second Chapter of Silicate Crystallochemistry. ,N. V, Belov
Structure of Epididymite NaBeSi30,0H. A New Form of Infinite Silicon
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Phases Formed in the System Chromium?Boron in the Boron-Rich
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Crystal Structure of the Ternary Phase in the Systems Mo(W)?
Fe(CO,Ni)?Si. E. I. Gladyshevskii and Yu. B. Kyz'ma
Complex Compounds with Multiple Bonds in the Inner Sphere.
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A. E. Kalmanson
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EDITORIAL BOARD OF
ATOMNAYA ENERGIYA
A. I. Alikhanov
A. A. Bochvar
N. A. Dollezhar
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
( iteeistant Editor)
A. P. Zefirov
THE SOVIET JOURNAL Of
ATOMIC ENERGY
A translation of ATOMNAY A ENERGIY A,
a publication of the Academy of Sciences of the USSR
(Russian original dated June, 1960)
Vol. 8, No. 6 July, 1961
CONTENTS
RUSS.
PAGE PAGE
The 50-Megawatt SM Research Reactor. S. M. F einber g, S. T. Kono -
beevskii, N. A. Dollezhal', I. Ya. Emel'yanov, V. A.
Tsykanov, Yu. M. Bulkin, A. D. Zhirnov, A. G. Filippov,
0. L. Shchipakin, V. P. Perfil'ev, A. G. Samoilov,
and V. I. Ageenkov 409 493
New Ideas in the Structural Design and Layout of Nuclear Reactors.
A. N. Komarovskii 420 505
Mechanical Properties and Microstructure of Certain Construction Materials
After Neutron Irradiation. I. M. Voronin, V. D. Dmitriev, Sh. Sh.
Ibragimov, and V. S. Lyashenko 429 514
Extraction of Uranium from Solutions and Pulps. B. N. Lask or in , A. P.
Zefirov, and D. I. Skorovarov 434 519
Interaction of Uranium Hexafluoride with Ammonia. N. P. G a lkin,
B. M. Sudarikov, and V. A. Zaitsev 444 530
The Flocculation of Pulp and Polyacrylamide-Type Flocculents. I. A. Yak ub ov i ch . 449 535
Determination of Absorbed Doses in Organisms Exposed to Emanations
and Their Daughter Products. L. S. Ruzer 455 542
LETTERS TO THE EDITOR
Absorption Section of Fast Neutrons. T. S. Be 1 novaa 462 549
Convergence of the Series in the Many-Velocity Theory of Neutron Diffusion.
A. V. Stepanov 464 550
A Ring Cyclotron Accelerator with Vertically Growing Magnetic Field.
A. P. Fateev and B. N. Yablokov 468 552
Some Properties of Accelerator Orbits Where Similitude Is Observed.
A. A. Kolomenskii and A. N. Lebedev 471 553
Measurement of the Radiative-Capture y -Emission Spectra of Neutrons in Some Rocks.
A. A. Fedorov, M. M. Sokolov, and A. P. Ochkur 474 555
Slowing Down of N'eutrons in Steel-Water Mixtures. L. A. Geraseva
and V. V. Vavilov 476 556
Determination of Degrees of Equilibrium of Short-Lived Radon Daughters in Air.
L. S. Ruzer 478 557
Luminescent Dosimeters Based on the CaSO4 ? Mn Phosphor for the Detection
of Gammas, Betas, and Neutrons. V. A. Arkhangel'skaya, B. I.
Vainberg, V. M. Kodyukov, and T. K. Razumova 481 559
Annual subscription $ 75.00 ? 1961 Consultants Bureau Enterprises, Inc., 227 West 17th St., New York 11, N. Y.
Single issue 20.00 Note: The sale of photostatic copies of any portion of this copyright translation is expressly
Single article 12.50 prohibited by the copyright owners.
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CONTENTS (continued)
RUSS.
PAGE PAGE
SCIENCE AND ENGINEERING NEWS
Letter from a Reader (on the Article "Entropy Trapping of a Plasma by a Reversal of the
Magnetic Bottle Configuration"). L. A. A rt s imo vich 485 562
Generation of a Z-Hyperon by Negative Pions with a Momentum of 8.3 Bev/c.
M. I. Solov'ev 486 562
Hungarian Exhibit of Instruments for Experimental Nuclear Physics Research 488 563
[Economics of Organic-Cooled Organic-Moderated Low-Power Reactors. V. V. Batov 564]
[The Piqua Organic-Moderated Reactor. V. V. B at ov 565]
[The New British Research Reactor (Jason). A. Seligman 568]
[Nuclear Power Developments in West Germany. Yu. M it yaev 570]
[Uranium Production in the Union of South Africa. R. R a f al' ski i 572]
Recent Data on C14 Concentration in the Atmosphere. Yu. V. Sivint sev 489 573
Applications of Alpha Radiation from Radioactive Isotopes for Quality Control in
Grinding Operations. V. V. Kondashevskii, A. N. Chertovskii,
V. S. Pogorelyi, and A. M. Gutkin 492 576
Brief Communications 494 578
BIBLIOGRAPHY
New Literature ? Books and Symposia 495 581
INDEX FOR JANUARY-JUNE, 1960
Table of Contents, Volume 8
Author Index xiii
NOTE
The Table of Contents lists all material that appears in Atomnaya fnergiya. 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.
Consultants Bureau Enterprises, Inc.
ERRATA
Vol. 8, No. 4, June, 1961
Page Column Line Reads
277 left 7-8 j/k2c2< 1
278 left 14-18 The increase . . .
Should read
4/k2c2 '5' 1
The increase in the magnitude of the longitudinal veloc-
ity component v11 c?-? vo to values> vo means
that it is necessary to take account of the additional re-
duction of Tc (E) due to the reduction in the time spent
by the ion (r11) in the region of the heating section:
T c 1 kVii.
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THE 50-MEGAWATT SM RESEARCH REACTOR
S. M. Feinberg, S. T. Konobeevskii,
N. A. Dollezhal', I. Ya. Emel'yanov,
V. A. Tsykanov, Yu. M. Bulkin, A. D.
Zhirnov, A. G. Filippov, 0. L. Shchipakin,
V. P. Perfil'ev, A. G. Samoilov, and V. I.
Age enkov
Translated from Atomnaya gnergiya, Vol. 8, No. 6,
pp. 493-504, June, 1960
Original article submitted March 15, 1960
The present article describes the design of the Soviet SM research reactor, where the neutral flux attains 2.2.1015
neutron/cm2 ? sec. The intensive neutron and y-ray fluxes that are obtained in the reactor make it possible to per-
form many investigations in the fields of nuclear physics and reactor techniques.
The SM reactor operates with intermediate neutrons and is characterized by a high ratio of the maximum
neutron flux to the heat outppt. This ratio determines the efficiency of any research reactor. For the SM reactor
this ratio is equal to 4.4.1015 neutron/cm2.sec.kw.
The installation of this reactor represents a considerable advance in the development of Soviet reactor con-
struction. The experience gained in the design and utilization of Soviet research reactors was used to a great
extent in constructing this reactor.
This article is not concerned with a description of the physical characteristics of the reactor; the main
emphasis is laid on the description of the engineering solutions on which the reactor design is based.
Purpose of the Reactor
The SM reactor is intended for scientific-research investigations connected with the use of the intensive
fluxes of thermal and fast neutrons and y -rays which are obtained in the reactor. The reactor operates with inter-
mediate neutrons and has a sufficient reactivity margin for simultaneous investigations in all experimental channels
as well as in beams. In correspondence with a planned program of investigations in the SM reactor, the following
experiments will be performed first;
1) production of new transuranic elements; ?
2) a study of the properties of fissionable and structural materials in neutron and y-ray fluxes at different
temperatures (from 20? K to 2000? C) and in different media (gas, water and pressures from 50 to 350 atm, liquid
metals, etc.);
3) investigations of the spectrum of intermediate neutrons by means of the spectrometry method;
4) investigations of the spectrum of y-rays from the (n, y) reaction;
5) a study of short half-life radioactive isotopes;
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6) investigations pertaining to neutron diffraction photography.
The reactor design also makes it possible to perform other investigations.
Basic Characteristics of the Reactor
The efficiency of experimental reactors depends to a considerable extent on the ratio of the maximum neu-
tron flux in the reactor to its heat output. The larger this ratio, the smaller the fuel expenditure. Therefore, in
designing the SM reactor, efforts were made to secure the maximum value of this ratio. Investigations showed
that, in intermediate reactors, this ratio can be considerably greater than in thermal reactors*.
The core of intermediate reactors has a smaller volume; moreover, for these cores, one can use structural
materials which permit the attainment of high energy intensities. Since the density of neutrons which are slowed
down in the reactor central water volume and in the lateral reflector is proportional to the energy intensity, such
reactors make it possible to obtain a great thermal neutron density for a comparatively low power level.
9)(r)
Fig. 1. Neutron flux distribution along the reactor radius. I) Core; II) central water
volume; III) reflector; IV) central experimental channel; 1) U235 fission density
measured by means of a fission chamber; 2)U235 fission density measured by means
of a fission chamber encased in cadmium; 3) without a specimen in the channel; 4)
with a specimen in the channel; the specimen absorbing power (with respect to ther-
mal neutrons) corresponds to 15 g of U235.
Thus, in the SM reactor, for a power of 50 Mw, the maximum flux of thermal neutrons in the lateral reflec-
tors attains 5.1014 neutronsicrnk sec, and, in the central water volume, this flux attains 2.2.1015 neutron/cm2 sec.
Thus, the ratio of the maximum thermal neutron flux to the power level in the SM reactor is equal to 4.4.1010 neu-
tron/cm2.sec?kw. In certain thermal reactors, this value is considerably lower. The neutron flux distribution along
the SM reactor radius is shown in Fig. 1.
Another important advantage of intermediate reactors is their longer period of operation. Thus, for a U235
burn-up depth of 25%, the SM reactor can operate without recharging over a period of 60 to 65 days. However,
from the physical point of view, it is advisable to replace some of the individual burned-up slugs with fresh ones.
In this case, the U235 content in the core will change very little during operation and the rate of the fuel burn-up
will be maintainedat the approximately average level, which is equal to 12.53!o. At the same time, the duration of
S. M. Feinberg, et al., Proceedings of the Second International Conference on the Peaceful Uses of Atomic Energy
(Geneva, 1958). Reports by Soviet scientists; Nuclear Reactors and Nuclear Power Engineering [in Russian) (Atomizdat,
Moscow, 1959) Vol. 2, p. 334.
410
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the reactor run without recharging is determined by the capacity of the box with spare slugs, which is placed in-
side the reactor. The capacity of the SM reactor box secures a continuous operation over a period of 40 days.
Moreover, an intermediate reactor is much less subject to contamination by xenon, and it makes, it possible to
produce neutron fluxes witAifferent energy spectra.
The reactor cc,:r nsists of slugs with fuel element plates containing 90%-enriched uranium. The lateral
reflector is made of beiyilium oxide. The lateral reflector size was determined by taking into account the neces-
sary arrangement of all experimental devices in sufficiently large fluxes of thermal and fast neutrons. The total
volume of the lateral reflector is 485 liters, while the volume of the experimental devices amounts to 43 liters,
and the volume of the cooling water flowing through the slits between the beryllium oxide bricks is 17.5 liters.
The critical mass of the reactor without the experimental channels is 7.3 kg of U235, and the critical mass
when the experimental channels are taken into account is 9.5 kg of U235. The reactor charge for the medium
burn-up rate, when the core consists of 20 slugs with different burn-up depths, is ???41.7 kg. For a nominal power
of 50 Mw, the average energy intensity is 2100 kw/liter, and the maximum thermal flux emitted from the surface
of the fuel elements is equal to 5.5'106 kcal/m2.hr.
In order to secure reliable heat dissipation for such large thermal fluxes, it was necessary to raise the pressure
in the reactor vessel to 50 atm. The cooling water velocity in the core is equal to m/sec for an over-all cool-
ant discharge of 2000 t/hr in the first loop. Under these conditions, the maximum temperature of the fuel element
surface does not exceed 195?C.
The Reactor Experimental Channels
For experimental purposes, the SM reactor is provided with five horizontal, one inclined, and 15 vertical
channel's.
The horizontal channels are located in the plane of the central reactor core cross section (Fig. 2 and 3). The
channels are extended to the physical measurement room. Each channel is provided with a protective device,
which makes it possible to make preparations for the experiments while the reactor is in operation, The inside
space of the channels, including the protective devices, can be evacuated if necessary. The neutron flux at the
channel exits is equal to ~3-10" neutron/cm;sec.
With the exception of the central experimental channel, all the vertical channels are located in the reflec-
tor at different distances from the core center.
The following vertical channels are installed first.
1) Three channels for the production of transuranic elements, one of which (with a diameter of 90 mm) is
placed at the core center, and the other two are placed in the reflector. The channels are cooled by water which
is under a pressure of 50 atm and whose temperature is 60? C. However, the technological layout provides the
possibility for the channel operation at higher temperatures and pressures.
2) Two low-temperature channels with a water coolant for metallographic investigations. The channels have
a special cooling loop with water at temperatures from 30 to 80? C and under pressures of up to 50 atm.
3) Two high-temperature channels with a water coolant for the testing of fuel element specimens, investiga-
tion of water chemistry problems, and for studies of the corrosion of structural materials. The channels have a
special cooling loop with the following water parameters: temperature at the channel upstream end: up to 400?C;
pressure: up to 350 atm; water discharge through a single channel; 30 t/hr.
4) Five channels with a gaseous coolant for studying the behavior of fissionable and structural materials in
fluxes of fast and thermal neutrons and 7-rays in the temperature range from 0 to 600? C. In order to improve
the removal of heat from the specimens and to eliminate the effect of the medium, high-purity helium is used as
the coolant. All the channels are combined into three independent loops, which makes it possible to perform
simultaneous experiments under three different regimes. The helium pressure in the loops is from 30 to 50 atm,
and the gas discharge rate in each channel attains 350 kg/hr. The channel design makes it possible to unload the
specimens under investigation without shutting down the reactor. Each of the channels is provided with hermetically
sealed electrical lead-outs for connection to measuring instruments.
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1,1b i.
MECHANISM
!II !MUNI III Ir 41111
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reflector; 2) control rods; 3) internal reloading
mechanism; 4) reserve cells for slugs; 5) protect-
ive screen; 6) reactor vessel; 7) lid; 8) packing-
gland lead-out of the reloading mechanism; 9)
packing-gland lead-out of the control mechanism;
10) drive of the emergency protection mechanism;
11) horizontal experimental channel.
5) A channel cooled by gaseous helium or liquid
hydrogen,which is used for studying the behavior of mater-
ials at low temperatures.
6) A channel with a gaseous ceolant for the testing
of specimens at temperatures of the grder of 2000? C.
7) A channel with a liquid-metal coolant whose
temperature is approximately equal to 1000? C, which is
used for studying the behavior of coolants and testing
fuel element specimens.
Reactor Design
From among the specific requirements which main-
ly determined the reactor design, the following should be
mentioned:
1) provision of a small-size core which would be
capable of withstanding very large thermal loads over a
long period of time;
2) provision of adequate core cooling for the above
conditions;
3) placement of a maximum number of experimen-
tal channels in the immediate vicinity of the core as well
as in the reflector;
4) provision of facilities for unloading the slugs with-
out lowering the pressure in the system.
' In designing the reactor, special attention was paid
to the calculation of the carrying structures in the immedi-
ate vicinity of the core, which, while subjected to exter-
nal loads (proper weight, weight of the mounted parts, and
pressure drops), simultaneously experience considerable
thermal stresses due to the internal radiation heat release.
The general reactor layout is shown in Figs. 2-5.
The reactor vessel contains the core, the reflector, the
protective screens, the control system rods, the mechanism
for the internal reloading of slugs, and a number of carry-
ing structures.
The slug cells are located in a square space with a
surface area of 420x 420 mm, which is separated from the
reflector masonry by zirconium sheets. Instead of the
four corner cells in the square, guiding zirconium tubes,
inside which the emergency protection rods move, were
installed. The remaining 32 cells can be occupied by
slugs, which are fixed by their stems in the support slab.
During the reactor operation, several of the cells
can by occupied by beryllium oxide blocks or remain
free. Thus, for instance, if four slugs are withdrawn from
the core center, a water-filled space with an area of
140X 140 mm is formed, where the central vertical channel
can be inserted.
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MI
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Cells for the storage of fresh slugs and the temporary storage of spent slugs are provided above the reflector.
The slugs are transferred by remote handling by means of the internal reloading mechanism without lowering ,the
pressure in the reactor. This mechanism is mounted under the reactor lid and it has two mobile rods, each of
which serves one half of the core. After all the spare slugs are used up, the spent slugs are discharged through an
inclined channel into the storage space. Two television cameras are provided for observing the reloading in the
reactor.
Reactor Vessel and Lid. The reactor vessel (see Fig. 2) is welded; its cylindrical part is made of Stainless
Steel 1Kh18N9T sheets,which are 36 mm thick. At the level of the middle core cross section, five horizontal
branch pipes for channel mounting are welded to the reactor vessel. Two inclined branch pipes, one of which is
intended for the unloading of slugs and the other for mounting the channel with the liquid-metal coolant, are in-
stalled at an angle of 90? (plan) above the reflector. Eight branch pipes which serve as the water coolant inlets
and outlets are welded to the forged flat bottom. Eight branch pipes which serve as the lead-ins for the drive
shafts of the compensating rods and the automatic regulator rods are mounted in the upper cylindrical part of the
vessel. Ten nipples, which serve for the outlet of pulse tubes of the system for controlling the activity of the
water leaving the slugs, are welded to the same part of the vessel. The vessel lid was made flat in order to facili-
tate the arrangement of experimental channels on it. The packing between the vessel and the lid consists of a
thread-like gasket. Several openings of different diameters were made in the lid for the experimental channels,
control rod drives, shafts of the internal reloading mechanism, for the charging with fresh slugs, and for the lead-
outs of television camera cables. The unused openings are closed by means of plugs.
The wall thickness of the vessel shell was calculated with respect to the portion in the vicinity of the core,
where the vessel strength is weakened by the five welded horizontal channel nipples and where the thermal stresses
are at the maXimum. Moreover, the spots at the joint between the cylindrical portion and the bottom with the
flange were investigated on an optical model. These investigations made it possible to find the optimum joint
shape.
In making the vessel, special attention was paid to securing a coaxial position of the horizontal branch pipes
with respect to the openings in the screens and the bushings in the reflector as well as a coaxial position of the
cells in the reflector with respect to the openings in the lid. In order to satisfy these requirements, a special arrange-
ment simulating the reactor inside space was used at the plant where the reactor was built.
413
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Fig. 4. Top view of the reactor lid.
Fig. 5. Reactor section through the horizontal channels.
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Top view
Cross section B-B
tzis).
2.)
g416"-6T
1) Fuel element; 2) slug jacket; 3) spacing
rack; 4) stem; 5) holding head.
,s-ftl- A--ss e,;,?%1 A 6^' ' 7')
L F44 kij
Reflector. The reflector (see Fig. 2) forms a structural unit
which includes the beryllium oxide brickwork, a slug holder plate,
?and a pulse tube system for controlling the water activity at the
outlet from the slugs.
The beryllium oxide brickwork is enclosed in a shell, which
is made of stainless steel 25 mm thick. The reflector shell serves
at the same time as an additional reactor vessel shield. In order
to secure compact packing, the entire reflector brickwork was
divided (in the plan) into eight segments, which were separated by
means of zirconium sheets.
A large amount of heat is liberated in the reactor reflector
(,.40 w/cm3 in the first layers) due to the absorption of y-radiation
emanating from the core. For this reason, the dimensions of each
block were determined by taking into account the maximum atibw-
able thermal stresses in these blocks. These conditions as well as
the presence of 24 vertical and five horizontal channels in the re-
flector led to a great diversity of reflector block shapes (approxi- ?
mately 65 types). The upper and the lower grids between which
the blocks are installed are made of thin stainless steel sheets,
which are reinforced by ribs.
The slug holder plate consists of 32 cylindrical bushings,
which are assembled between two sheets. The bushings, the spac-
ing between which is equal to 70 mm, have a bored collar with
exact tolerances at the lower end and a conical outlet at the upper
end. The positive placement of slugs is secured by the accuracy
in making the slug stems and the holder plate, which is fixed on
a cylindrical bushing, which is welded to the bottom of the reflec-
tor shell. The system of pulse tubes for controlling the water acti-
vity is fixed on the holder plate from below.
In order to reduce the number of pulse tubes which are ex-
tended from the reactor vessel to the data transmitters', the coordi-
nate system of water sampling was used. The operating principle
of this system consists in sampling'water from each row of slugs
into collectors, the number of which corresponds to the number of
rows. Another row of collectors, whose arrangement is similar to
that of the first row, is installed perpendicularly to the first row.
Thus, if a fuel plate is damaged, the water with higher activity
enters two mutually perpendicular collectors; the damaged slug
is directly determined with respect to the designation numbers of
these two collectors.
Slugs (Fig. 6). The fuel element type and the design of the
jacket for the fuel elements were chosen with the aim of securing
a compact core with the greatest possible heat exchange surface.
The fuel element (Fig. '1) is of laminar shape and is prepared according to the following method;
1) The core is pressed from uranium oxide powder and electrolytic nickel;
2) The core is inserted in the nickel frame and is enclosed on both sides by nickel sheets which are welded
to the frame along the entire contour;
3) The frame and the core are hot-rolled in an inert-gas medium until the assigned dimensions are obtained;
in rolling, the nickel in the core is diffusion-welded to the covering sheets and the frame.
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Fig. 7. Fuel element. 1) Fuel (core);
2) frame; 3) covering.
The design of the fuel element and the technology used in
its manufacture provide a reliable thermal contact between the
heat exchange surface and the granular uranium oxide.
After rolling and cutting, the final dimensions of a plate
are the following:
Thickness, mm
0.8
Length, mm
280
Width, mm
33.4
Core thickness, mm
0.5
Core length, mm
250
Core width, mm
29
Over-all plate weight, g.
60
U 2 '3 5 weight, g
12.5
The plates are assembled in the slug jacket. Each slug con-
tains two rows with 54 plates in all, which are spaced by means
of racks at the top and bottom and by means of corrugated strips
made of stainless strip steel 0.1 mm thick along the entire length. The slug jacket (see Fig. 6) is made of two
halves, which are connected by spotwelding. Each half is made of a stainless steel sheet 0.5 mm thick. At the
upper end, the jacket is welded to the rack, which is provided with a holding head, and, at the lower end, the
jacket is welded to a cylindrical stem with gradual transition to a square cross section.
Since the fission density of uranium nuclei at the interface between the active zone and the water-filled
space sharply increases, the end plates in some slugs contain a smaller amount of uranium in order to equalize
the heat release. Thus, three types of slugs, which differ from each other by their end plates, are placed in the
reactor. In order to avoid possible errors in placing them, the slug stems are provided with grooves which are
oriented with respect to the plates with a lower uranium content, and the holder plate is provided with pegs which
fit the stem grooves.
The correctness of plate mounting, the spacing method, and the jacket rigidity were checked by experiment.
At the same time, the plates were tested in the reactor loop under conditions close to the actual operating conditions.
Moreover, the problem of the heat transfer in narrow slit channels was studied, as a result of which the necess-
ary data on the wall temperatures and the critical thermal loads were obtained.
Protective shields. The reactor vessel, which is stressed by the internal pressure, is protected from intensive
neutron and y-ray fluxes by the reactor shields (see Fig. 2).
The thickness of the shields and of the reflector shell was chosen with the aim of securing a sufficient flux
of neutrons beyond the reactor vessel for a normal operation of the ionization chambers and of keeping the ther-
mal stresses in the vessel below the maximum allowable value. The shields have a cylindrical shape; the inside
cylinder is welded to a support ring, and the outside cylinder is fastened to the lower adapter. The inside shield
cylinder has a flange, on which the reflector shell is mounted by means of the support ring. The end of the out-
side cylinder is located above the core. A receiver for spent slugs, cells for the storage of fresh slugs, and an ad-
ditional shield, which protects the upper reactor vessel part from radiation, are fastened on the cylinder upper end.
Slabs which protect the vessel bottom from radiation are installed on the lower shield support ring.
The shields also serve for dividing the reactor inside volume into two zones. This division is secured by
closely fitting the lower portion of the shields to the vessel bottom. As a result of this, two volumes are formed:
a ring-shaped volume with the four branch pipes for the water coolant supply and the central volume with the
welded four outlet branch pipes.
Water coolant circulation in the reactor. The water enters the ring-shaped clearance formed by the shields
and the reactor vessel through four branch pipes, which are welded to the reactor bottom. In the ring-shaped clear-
ance, the water supply is divided: one portion is directed under the reflector, where it cools the beryllium oxide
blocks, and the other portion passes between the vessel and the shields. At the top, both water streams meet, and
the water flows downwards between the slugs, through the clearances between the demountable beryllium blocks,
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-.9pr
and along the emergency protection rods. The water then passes through the clearances in the lower protective
slab and comes out of the reactor through four central branch pipes, which are welded to the vessel bottom.
Since a certain given amount of water must'be transmitted through the reflector, adjusting baffles, by means
of which the water flow is regulated during preliminary adjustments, are fastened' to the upper portions of the
shields. During the adjustment operations, the water flow is controlled by means of a flowmeter, which is connect-
ed to two pulse tubes, one of which is installed upstream from the, reflector water coolant inlet and the other is
installed downstream from the outlet.
The presence of fitting spots between the water coolant inlet and outlet creates the possibility of water
leakages. In order to minimize the leakages, all the fitting spots (between the shields in the vessel, between the
reflector shell and the shields, and the spots where the slugs are placed) are carefully adjusted and individually
checked after final assembly.
The Control System
The reactor control system includes two automatic regulators with coupled rods, four compensating rods,
and four emergency protection rods, which can also be used as compensating rods.
Thirteen ionization chambers, which are placed outside the reactor vessel, serve as the control system data
transmitters.
The automatic control system maintains the assigned reactor power level within the 0.5-100% range.
A rack and pinion drive is used as the final mechanism of the automatic regulator rods. From the servomotor
and the reduction box, which are mounted on a platform above the reactor, the rotation is transmitted by two
shafts with universal joints to-the two pinions, which drive the racks with the rods. Packing glands with an arrange-
ment for collecting the leakage flow and directing it into a hermetically sealed draining system are provided at
the spot where the shafts enter the reactor. The maximum speed of the rods is 40 mm/sec, and the rod stroke is
450 mm.
The emergency protection system consists of three independent channels in the electrical circuit, and it
operates if the power level exceeds the nominal value by 25%. Moreover', the emergency protection system com-
prises a preventive protection arrangement, which is activated if the power level, which is maintained by the
power control device,.is exceeded by 10-15%. On the reception of this signal, the automatic regulator rods are
introduced into the core. The action of the preventive protection system stops if the power level drops by 2-7%.
The emergency protection servomechanisms are designed to secure a quick trip-out of the rods for any posi-
tion they may occupy. This provides positive emergency protection, and, at the same time, makes it possible to
use the rods for a partial compensation of the reactor reactivity. The emergency protection drive comprises a
motor which rotates at low speed. The motor rotor is rigidly connected to a ball nut. A screw with a core made
of magnetic stainless steel at its end moves inside the nut. The lower motor flange is fastened to the frame of
the power and signal coil system. A magnetic core, which is connected to a pole, at the end of which the rod
is suspended, is located inside this system. When the power coil system is energized, the resulting cohesive force
between the cores vanishes, and the rod falls into the core under the action of its own weight and the pressure
drop. At the end of the falling period, the braking device is activated.
The over-all rod stroke is 400 mm, and the reactivity compensation section is equal to 150 mm.
The data transmitters for the emergency protection signals consist of three ionization chambers, which are
surrounded by lead shields.
The design of the final mechanisms of the four compensating rods is similar to that of the automatic regula-
tor rods. The rod speed is 1 mm/sec, and the rod stroke is 450 mm.
Two types of rods of different shapes are used in the reactor. The emergency protection rods have a square
cross section, and the others have a round cross section. Each rod has an upper and a lower part. The upper (ab-
sorbing) part of the rod consists of a cadmium tube, which is enclosed in a stainless steel jacket. The tube is
filled with water. The lower part of the rod is made of beryllium oxide. Thus, in introducing the rod absorbing
part into the reflector, beryllium oxide is replaced by the cadmium tube, which is filled with water. The efficiency
of this type of rod in a reactor with intermediate neutrons is sufficiently high.
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C-"Cebro..v?
4 8---
- -
51
LeaTiq
Er% rt
VERTICAL
EXPERIMENTAL
HANNEL
"PIPING OF THE
I.VERTICAL
?CHANNELS
HORIZONTAL -
EXPERIMENTAL
'CHANNEL
SPACE, FOR
THE PROTECTIVE DEVICES
OF THE HORIZONTAL CHANNELS
Rea444E-pes414efiwinside-the-slineldv-.1) Reactor; 2) vertical experimen-
tal channel; 3) horizontal experimental channel; 4) unloading mechanism;
5) piping of the vertical channels; 6) rotary disk slab; 7) top protective slab;
8) thermal shield; 9) reactor room floor; 10) space above the reactor; 11)
space for the protective devices of the horizontal channels.
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Under start-up as well as operating conditions, the reactor power measuring system makes it possible to
measure the power level on a linear and a logarithmic scale, to record the power level, to vary its rise time, and
to provide sound signalization. It is considered that the lowest measurable reactor power level is 10-50/0.
The Reactor Shielding
The reactor location in the shield is shown in Fig. 8. The reactor rests on a support slab inside a concrete
pit. A passage, through which the water coolant piping passes,leads to the lower portion of the pit.
In order to secure the maximum intensity of neutron fluxes at the beam exits, the shield near the location of
the horizontal channels is made as heavy as possible;' it consists of air-cooled steel shields and heavy concrete.
The thickness of the protective shields was determined by taking into account the maximum allowable tempera-
ture drop in concrete. The air in the space between the reactor vessel and the thermal shield is rarefied to such
a degree that the active air does not penetrate into the space above the reactor and .into the steel shield cooling
system.
The horizontal channels at the outlet from the thermal shield are made airtight, whereby the space where
the protective devices are located is separated from the reactor space.
The top of the reactor emerges into the room above the reactor (Fig. 8). This space contains the control
system drives, the piping of the experimental channels, and the channels of the ionization chambers, which are
fastened to the shielding blocks that separate this space from the reactor space. Moreover, the space above the
reactor contains cable conduits leading to the electric motors, ionization chambers, leakage detection data trans-
mitters, and the experimental channel data transmitters.
The space above the reactor is separated from the reactor room by'means of a shielding slab, and, above the
reactor itself, this space is separated from the reactor room by a rotating cast iron disk with several interchangeable
plugs, which are located above the experimental channels. The placement and withdrawal of the specimen's under
investigation is effected by manipulation from the reaCtOr room, while the disk does not have to be rotated for this
purpose. The disk is rotated only in the case where it is necessary to withdraw a channel or to perform work on
tightening the reactor lid packing. All the equipment not connected with investigation work is contained in the
space above the reactor, which makes the experimenters' work safer and more efficient.
' ? ,. ?
Adjacent to the space above the reactor is the liquid-metal loop adjoining basement, the well containing
the mechanism for unloading spent. slugs, and the storage room for these slugs, which is filled with water. The
unloading mechanism is connected to the reactor through. an.inclined tube, which is covered by means of a de-
mountable lid. .?.
The slugs are withdrawn from the reactor in the following manner: the pressure in the reactor is released,
the inclined tube lid is taken off, and the unloading mechanism, which aligns the container with the inclined
tube, is activated. The adapter for gripping the cases is,then,lowered; the adapter pulls the slugs out of the reac-
tor and hauls them into the container, after which the container rotates until aligned with the inclined channel
connecting the mechanism frame with the storage room, and the slugs are lowered and placed into the storage
cells. The entire process.of unloading the slugs takes place under a protective layer of water.
The Reactor. Cooling Layout '
The water temperature' at the reactor 'upstream end is 50?C; at the downstream end, it is 80?C.
The reactor Cooling system comprises four independent loops, each of which includes a hermeti-
cally sealed 'circulation pump with a capacity of 500 a hr, Which produces a pressure of 10 atm,
a 15-Mw heat eichanger, and a. purifying filter. Two of-the four pumps are connected to the emergency-feed
line, and, in the case'of electric power failure, the power for these pumps is supplied from a storage battery.
Moreover, the layout comprises a loop for emergency reactor cooling. This loop is designed for 'maintaining a
steady water circulation during the time required for witching the electrical power supply lines of the main
pumps to the storage batteries, and it includes a.hermetically sealed pump and a heat exchanger. The pump is
in constant operation and is fed from a storage battery.
For continuous elimination of radioactive admixtures fromithemater and for removing detonating mixtures,
the layout includes ionexchange filters and ,a ,bypass system with contact apparatuses,.which are calculated for
the recombination of 20.10,7 norm ,liters of detonating mixture._
The system also cOntains volume compensators. The pressure in the loop is maintained by means of helium,
which is fed to the compensators through a pressure-reducing regulator from a tank system. In releasing the pressure,
helium is admitted into the receiving tanks.
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NEW IDEAS IN THE STRUCTURAL DESIGN AND LAYOUT
OF NUCLEAR REACTORS
A. N. Komarovskii
Translated from Atomnaya Energiya, Vol. 8, No. 6, pp. 505-513, ,
June, 1960
Original article submitted January 28, 1960
This article discusses methods used in siting reactor buildings, and their structural layouts. The advantages and
the disadvantages of the above-ground and below-ground approaches in the construction of reactor buildings are
considered. Current techniques for building concrete shielding around stationary power reactors are described.
Detailed treatment is given to problems in the use of explosion-proof and missile-proof protective shells, and
problems concerning the economics of the use of various compositions of concrete for biological shielding of re-
actors are elucidated.
Reactor Siting Methods
Siting of reactors below ground level has been a common approach to reactor siting up to the present. As
some examples of such comparatively costly and laborious installations, we may cite the reactors of the World's
First Nuclear Power Station (USSR) and the Shippingpott nuclear power station (USA). This solution is owed to the
effort to secure greater safety for the population of the surrounding areas in case of damage to the reactor. Under-
ground excavation of reactor sites has been practiced primarily when building new reactors of untested design, and
especially in cases where pressurized water was employed as coolant.
At present, most reactors are being built above-ground, with the advantages, consisting, as a rule, of greater
simplicity in design, lower construction costs, and shortened construction time, resulting from elimination of the
need to excavate deep foundation pits or build foundation walls designed to withstand heavy earth pressure, not
to mention being leaktight against passage of ground water. According to the data reported by Bergstrom and
Chittenden, in reference to the design of the experimental power reactor EBWR, it was found that excavation
costs increase the total cost of the reactor installation by as much as 7010 [1]. Inspection and overhaul of the reactor
is much simpler when the reactor is built above-grade. It should also be noted that in the case of reactor damage
and attendant leakage of radioactive fluids, there is a great hazard of these contaminants getting into the surround-
ing earth in the case of a reactor recessed below ground level. The above-grade position of research reactors
facilitates convenient placing of all the experimental equipment servicing the reactor. The above-grade variant
is mandatory for horizontal reactors, inasmuch as removal of fuel elements from a below-ground horizontal reactor
would require an underground reactor hall of enormous size.
The present shying away from underground installations is also explained by the increased reliability of the
control systems of reactors, and likewise by the experience accumulated in operation of reactors of many types.
It is very seldom that reactors are built completely below ground level, since this would entail greatly in-
creased costs and much more troublesome operating conditions. Below-grade siting is motivated primarily by a
striving to secure airtight shielding of the object. However, in Scandinavian countries and in Switzerland, with
their great knowledge of tunnel and excavation engineering, construction of underground reactor facilities is still
being planned and carried out. Examples representative of such reactors are the Norwegian heavy water reactor
at Halden [2, 3], and the Swedish power reactors R-1, R-3,"Adam", and "Eva" (Fig. 1).
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A unique approach to exploitation of the advantages in underground siting is to be noted in the draft project
for a large-scale power reactor [4] drawn up by Swedish engineers. The steel pressure vessel of the reactor, capable
of withstanding an internal pressure greater than 10 atm, and having relatively thin walls, is housed inside a below-
grade pressure-tight chamber such that the pressure inside the pressure vessel is transmitted through the gas flooding
the space between the walls of the pressure vessel and the chamber enclosure to the surrounding rock. The walls
of the pressure vessel actually play the role of an intermediary membrane in this setup (Fig. 2).
C
1
r
111?.
.--
_ __?___,,w,m1 bilifillio '
-:---.44---
Fig. 1. "Eva" underground reactor installation (Sweden): 1) Heat-exchanger shell;
2) fuel-element charge-discharge; 3) fuel-element storage house; 4) expansion
tank; 5) tunnel connecting reactor building to turbine house; 6) turbine house.
The Swedish reactor designers drew the following conclusions from their experience in planning and build-
ing underground reactor facilities [4].
The advantages of such reactors are:
1) the installation built below-grade will be capable of withstanding excess internal pressure concurrent with
reactor damage, without any additional complex safety measures;
2) these facilities can be reliably protected against wartime destruction;
3) the surrounding rock body provides excellent biological shielding;
4) leakage of radioactive gases out into the surrounding medium in the event of damage is significantly
diminished by the filtering action of the overlying rock and soil layers, and in consequence of the decay of radio-
isotopes during their traversal of the walls;
5) the underground siting is to be preferred at locations where it is important to preserve the natural landscape
intact.
The disadvanlages inherent in underground reactor installations are: ?
1) the need for an appropriate rock cover, which narrows down the available freedom of choice for the site;
2) the need for careful inspection of ground-water level and ground-water movement, with resulting addition-
al expenses;
3) inconveniences due to the limited size of the installation.
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Liljeblad and Madsen [4] note that reduction of costs in below-grade excavation work in Sweden in recent
years has had the effect of appreciably offsetting the more protracted construction schedule required for under-
ground construction, compared to above-grade work. According to them [4], a tunnel 20-50 m2 in cross section
and 100 m in length can be driven in a month and a half; a rock chamber 20,000 m3 in volume can be blasted
out and excavated in 7 to 10 months. Swiss specialists submit pretty much the same considerations in their argu-
ments favoring underground nuclear electric power generating stations.
Fig. 2. Cross-section diagram;
Swedish plan for an underground
power reactor.
On the basis of the experience accumulated in underground construction
of reactors in Sweden, it is felt that compartments of the order of 25 m may
be safely excavated in rock where the roof thickness is 40-50 m. For roof
thickness in excess of 50 m over chambers of cylindrical cross section, com-
partments 35-40 m can be safely tolerated. Important attention is given to
providing watertight concrete lining for the tunnels and excavated chambers.
Concrete lining of the walls of underground rock excavations may be
achieved by placing concrete or by applying gunite with a sand-cement grout.
The advantage to be gained in discrete upright foundation walls for the under-
ground installation is their accessibility from either side, and the possibility
of using the voids for additional drainage of ground water. The disadvantage
resides in the increased hazard of damage to the walls if the reactor should
sustain an injury of an explosive nature. Gunite injection is a lot cheaper
than concrete lining. However, this incurs a penalty of much thinner density
of lining, not to mention the uneven surface,which is highly undesirable from
the standpoint of deactivation operations (particularly after possible damage,
with radioactivity release, has occurred).
It is considered that no concrete lining of any kind is capable of being
completely gas-tight. In the Swedish reactors, a concrete-lined chamber
with 5000 m2 surface area was found to leak 0.01-0.02 m3/sec air at a pressure
of 1 atm. If specifications for a gas-tight lining are very exacting, a lining
of sheet steel or plastic will be required.
According to Swedish data, at roof thickness of 20 m and an excavated
span of 20 m, rock slides may occur at an overpressure of 15-20 atm, which
is 5-6 times greater than the usual overpressure which is designed for.
It is the prevailing opinion in the USSR that primarily because of cost considerations and the compressed
construction schedules, recourse may be had to underground siting of reactors only where already existing excava-
tions are available (e. g., excavations completed previously for some other purpose, but not now in use), or else
mining pits or natural caverns may be utilized.
Some Problems in the Construction of Reactor Enclosures.
Two-echelon reinforced concrete shielding has been resorted to in some of the recent reactor power instal-
lations. The inner shielding ring contacts directly with the reactor proper; the outer echelon encases the pipes
and conduits leading to the steam generating units. This was the decision taken, e. g., in the design of the British
nuclear power stations at Hunterston and Hinkley Point (Fig. 3). In individual cases, the outer echelon of the shield-
ing also reaches to the steam generators Je. g., in the U S "Yankee" nuclear power station, see Fig. 4).
In double-echelon shielding, the thickness of the outer echelon is usually about 600/0 of the thickness of the
inner shielding. This facilitates overhaul of the radioactive piping and steam raising units (when they are shut
off) without shutting down the reactor. When the heat exchanger units are separated by baffles, the possibility of
closing off and overhauling the individual heat exchanger units without having to interrupt the operation of the
reactor is secured.
The lateral concrete biological shielding of reactors, heat exchangers, and the shielding walls of the reactor
buildings are placed in monolithic form, as a rule. An exception is presented only by some small experimental
reactors where the latter and top shielding is in the form of a module of removable concrete blocks.
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Fig. 3. An example of double-echelon shielding in
the reactor building of the Hinkley Point nuclear
power plant: 1) reactor; 2) inner ring of biological
shielding; 3) outer ring of biological shielding; 4)
gas duct; 5) steam generating unit.
The striving to avoid such sectionalized concrete-
block shielding, or combinations of sectionalized and
monolithic shielding (i. e., with subsequent concret-
ing of the arrays of blocks) in the biological shielding
of large reactors is to be explained first of all by the
danger of crevices forming between the blocks, par-
ticularly in the outer extended zones of the shielding
walls, which are subjected to heat on one side only.
However, it has by now become established that local hairline cracks in concrete exert no material effect on the
efficiency of biological shielding. This accordingly provides justification for making wider use of modular block
shielding and combined shielding designs, facilitating in great measure the industrialization and speeding up of
reactor construction work. The wall of one face of the building housing an experimental power reactor now being
constructed in the USSR will be assembled in modular fashion, and will consist of reinforced-concrete blocks
measuring 5.4 by 2.4 by 1.5 m (Fig. 5). The weight of one block is 46.5 tons. The blocks present a broken-line
configuration seen in vertical cross section; they are lined with stainless steel on the side facing the reactor core.
Combined modular-monolithic constructions of shielding enclosures (e. g., used for nuclear-reactor enclosures)
of the type appearing in Fig. 6 have been gaining favor in the building of more recent reactor installations in the
USSR. Large constructions in the form of assembled reinforced-concrete columns, girders, and slabs using standard-
ized prefabricated modular components to maximum advantage are employed in the building housing the experi-
mental power reactor, as well as the building housing the reactor of the Novo-Voronezh nuclear electric power
generating station. Such construction designs obviate any need for timber, scaffolding, formwork for placing con-
crete, etc., and greatly facilitate industrialization and speed-up of a construction work on enclosures.
While on the subject of interchangeable reactor shielding components, we might mention that the system of
constructing the top shielding of several layers of reinforced-concrete slabs or blocks with layers of other materials
(lead, etc.) sandwiched in between, popular in the past, is more recently being crowded out by the tendency, fully
Justified by operating conditions, of arranging the top shielding in the form of detachable large "plugs," comprising
a metal sheathing encasing a concrete "meat." The weight of each of the constructions amounts to 20 tons and
higher.
Fig. 4. An example where the steam generators are
located inside the outer shielding echelon ("Yankee"
nuclear power plant): 1) inner shielding echelon; 2)
outer echelon; 3) reactor; 4) steam generator; 5)
containment enclosure; 6) support column.
Applications for Various Concrete Mixes
Data [5] proving the unfeasibility of employing special heavy concretes as biological shielding, except for
those cases where it is absolutely imperative to keep shielding thickness within bounds. Others [6-8] assert that
the use of metal aggregates and concrete of weight by volume exceeding 4.8 tons/m3 should be restricted to those
cases where the high cost of the 'concrete is offset by the total savings achieved as a result of minimizing shielding
thickness. Horton [9] notes that a concrete shielding arrangement using aggregates drawn from local quarries is al-
most always cheaper than using heavier coarse aggregates transported from afar. The curves seen in Fig. 7 demon-
strate that with the exception of small reactors) the use of conventional concrete optimizes costs.
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3 1 2 4
Fig. 5. Modular
design biological
shielding for ex-
perimental power
reactor under con-
struction in the
USSR: 1) space
for reactor core;
.2) concrete blocks; -
3) stainless steel
lining.
?
Air.:::::53t7ZeZMNZAVVR
Fig. 6. Layout of the shielding enclosure (cross sec-
tion through the girders): 1) monolithic concrete;
2) supporting layer of prefabricated reinforced-con-
crete beams; 3) temperature-expansion seam; 4)
top layer.
As for applications in reactors enclosed within steel containment
shells, a paper (10) based on the experience of the USA's Stone and
Webster Engineering outfit, engaged in the design and construction of
shielding enclosures for nuclear reactors, mentions that the cases con-
sidered by this firm involving the use of high-density concrete as biologi-
cal shield with a consequent reduction in shielding dimensions brought
about economy in steel costs which more than covered the additional
outlays for the high-density concrete. ,However, there are no published
economic calculations or data available on this question at present, and
the significant difference in costs of conventional as against special
heavy concretes in the USSR compared to prices prevailing in the USA
and the UK give us reason to doubt that this view will retain its validity
for conditions in the USSR as well.
Until recently, concrete with limonite ore (brown hematite) having a high content of chemically bound
water, higher than that characteristic of conventional magnetite and hematite ores (10-17% as against 1-4%), as
aggregate has been used in some particular cases for biological shielding. The idea was to bring about a substantial
increase in the efficiency of shielding against neutrons. Calculations
performed at the Institute of Atomic Energy have shown that to get
the same shielding efficiency against neutrons from the use of con-
ventional concrete, the shield thickness would have to be 20% greater
bo than when limonite concrete is employed.
:q
10-3
0
0 No /04
bo
1000
t70
?e4
"0
0
100 200 300 400 500
Core radius, cm
Fig. 7. Weights and costs of a representative
set of shielding concretes (data from (8D: 1)
barytes concrete (3.054 g/cm3); 2) concrete
with pig-iron punchings(5.3 g/cm3); 3) con-
ventional concrete (2.5 g/cm3).
424
One of the designing institutes in the USSR carried out, for
the purpose of a feasibility evaluation of the use of limonite con-
crete as against ordinary concretes as biological shielding, a pro-
jection and comparison of estimated construction costs for a large-
scale reactor installation using shielding of ordinary concrete and
using the variant of limonite concrete shielding. The comparison
was carried out with reference to different regions of the USSR.
The actual prevailing conditions in the production of limonite Ore,
transportation of the ore by railway to the construction site, and
practical costs of locally procured concrete aggregates were all
taken into account in the estimate. The calculations showed that
the use of limonite concrete for shielding in large reactors results
in the following cost increases in shielding for the reactor buildings
(percentagewise):
For centrally located regions of the USSR by 12010
For the mid-Ural belt by 15
For Western Siberia by 128
For Eastern Siberia by 132
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This comparative estimate permits the inference that from an engineering and costs standpoint, the use of
limonite concrete for biological shielding of reactors in the USSR will result in over-allhigher construction costs.
The use of concretes incorporating limonite or hydrohematite ore aggregates might be justified only in case the
reactor is being built in the immediate vicinity of the location where those ores are mined.
Compacted dried iron ore is worthy of mention as a low-cost material for horizontal biological shielding
placed in reinforce:1-concrete laminates limited to the thickness specified in the design. In the shielding assembly
of one of the buildings for the proton synchrotron (synchrophasotron) of the Joint Institute for Nuclear Research at
Dubna, consolidated dry magnetite ore from Krivoi Rog, with a dry volume by weight 2.62 tons/m2 at a thickness
of 0.9 m for the shielding layer and a total volume of 500 tri3 was used. The ore composition included 85% mag-
netite crushed rock and 15% magnetite sands. The components of the shielding were laminated into layers 15 cm
thick by a 10-ton road leveler. Less tractable lumps were consolidated by using electrically driven tamping rods
weighing 200 kg each. The weight by volume of the consolidated magnetite mix was 3.3-3.5 tons per cubic
meter. This resulted in savings of 588 thousand rubles compared to the previously planned shielding assembly of
heavy-concrete blocks (weight by volume 3.2 tons/m3). The shielding cost per cubic meter was cut from 933
rubles, according to the first cost estimate, to 464 rubles. Placing of a layer of consolidated iron ore may thus
be found useful in horizontal biological shielding units, not loaded statically, for nuclear reactors.
use of Protective Shields Against Missiles and Radioactivity Release
Despite the fact that no important accidents of an explosive nature accompanied by the release of radioactive
liquids, gases, or fragments have occurred in the actual operation of reactors, most large-scale foreign reactors
have protective containment shells, mostly metal structures,incorporated into their design. The building of these
containment enclosures,despite the appreciable increase in cost and complexity of the reactor installations involved,
is due to the effort to ensure the safety of the area surrounding the reactor plant in the case of a theoretically pos-
sible accident involving explosion of the reactor core or a rupture of the lines carrying radioactive and pressurized
liquids or vapors.
The problem of the use of protective missileproof containment shells has been the subject of theoretical and
design-engineering scrutiny in the USA in the recent period. Experimental work is also being carried out on a
comparatively large scale. However, one source reports [10] that a decision to forego building such structures
might be justified by the distance from populated centers, by the reliability characteristic of a given reactor type,
or by the "mild" operating conditions., The safety of reactor performance would be achieved preferably by im-
proved reliability of the associated equipment, rather than by housing the reactor inside containment shells. It
is possible that a reactor type with a large margin of safety may be acknowledged as suitable for use on a wide
scale, and that favorable experience in the operation of that type will be accumulated; it would then be possible
to make a definitive decision to do without such containment structures. The use of the containment vessels acts
as a serious hindrance to achieving low nuclear power costs, and all measures should be taken to cut the costs of
the containment enclosures or to circumvent them altogether.
As stated in another report [11], the probability (for each set of 100 reactors) of fatalities resulting from a
reactor accident is 2-10-8 in the USA. At the same time, the probability of fatalities in automobile accidents in
the USA is 2-10-4.
Several American engineers feel that to avoid excessive costs incurred in building containment enclosures,
the strength of the latter should be designed to match only 40% of the maximum possible energy of the conjectured
reactor explosion.
One view [12] is that the question of the need to build such containment vessels is still unresolved. And only
in the case of a substantial potential hazard of a reactor accident plus absence of any possibility of accurately
determining the extent of the possible damage would the additional costs associated with erecting the enclosure be
justified.
In the 21st semiannual report of the US Atomic Energy Commission (AEC) [13], the need to build contain-
ment shells was justified by the argument that one day following shutdown of a reactor rated at 60 Mw power,
the amount of radioactive material present within the reactor is equivalent to 300 tons of radium. An explosion
of the reactor accompanied by the ejection of radioactive materials into the surrounding locality could be more
dangerous than a nuclear explosion insofar as radioactive contamination of the locality is concerned, although
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less dangerous with respect to its mechanical effect than any large-scale explosion which might occur in any other
branch of industry.
If the installation cannot be isolated, then gas-tight enclosures (vessels) must be constructed around the re-
actor. The scale of this secondary shielding for each particular reactor will be determined by the location of the
reactor. Laying great stress on the correct design of protective containment shells, the US AEC conducted tests on
designs of containment structures at the ballistics research laboratory (Aberdeen, Maryland). Experiments on
mockups scaled to one-fourth natural size were staged, using appropriately shaped charges of explosi es.
The experiments showed that when models of a radioactive reactor vessel are destroyed with t aid of a
corresponding amount of explosives, a shock wave is set up which imposes on the outer containment shell a
pressure increasing with very high speed.
Despite the acknowledgment that reactor performance records show excellent safety and that containment
shells entail high costs, and even in the absence of any convincing proofs justifying the erection of such contain-
ment shells, large-scale nuclear installations are enclosed by primarily metal protective enclosures in a number
of countries. Furthermore, recent reactor design practice shows that several large reactors are being built without
containment shells (e. g., the British nuclear electric power stations with graphite-moderated gas-cooled reactors,
the Soviet nuclear power stations, the Kurchatov generating station at Beloyarsk, the Voronezh station, etc.). The
following conclusions may be arrived at on the basis of the above:
1. At the present state of the art of reactor design, building of containment shells is expensive and in many
cases superfluous overdesign;
2. Concrete biological shielding can handle the job of preventing release of radioactive elements in case
of a reactor accident, without incurring additional cost penalties. In that case, taking account of the design in-
ternal overpressure, it is usually necessary to simply strengthen the reinforcement of the concrete in the lateral
and top biological shielding and to make sure that the reactor chamber is leak-proof. This measure is taken in
an experimental power reactor now under construction in the USSR, and will undoubtedly provide for sufficiently
reliable localization of radioactivity release within the reactor chamber in case of an explosion [14].
3. Erection of special containment structures around the reactor may be justified only in the case where,
in the first instance, the reactor 'is an experimental type and is prone to an explosive accident both with respect
to basic design and with respect to the nature of .the moderator or coolant employed; in the second instance, if
the design of the biological shielding does not enablethe latter to localize radioactive elements which might
become ejected in the event of an accident to the area within the reactor hall; and in' the third instance, if the
reactor is situated so close to populated localities or buildings occupied by personnel that these could become
targets of radioactive contamination in the event of an explosion.
4. In certain cases, the presence of the protective containment shell may even aggravate the danger of ex-
plosion. For example, the release of hydrogen contained in the pressurized coolant water by the explosion may
result in the formation of an explosive mixture with the air trapped within the containment shell [10].
Taking into account the theoretical possibility of destruction or damage to the steel containment structure
by missile fragments of equipment, parts, or tubes torn loose in a reactor explosion, recent practice has added the
feature of a reinforced-concrete inner lining of 0.6 m thickness to provide an added safeguard inside the steel
containment shells, in some designs. *However, even this approach cannot fully guarantee that the steel shell and
the reinforced-concrete lining will not be penetrated by fragments of equipment, since the calculated weights of
such "missiles" and the predicted velocity imparted to them by the explosion are entirely arbitrary.
In the design of the USA package power reactor APPR-1, a power plant with a pressurized-water-cooled
water-moderated reactor rated at 10 Mw thermal power, it was found through theoretical calculations that
potential "missiles" could attain speeds up to 15 m/sec while traversing a path of 12 m inside the containment
volume. Reinforced-concrete lining 0.60 m thick would under such conditions be pierced by such "missiles" as
pieces of piping, valves 50 mm in diameter, pieces of control rods, etc. [10].
One positive quality that May be attributed to an interior reinforced-concrete lining is the fact that it pro-
vides some additional biological shielding, and another is that it limits the temperature rise of the steel shell in
case of an accident, and consequently diminishes the attendant thermal stresses. However, the usual requirements
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imposed on the biological shielding around the reactor and the steam generators for the safety of the operating
personnel .present inside .the containmentrshell.volume during the operation of the reactor makes any additional
shielding such as provided by the reinforced-concrete lining of the containment shell superfluou.s. The thermal
stresses in the containment shell are not determining factors in deciding shell thickness, in most cases. More-
. ,
over, practice in building and placing this type of reinforced-concrete structure has revealed great difficulties
in erecting and reinforcing them. The above-mentioned iourcg [10] confirms the fact, based on experience in.de-
signing containment shells, that linings of reinforced concrete are costly in most cases, and that internal concrete
barrier shields not integral with the steel containment structure are more practical. These concrete shields are
constituted in some cases by the walls of the reactor biological shielding, by the heat exchangers, or by other
radioactive equipment forming, together with the reactor top shield, pressure-tight chambers capable of contain-
ing, within certain limits, any radioactive release (in the event of a reactor explosion) and also capable of local-
izing the "missile" fragments which might be sent flying by such an explosion.
The striving to nullify the probability of "missiles" being
torn loose by a reactor explosion has decided the choice of material
for the pressure vessel and piping of the primary loop in the case of
the reactor for the US "Yankee" nuclear power station, where car-
bon steel lined with stainless steel is employed in the pressure
2 I vessel, and tough austenitic stainless steels are used for the ducting.
Nevertheless, this reactor setup is buttressed by a "missile" shield
incorporating concrete liners surrounding the reactor [9]. The APPR-1
package power reactor, with a steel containment shell backing up
an external concrete biological shield 0.915 m thick, has in addition
an interior reinforced-concrete "anti-missile" lining 0.60 m thick,
which in essence forms part of the reactor biological shield. This
EEC lining is encased by a sheet steel envelope designed to facilitate
deactivation of the reactor house (Fig. 8). Erection of the steel
containment shell is no easy job, and will not be undertaken in
future projects, as stated earlier. In conclusion, it is well to note
that the variant of "missile" fragments flying loose in all directions
in the event of a reactor explosion is highly improbable and the
approach must be geared to eliminating the possibility of such missiles
flying loose by judicious engineering of the basic equipment, pipe-
work, fastening and hold-down devices. In giving thought all the same
to the problem of protecting the steel shell against impact of these
missiles, reliance should be placed on the concrete structure of special
inner linings of reinforced concrete. Linings of this type may prove
feasible only in those rare cases when it is decided not to place a biological shield directly around the reactor and
ancillary equipment, and to make use of only an outer shield which in this case will incorporate the anti-missile
lining.
Fig. 8. Vapor containment shell of the
APPR-1 reactor; 1) containment shell;
2) concrete lining; 3) steel shell; 4) sec-
ondary shield; 5) primary shield; 6)pres-
surized-water reactor.
LITERATURE CITED
1. R. Bergstrom and W. Chittenden, Nucleonics 17, 4, 86 (1959).
2. Nucleonics 16, No. 2 (1958).
3. Physical and Engineering Institutions in Norway, Atomnaya Energ. 5, 4, 468 (1958).?
4. R. Liljeblad and K. Madsen, Proceedings of the Second International Conference on the Peaceful Uses of
Atomic Energy (Geneva, 1958) Paper 2419.
5. A. N. Komarovskii, Structural Materials for Radiation Shielding of Nuclear Reactors and Accelerators [in
Russian] (Atomizdat, Moscow, 1958).
6. A. Brake, Nuclear Energy Eng. 13, 130 (1959).
7. H. Davis, J. Am. Concrete Inst. 29, No. 11 (1958).
8. D. Campbell-Allen, J. Am. Concrete Inst. 30, 6, XII, (1958).
9. C. Horton, Nuclear Eng. 3, 33, 515 (1958).
*Original Russian pagination. See C.B. translation.
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10. C. Chave and 0. Balestricci, Proceedings of the Second International Conference on the Peaceful Uses of
Atomic Energy (Geneva, 1958) Paper 1879.
11. D. Shortall, "New US achievements in reactor 'design," Paper read Nov. 22, 1957, at the Autumn Nuclear
Week Conference in New York.
12. B. John Garrick, Civil Eng. 28, No. 9 (1958).
13. Twenty-first Semiannual Report of the AEC, USA (July-December 1958).
14. A. N. Komarovskii, Atomnaya Energ. 7, 3, 205 (1959).*
? Original Russian pagination. See C.B. translation.
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MECHANICAL PROPERTIES AND MICROSTRUCTURE
OF CERTAIN CONSTRUCTION MATERIALS AFTER
NEUTRON IRRADIATION
I. M. Voronin, V. D. Dmitriev, Sh. Sh. Ibragimov,
and V. S. Lyashenko
Translated from Atomnaya gnergiya, No. 6, pp. 514-518, June, 1960
Original article submitted August 27, 1959
The article presents data on the influence of irradiation during normal operation of the reactor at the First Atomic
Power Station on the mechanical properties and microstructure of a number of austenitic, ferrite, ferritemartensitic
steels and molybdenum. The authors demonstrate a sharp falloff in the plasticity of the austenitic and ferrite steels
and molybdenum as a result of irradiation with a total flux of (0.9-3.4).1020 neuticm2 at a temperature of 450-650?
C. It is found that the causes of steel and molybdenum becoming brittle differ from one another in their nature.
The information presented may prove useful in the planning of new atomic power plants.
The intent of the present paper was to find out the degree of influence of the operating conditions in the re-
actor at the First Atomic Power Station on the mechanical properties and microstructure of technically pure moly-
bdenum and structural steels of the following brands: 1Kh18N9T, Kh2ON14S2, 1Kh15N11M2C2T, 2Kh13, 1Kh13BMS2,
and Kh10Yu5ST. Test steels, whose chemical compositions are listed in Table 1, were also investigated.
TABLE 1
Chemical composition of test steels
Steel brand
Content of elements, %
Cr
Nb
Mo
Si
Cu
1Kh13BM* . .
0,12
12,4
1,3
0,6
--
1Khl3B2M
0,10 --0,15
12--14
1,5-2,0
0,8-1,2
--
1K1113B2MSFB..
0,10 --0,15
12-14
1,5-2,0
0,5-1,0
0,5-1,0
0,2--0,5
0,2-O,5
2Kh17B2M .
0,15-0,20
16-18
1,5-2,0
0,8-1,2
--
2Kh17B2MS2 . .
0,15 --0,20
16-18
1,5-2,0
0,8-1,2
1,5-2,0
2Khl7B2MS3
0 15-L0,20
16-18
1,5-2,0
0,6-0,8
2,5-3,0
Khl7MS3D2?
0,03
18,0
--
1,0
3,0
1;75
2Kh17B21vIS3P2,
0,15-0,20
16-18
1,5-2,0
0,6 - -0,8
2,7-3,2
1,5--2,0
*Content by analysis.
To irradiate the samples we used the SUZ channel of the outer ring of the reactor at the First Atomic
Power Station [1, 2]. In-the empty graphite sockets of the channel, which were supposed to be in the active zone,
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four cylindrical openings with a diameter of 18 mm were drilled out along the axis of the socket, and into these
were placed the samples, contained in unsoldered quartz ampoules. The ampoules were in contact with the nitro-
gen atmosphere that filled the reactor pile.
The temperature was controlled by thermocouples located at three points of the active zone of the channel,
one high, one middle position, and one low. The temperatures in between were determined according to the dis-
tribution of heat liberated inside the graphite pile over the height of the reactor. Depending on the location of the
samples, as well as the operating power of the reactor, the temperature thus determined was found during the ir-
radiation to lie within the interval 450-650? C. Heat liberation in the samples was not accounted for.
In preparation a part of the samples was stamped out of sheet material, the rest being cut from rods. The
corresponding heat treatment was performed in vacuum (,-1.10-3 mm Hg). Cold plastic deformation was realized
(after quenching of the steels) by drawing.
The number of test samples for each point was equal to from three to five. The testing was done at room
temperature.
The total neutron fluxes shown further in Tables 2-5 are given for thermal neutrons and are approximate.
It is assumed that the number of neutrons of higher energy amounts to ?25% of the number of thermal neutrons.
The mechanical properties of austenitic steels before and after irradiation are shown in Table 2. From
the data obtained it follows that the initial state of the material plays a significant role in the change in proper-
ties during irradiation. For example, the microhardness of quenched steel 1Kh18N9T after irradiation increased
by 65 kg/mm2, the relative elongation of the quenched steels Kh2ON14S2 and 1Kh15N11M2S2T decreased by a
factor of three or more, and for steel 1Kh18N9T by approximately 20%.
TABLE 2
Mechanical properties of austenitic steels
Initial state
Steel brand
Total flux
20
X 10
neutrons/cm2
Irradiation
temp, ?C
Yield strength,
kg/mm2
Relative
elongation,
Microhardness,
kg/mm2
Before
irrad.
After
irrad.
Before
irrad.
After
irrad.
Before
irrad.
After
irrad.
Quenching from
1Kh18N9T
3.4
?650
57
58
39
31
138
203
1100? C inwater
Kh2ON14S2
{ 2.4
3.4
?500
?650
56
56
53
54
45
45
17
11
156
156
199
187
1Kh15N11M2S2T
2.7
?550
68
63
46
15
185
225
Quenching from
1Kh18N9T
3.4
?650
88
74
16
23
317
265
1100? C and cold
Kh2OH14S2
1.2
?500
96
91
16
17
deformation*
1Kh15N11M2S2T
3.1
?600
139
110
4.4
8.8
432
412
Quenching from
1Kh18N9T
3.4
?650
65
63
26
25
209
195
1100? C, cold de-
Kh2ON14S2
0.9
?450
69
67
33
20
formation ? and
1Kh15N11M2S2T
2.7
?550
86
82
20
11
273
283
3-hour anneal at
850?C
The same with
1Kh18N9T
I 2.4
?500
64
69
29
22
180
236
10-hour anneal
1 3.4
?650
64
64
29
31
180
178
at 650? C
Kh20N14S2
1.0
?500
71
65
35
18
254
277
1Kh15N11M2S2T
2.7
?550
87
86
15
13
286
300
*The shrinkage for
steel 1Kh15N11M2S2T was ?90%, for the rest ? 0%.
Figure 1 shows the microstructure of austenitic steel Kh2ON14S2 in the quenched state before and after irradia-
tion. The same changes in structure are noted for steels 1Kh18N9T and 1Kh15N11M2S2T. It is noteworthy that
the quenched austenite grain assumed a more uniformly circular shape after irradiation at 500-650? C, while at
the grain boundaries carbides were precipitated. The same change of structure is also observed after the action of
a single temperature and, consequently, is not the result of the specific action of the neutron field. It is known
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that being held in the region of very ugh temperaturei gives rise waging (dispersion-hardening) of austenitic.
steels. With the 'simultaneous acticin- of a rieUtion field andlemperature, the aging .process'is probably accelerated
[3].
a
Fig. 1. Microstructure of steel Kh2ON14S2 in the quenched state (X. 325):
a) Before irradiation; b) after irradiation at a temperature of 650? C by a
total flux of 3.4.1020 neutrons/cm2.
As a result of the work of various investigators [4] it is presently known that temperatures of the order of
those considered in the present article are sufficiently high for the indicated steels from the viewpoint Of being
able to retain in these a large 'fraction of primary radiation dislocations. Nevertheless, since dislocations during
the time of neutron bombardment must arise at any temperatures (although with an increase in the temperature
at which bombardment is taking place the defects that appear will be annealed out in an ever-increasing number
and at an ever-increasing rate) these dislocations should effect those processes which occur at these temperatures
in the nonirradiated steels. With increasing temperature the relative participation of the neutron flux inte.nsitY
should increase:
For cold-deformed steels a reduction in yield strength and hardness, as well as an increase in relative,
elongation are observed as the result of irradiation.
In the present instance there occur simultaneously in the steels two processes, which have opposite effects
bri their properties: relief of cold hardening by anneal and aging accompanied by the separation of hardening
phases. There is no doubt that the observed dehardening of cold-deformed steels was the result of the former
, process.
It is known [5]that preliminary cold deformation augments the subsequent agidg of.steels. Figure 2.illuStrates
the austenitic decomposition of cold-deformed .(~300/0 deformation) steel Kh2ON14S2 as the result of irradiation
at a temperature of ?506? C. From a comparison of Fig. 2 and Fig. lb we can see the difference in the degree
of austenitic decomposition and in the distribution of separations between the quenched and cold-deformed steels.
Stabilization of the structure of cold-deformed austenitic steels by anneal at 850?C for a period of 3 hours
or two-stage anneal at 850? C for 31ours and at 650? C for 10 hours leads to a reduction in. the degree of decline
in plasticity as the result of irradiation. The yield strength of the stabilized steels as the result of irradiation was
not altered significantly.
Consequently, it may be concluded that the observed change in the properties of austenitic steels as the result
6f irradiation by a total flux of 0.9-3.4.1020 neutrons/cm2 at temperatures of 500-650? C are connected primarily
' with the nonequilibrium state of the material and with the processes taking place in it.
A neutron field does not exert any essential influence on the mechanical properties of ferrite and ferrite-
martensitic steels (Table 3). This is clearly evident from a comparison of the properties of the steels tested after
irradiation at 600? C and after being held at 600 and 700? C without irradiation* (Table 4). ?
* The information relating to the properties of the test steels after prolonged exposure to temperatures of 600 and 700? C
without irradiation was obtained by M. D. Abramovich.
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Fig. 2. Microstructure of quenched cold-deformed
steel Kh2ON14S2 after irradiation at a temperature
of ?500? C by a total flux of 1.2-1020 neutrons/cm2.
TABLE 3
Mechanical properties of ferrite-martensitic steels
Unlike the austenitic and ferrite steels, in the region
of the temperatures investigated there was no appreciable
aging accompanied by separation. But in ferrite 1752
-
chrome steels containing a large quantity of silicon there
occurred as the result of irradiation a very strong drop,
unobserved in ferrite steel containing no silicon, in the
magnitude of the relative elongation. This change in
plasticity was evidently caused by those factors (e. g.,
the x-ray detected separation of the a -phase) which give
rise to brittleness of the given steels after long exposure
to temperatures of about 600? C without irradiation.
In 1310- chro m e steels belonging to the ferrite-mar-
tensitic class, heat-treated prior to irradiation to the
stable ferrite-perlite structure, no significant changes in
the properties were observed after irradiation.
-
Steel brand
Initial state
Total flux
10-20
neutrons/cm2
Irradiation
temp, ?C
Yield strength,
kg/mm2
Relative
elongation,%
Microhardness,
kg/mm2
Before
irrad.
After
irrad.
Before
irrad.
After
irrad.
Before
irrad.
After
irrad.
2Kh13
1Kh13BMS2
Kh1OYu5ST
(ferrite)
1-hour anneal
at 750? C.. .
1-hour anneal
at 750? C. . .
As delivered.
0.9
0.9
2.7
?450
?450
?550
59
65
68
60
67
71
28
25
15
24
23
13
234
225
?
230
254
?
TABLE 4
Mechanical properties of test steels
Steel brand
Yield strength, kg/ mm2
Relative elongation, 010
Initial
state*
After holding
at 600?C for
1560 hr
After holding
at 700?C for
1000 hr
After irra-
diation at
-,600?C
(total flux
3.1 ? 102?
neutrons/
/cm2)
Initial
states
After holding
at 600?C for
1560 hr
After
hold-
ing at
700?C
for
1000 hr
After ir-
radiation
at ?600?C
(total flux
3.1 1020
neutrons/
!cm2)
1Khl3BM
60
58
54
51
23
23
20
20
1Khl3B2M
59
56
53
48
21
22
24
23
1Kh13B2MSFV
63
61
52
55
18
20
22
19
2Khl7B2M
60
56
55
52
20
21
23
19
2Khl7B2MS2
73
72
68
72
21
12
23
8.0
2Khl7B2MS3
83
75*?
69
72
19
2.0**
3.0
2.5
Khl7MS3D2
76
73
66
86
18
3.0
9.0
4.3
2Kh17B2MS3D2
81
74
67
84
18
1.5
5.0
1.2
*The 13 steels were annealed at 750?C for 2 hr, the rest were normalized from 900?C.
**Held for 1000 hr.
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TABLE 5
Mechanical properties of molybdenum before and after irradiation
Molybde-
num batch
Initial state
Total flux
20
X 10 -
neutrons/
/cm2
Irradia-
tion tern-
perature,
?C
Yield strength,
kg/ mm2
Relative elong-
gation, aio
Microhardness,
kg/ mm2
Before
irrad.
After
irrad.
Before
irrad.
After
irrad.
Before
irrad.
After
irrad.
I
II
30-min anneal at
1250?C
2-hour anneal at
1000?C ,
1.2
2.0
-650
-.450
65
72
84
110
21
7
2.6
1.7
187
366
Table 5 shows the change in properties of technically pure molybdenum after irradiation. The increase in
strength and hardness of molybdenum as the result of irradiation and the sharp reduction in relative elongation
strongly depend on the initial state of the molybdenum, as noted earlier [6].
It appears that in connection with the high melting temperature and high elastic properties of molybdenum,
no amount of exposure to temperatures of 450-550?C without irradiation will give rise to any processes that are
able to strongly modify the properties of molybdenum. The conclusion may therefore be made that radiation
dislocations arising during irradiation and influencing the mechanical properties are retained even at tempera-
tures of 450-550?C. In steels subjected to the same conditions a large part of the radiation dislocations apparent-
ly succeed in being annealed away.
The authors wish in conclusion to express their gratitude to E. V. Chermashentsev and A. Ya. Ladygin for
their part in the work undertaken.
LITERATURE CITED
1. D. I. Blokhintsev, N. A. Dollezhal', and A. K. Krasin, Atomnaya nerg. 1, 10 (1956).*
2. D. I. Blokhintsev, M. E. Minashin, and Yu. A. Sergeev, Atomnaya gnerg. 1, 24 (1956).*
3. G. Murray and W. Taylor, Acta. Met. 2, 1, 52 (1954).
4. N. F. Pravdyuk, et al., Proceedings of the Second International Conference on the Peaceful Uses of Atomic
Energy (Geneva, 1958). Papers by Soviet Scientists. Nuclear Fuel and Reactor Metals [in Russian]
(Atomizdat, Moscow, 1959) Vol. 3, p. 610.
5. E. Goudremond, Special Steels [Russian translation] (Metallurgizdat, Moscow, 1959) p. 239.
6. C. Bruch, W. McHugh, and R. Hockenbury, J. Metals 7, 2, 281 (1955).
*Original Russian pagination. See C. B. translation.
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Declassified and Approved For Release 2013/02/19: CIA-RDP10-02196R000100050006-3
EXTRACTION OF URANIUM FROM SOLUTIONS AND PULPS
13. N. Laskorin, A. P. Zefirov, and D. I. Skorovarov
Translated from Atomnaya Energiya,Vol. 8, No. 6, pp. 519-529,
June, 1960
Original article submitted July 18, 1959
Extraction processes are finding increasing application for the processing of uranium from raw materials. The
.high selectivity and the fluid aggregate state of the extraction agents give this method advantages over the pre-
viously known processes.
This article gives data on the extraction of uranium from sulfate, nitrate, chloride and phosphate solutions
and pulps, most frequently encountered in the hydrometallurgy of uranium. The extraction agents suitable for in-
dustrial use are the esters of carboxylic acids, the esters of phosphoric and phosphinic acids and also liquid cation-
exchange materials and anion-exchange materials, in a number of cases (in the excaction of uranium from colored
solutions) have advantages over solid ion exchange materials. A system is described for the extraction of uranium
from dense ore pastes.
Introduction
Extraction methods were used for uranium at the start of the development of the uranium industry. In hy-
drometallurgy, extraction was used at the stage of obtaining pure uranium compounds. At a later stage, together
with the development of sorption processes for extracting uranium from solutions and ore pulps, work was carried
out on extraction from nitrate, sulfate, chloride desorption and other solutions and ore pulps of varying thickness.
By drawing an analogy between the mechanism of absorption of sorbents and extraction agents it is possible
to find and use new classes of organic solvents which are also liquid ion-exchange materials.
In a number of cases the difference in the processes of sorption and extraction consists mainly in the differ-
ence between the aggregate state of the organic phase (absorber). For example, the mechanism of extraction by
alkyldithiophosphoric, alkylphosphoric and other acids is similar to the mechanism of sorption by cation exchange
materials:
ni.J0;2.F m42- 2n (HA)2 (UO2)n(HA)7A 2nH?,
and the mechanism of extraction by alkylamines is similar to the mechanism of anion exchange:
ri
UO2(934)71.1-(2/1-2) (n ? 1)[(113IN1-I)2SO4] -7=1" RB3NE1)2?_2U02(SO4)l + (ii?
where n is equal to 2 or 3.
The difference in the aggregate state of extraction agents and sorbents provides new possibilities in the use
of liquid absorbents. Compared with solid absorbents, liquid absorbents have a much higher rate of absorption and
capacity, and hence there is a greater degree of concentration of uranium during extraction from dilute solutions.
In particular, liquid absorbents can be used for extraction from anhydrous ore materials whereas sorption processes
can only be used with ore pulps containing not more than 40% solid materials. In a number of cases solid extraction
agents are of interest, for example, porous active sorbents impregnated with extraction agents (gels and activated
carbons).
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TABLE 1
The Solubility of Esters of Acetic Acid in Water and Selectivity with Regard to Uranium
_
Extraction Agent
Chemical Formulapurities
Molecular
weight
Solubility in
water at 20?C,10
.
Total of im-
in
U308, To
Isopropyl acetate
CH3COOCH(CH3)2
102
3
2.6
Butyl acetate
CH3C00C41-18
116
0.6
0.15
Isoamyl acetate
CH3C00(CH2)2CH(CH3)2
172
0.25
- 0.06
TABLE 2
Selectivity of Certain Extraction Agents for Uranium
Extraction Agent
Chemical Formula
Total of impurities
in U308, To
Isooctyl alcohol
CH3(CH2)3CH(CH2NCH2OH
1.7
Dibutyl ether
(C4H9)20
0.2
Isoamyl acetate
CH3C00(CH2)2CH(CH3)2
0.06
Isooctyl acetate
CH3COOCH2CH(C2118)(CH2)3CH3
0.05
Tributyl phosphate
(C41180)3P0
0.5
Extraction of Uranium from Solutions
Studies have been made of the extraction of uranium by different classes of organic compounds (alcohols,
ethers, ketones, diketones and their halogen derivatives, esters, carboxyl, phosphoric, phosphinic, alkyldithiophos-
phoric acids, aliphatic amines, etc). This section will only deal with the properties of a few of the solvents.
The extraction of uranium by alcohol, ethers and esters of carboxylic acids. The extraction agents of this
group extract uranium from nitrate solutions and can only be used in the presence of salting-out materials since
under ordinary conditions they have a low distribution coefficient (Kp = 0.2-0.7). With increase in the length
of the aliphatic radical, Kp of the solvents decreases in a definite order for each homologous series.
It is characteristic that with increase in the length of the hydrocarbon chain by a CH2 group in the acid
radical of the carboxylic esters, Kp is reduced more noticeably than with a corresponding increase in the length
of the hydrocarbon chain in the alcohol group. Increase in the temperature of the solutions leads to a decrease
in K
P'
The value of carboxylic esters and ethers consists of their high selectivity with respect to uranium. It has
been found that with increase in the molecular weight of the solvent the selectivity of uranium extraction in-
creases.
It can be seen from Table 1 that the selective capacity of acetic esters increases with their solubility in
water. The selectivity of extraction agents is reduced in the order: esters of carboxylic acids, ethers, alcohols.
Their data are given in Table 2; for comparison purposes information is given on tributyl phosphate (TBP) which
has been extensively used in the extraction of uranium.
Kp for the extraction of uranium by alcohol, ethers and esters of carboxylic acids depends strongly on the
concentration of salting-out materials in the aqueous solution. This relationship is shown in Fig. 1 for isoamyl
acetate and dimethyl phthalate. The extraction of uranium by these types of solvents is strongly inhibited by
sulfate, phosphate, arsenate and other ions.
In a number of cases, including extraction of uranium from thick nitrate pulps, effective use can be made
t:oth of carboxylic esters and ethers.
The literature [1] indicates the possibility of effective use of such compounds as tetrahydrosylvan and tetra-
hydropyran for the extraction of uranium from nitrate solutions with small amounts of salting-out materials.
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5
4
3
2
1
0
2
100 200 .300 400 500
Ca00312.g/ liter
Fig. 1. The salting-out action of Ca(NO3)2 in the
extraction of uranium by isoamyl acetate (curve
1) and dimethyl phthalate (curve 2).
Ir',
3
2
0 0
0 100 200 300 400 500 600 700
Concentration of H2SO4, g/liter
Fig. 2. Extraction of uranium by 100%
TBP from sulfuric acid solution, uranium
concentration 10 g/ liter; ratio of volume
of organic phase to water, 10.4.
Esters of phosphoric acid. Trialkyl phosphates are a well-known class of selective solvents for uranium. Kp
of the trialkyl phosphates increases with increase in the hydrocarbon radical up to C5 C6, then gradually de-
creases. The nature of the hydrocarbon radicals is then of decisive importance. For example, the triaryl phos-
phates are almost unable to extract uranium.
Increasing the temperature for ethers and esters of carboxylic acids leads to a reduction in Kp. The selec-
tivity of trialkyl phosphates increases with the molecular weight of the extraction agent (Table 3). The intro-
duction of salting-out materials (nitric acid and nitrate salts) causes an increase in Kp of uranium. The extrac-
tion is strongly inhibited by sulfates, arsenates, phosphates and fluorides. It is of interest that TBP also extracts
uranyl sulfate but only from solutions with a high concentration of sulfuric acid. As can be seen from Fig. 2,
with a sulfuric acid concentration of 700 g/liter,Kp is equal to three for extraction by 100% TBP.
TABLE 3
Extraction Agent
Chemical Formula
Molecular weight
K (for 1 M solu-
tion of extraction
agent in kerosene)
Total of impuri-
ties of U308, %
Tripropyl phosphate
Tributyl phosphate
Triamyl phosphate
Trihexyl phosphate
Trioctyl phosphate
Triphenyl phosphate
Tricresyl phosphate
(CiH70)3P0
(C4H30)3P0
224 7.4
266 8.3 0.5
(C5I1110)3P0
310
9.3
0.3
(C61-1130)3P0
352
9.6
(C8H170)3P0
434
9.3
0.13
(C6H50)3P0
326
0.025
(CH3C6H40)3P0
368
0.0015
Note; K was determined at equilibrium conditions in equal volumes of the phases. The initial solution con-
tained 30 g/liter of uranium and 25 g/liter of nitric acid.
TBP was used in the extraction and purification of chemical concentrates and also in connection with the
extensive development of sorption technology in the purification of nitrate desorption (regeneration) solutions of
uranium (Fig. 3).
Trialkyl phosphates can be used as plasticizers in films of cellulose esters, polymonochlorostyrene and other
materials. It was found that these solid extraction agents can extract uranium from nitrate solutions. For
example, 1 g of a cellulose film containing 16% TBP absorbs 35 mg of uranium from solution providing the
initial solution contains 5 g/liter uranium, 200 g/liter nitric acid and 300 g/liter ammonium nitrate.
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Ore Nitrate mother liquor Saturated Reextraction agent
112504
Leaching and
classification
Sand to
waste
Desorbent HNO3
Ion
exchange
Slurry to
waste
extraction agent
Extraction
with 30/0 TBP
Desorption
solution
Reextraction
Extraction agent
Fig. 3. Flow sheet for the extraction purification of nitrate desorbed solutions of uranium.
Precipitation
of uranium
Finished
product
Diisoamyl ester of methylphosphinic acid (DAMPA). DAMPA [i(C5H110)2POCH3] is one of the lower members
of the alkylphosphonate series. The representatives of this series with large molecular weights, for example the
dihexyl ester of hexylphosphinic acid or the dioctyl ester of methylphosphinic acid are also very interesting since
they are less soluble in water and very effective in the extraction of uranium.
DAMPA is a transparent colorless liquid of specific gravity 0.953, refractive index 1.4284. The solubility
of DAMPA in water is 0.3 g/liter, boiling point 256?C. DAMPA extracts uranyl nitrate and nitric acid, which
is described in [2]. DAMPA is readily soluble in various organic solvents and can be used to extract uranium
from nitrate and chloride solutions. In the extraction of uranium from a 1 M solution of nitric acid.K for 100%
DAMPA reaches 100 with an initial concentration of uranium in the solution of 10 g/liter. Increase in the
nitric acid concentration leads to a considerable increase in Kp. It has been shown that DAMPA has fairly good
selectivity. In the extraction of uranium from nitrate solutions,iron is extracted to a very small extent and
aluminum and calcium are extracted in small quantities.
0
2.J 5 5 7 6 9 10 11 12
Fig. 4. Extraction of uranium by a 1010
solution of DAMPA in kerosene from
chloride solutions (initial solution 3.5
g/liter uranium, ratio of volumes of
organic phase to aqueous phase equal to 1).
In the extraction from chloride solutions K increases con-
siderably with increase in the hydrochloric acid concentration (Fig. 4).
Sulfate and phosphate ions inhibit the extraction of uranium
with DAMPA. Figure 5 shows a curve characterizing the inhibiting
action of phosphoric acid.
The presence of small amounts of nitric acid (15-20 g/liter)
considerably increases the extraction of uranium from phosphate
solutions. For example, in the extraction of uranium from a solution
containing 300 g/liter of phosphoric acid, KD = 0.23. When uranium
is extracted by a 10% solution of DAMPA in kerosene, equilibrium is
established after 2-3 minutes. The reextraction of uranium from
DAMPA can be carried out by 5-1053 solutions of sulfuric acid and
ammonium carbonate and also by other reagents.
Trioctyl phosphinoxide (TOPO). TOPO [(C81-117)3P0] forms
stable complex compounds with uranium. The mechanism of extrac-
tion has not been fully studied although there are reasons to assume
that compounds such as TOPO form complexes with uranium similar
to TBP. TOPO is readily soluble in carbon tetrachloride, aromatic
hydrocarbons, alcohols, esters and somewhat less soluble in petroleum
ether and kerosene. A 0.1 M solution of TOPO in kerosene can be
prepared. With increase in temperature the solubility of TOPO in
all solvents increases noticeably. The solubility of TOPO and conn-
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Ap
4
pounds of uranium with TOPO are similar. The impu-
rities which are normally present in industrial solutions,
except trivalent iron and chromium, are extracted very
little by TOPO. TOPO extracts uranium from nitrate
and chloride solutions. Phosphoric acid reduces the ex-
traction of uranium by TOPO much more strongly than
other impurities.
The introduction of complex-forming reagents
such as citrate and oxalate ions considerably increases
the extraction of uranium from solution. In the extrac-
o 100 200 3 0 0 tion of uranium by a 0.1 M solution of TOPO in kero-
//3PO4,g/liter sene from nitrate solutions, Kp= 900. Increasing the
concentration of nitric acid in nitrate solutions causes a
Fig. 5. The effect of phosphoric acid on the extrac-
very small increase in Kp. When the concentration of
tion of uranium by a 10% solution of DAMPA in
nitric acid is increased to 200 g/liter there is a certain
kerosene from solutions containing 20 g/liter of
reduction in Kp.
nitric acid.
In the extraction of uranium by TOPO, equilib-
rium is attained after 0.5-1 min. It has been shown
that at low concentrations of uranium K maintains a large value. This property means that TOPO can be used
for the extraction of uranium from very dilute solutions.
Carbonate solutions can be used in the reextraction of uranium from TOPO. For example, a 10% solution
of ammonium carbonate heated to 40?C in one operation lasting 1-2 minutes reextracts up to 99.5% uranium
with a ratio of volumes of organic phase to aqueous phase equal to 4. Reextraction of uranium by 10 N hydro-
chloric acid is less effective. TOPO has good chemical stability. Its extraction properties hardly change even
after repeated use. The prolonged action of concentrated acids does not produce any visible changes or any
changes in the extraction properties of TOPO.
The addition of alkyl phosphates, alkyl phosphonates, alkyl phosphinates and particularly alkyl phosphinox-
ides to solutions of alkyl phosphoric acids considerably increase the value of K in the extraction of uranium.
Dialkyl phosphites. Studies have been made of various dialkyl phosphites [(R0)2P(OH)] with hydrocarbon
radicals from C4 to C8. The methods for preparing dialkylphosphites were based on the reaction of phosphorus
trichloride with the appropriate alcohol.
The studied dialkyl phosphites were soluble in hydrocarbons, their chlorine derivatives, ethers, alcohols and
ketones. Compounds of uranium with the lower dialkyl phosphites (with C4-05) were insoluble in aliphatic
hydrocarbons. The solvate or uranium with dioctyl phosphite is readily soluble in kerosene.
Diisoalkyl phosphites containing the hydrocarbon radical C4-05 extract uranium from 0.3 N solutions of
nitric acid with K = 8-12 of uranium in the extraction of dialkyl phosphites increasing with increase in the con-
centration of nitric acid in the aqueous solution. Kp of dialkyl phosphites increases with increase in the mole-
cular weight. For example, in the extraction of uranium by 1 M solutions of dialkyl phosphite from nitrate
solutions containing 0.3 N l-IN03, Kp for dibutyl phosphite is equal to 2, for diamyl phosphite 2.8 and for
dioctyl phosphite ?800.
In contrast to dibutyl and diamyl phosphites, dioctyl phosphite extracts uranium from sulfate solutions with
high values of Kp (> 1000); with increase in the concentration of sulfuric acid from 50-100 g/liter there is
a considerable reduction in K.
Alkyl phosphoric acids. Among the widely used extraction agents there are dialkyl phosphoric (R0)2P(0)0H
and dialkyldithiophosphoric (R0)2P(S)SH acids, in particular di-(2-ethylhexyl) phosphoric and di-(2-ethylhexyl)
dithiophosphoric acids. Dialkylphosphoric compounds were described in [3, 4] and have been used in certain
plants in the US. Alkyldithiophosphoric acids are of interest since they are cheap and readily available extrac-
tion agents for uranium and can be used when treating chloride, sulfate and phosphate solutions. Compounds of
this type are liquid cation-exchange materials. Molecules of dialkyldithiophosphoric acids are apparently di-
merized like the molecules of dialkylphosphoric acids and react with the uranyl ions as an exchange reaction
described at the beginning of the article.
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120
110
100
90
80
70
6'0
50
30
20
10
44,
NCI
11230
PO
0 1 2 3 4
Acid concentration, N
Fig. 6. Relationship between Kp of uranium and
the concentration of mineral acids in the extrac-
tion by a 10% solution of dioctyldithiophosphoric
acid in kerosene (initial solution 1 g/ liter of
uranium, ratio of volumes of organic phases to
aqueous phases equal to 0.2).
Figure 6 gives curves for the dependence of Kp of
uranium in the extraction by a 1010 solution of dioctyl-
dithiophosphoric acids in kerosene on the concentration of
various mineral acids in the aqueous solution. For the ex-
traction of uranium from phosphate solutions with a high
concentration of phosphoric acids which are obtained in the
acid decomposition of uranium-containing phosphorites
use is made of the mono- and dialkylpyrophosphoric acids
(H2RP207, HR2P207), which are efficient in extracting ura-
nium. In particular, use is made of diisooctylpyrophosphoric
and diisodecylpyrophosphoric acids with varying structures
of the hydrocarbon radicals.
Amines. Long chain alkylamines and alkylarylamines
extract uranium from sulfate solutions and concentrated
solutions of hydrochloric and nitric acid in a similar manner
to anion-exchange resins [5]. All the rules established for
the sorption of uranium by anion-exchange materials from
sulfate, nitrate, chloride, phosphate and other solutions can
be used in the first approximation when considering pro-
cesses of uranium extraction. In the extraction, use was
made of primary, secondary and tertiary aliphatic and ali-
phatic-aromatic amines which were insoluble in water. The
solvent for the amine compounds was a mixture of kerosene
with alcohol. In view of the tendency of amines to form
emulsions with aqueous solutions the content of solid particles
in them should be at a minimum. To reduce the emulsifi-
cation of amines with water it is best to use amines with
branched aliphatic radicals. The emulsification can also
be reduced by increasing the acidity and temperature of
the solutions.
Long chain aliphatic amines have high chemical stability in acid and alkaline solutions. The selectivity
of alkylamines with regard to uranium increases on transition from the primary to tertiary amino compounds.
For example, the secondary long chain amines are somewhat more selective than the primary compounds but
in contrast to the latter they extract trivalent iron in large quantities. When using secondary aliphatic amines
it is therefore essential to reduce the trivalent iron. Kp of uranium depends to a considerable extent on the
length and structure of the hydrocarbon radical. The solvents of alkylamines can also be benzene, carbon tetra-
chloride, chloroform, etc.
Table 4 gives K of uranium for various amines. A comparative evaluation of the extraction properties of
alkylamines was carried out for decimolar solutions of amines in kerosene in the extraction of uranyl sulphate
from pure solutions at pH = 1 and from solutions containing 50 g/ liter of sodium sulfate for pH = 1 and a ura-
nium concentration in the initial solution of 1 g/ liter.
The following conclusions can be drawn from data on the dependence of Ko on the molecular weight and
structure of hydrocarbon radicals for a number of amine extraction agents (Table 4):
1) for the homologous series of n-alkylamines, Kp of uranium increases with the molecular weight of the
amines and has a maximum value for trioctylamine;
2) KD of asymmetric amines such as methylstearylamine and stearyldimethylamine, is comparatively small
(5-10) and depends very strongly on the nature of the diluent;
3) when using alkylamines the degree of branching of the aliphatic radicals is very important. The de-
ciding role is probably the spatial hindrances. For example, Kp of uranium for tri-n-octylamine is equal to 200,
and for tri-(2-ethylhexyl)amine it is less than 1, which can be explained by the screening of the nitrogen atom
by the ethyl radicals;
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4) in extraction from sulfate solutions special attention should be paid to benzylheptadecylamine, for which
Kp of uranium is equal to or greater then 1000;
5) increasing the concentration of alcohol in the solvent leads to a reduction in K.
In the extraction of uranium from sulfate solutions, nitrate, chloride and phosphate ions reduce the ex-
traction properties of the amine extraction agents. With regard to their depressing action they can be placed
in the following order: nitrate, chloride, phosphate and sulfate ions. They almost entirely inhibit the extrac-
tion properties of alkylamines with a concentration of nitrate and chloride ions of -.15-30 g/liter. Reduction
in the extraction properties of alkylamines by sulphate ions is observed to a noticeable extent only at very high
concentrations of sulphates (^, 200-300 g/liter) and at pH r=-'? 4.
Figure 7 shows the dependence of uranium extraction by a 0.1 M solution of tridecylamine in kerosene on
the concentration of mineral acids in the initial solutions. As can be seen from this diagram, trialkylamines
can be used to extract uranium from 6-10 M solutions of nitric and hydrochloric acids. Under these conditions,
using trialkylamines, efficient separations can be made of a mixture of thorium and uranium, plutonium and
uranium, and also a mixture of trans-uranium elements.
90
80
70
0- 60
? 50
"-d 40
ie) 30
as
ill 20
10
111
1
1
[
NM
IPS
til
1,1K
Nil
IIINIT
all
Mill
ILI
0 1 2 4 6 8 10
Concentration of acid, N
00
Fig. 7. Curves showing the dependence of ex-
traction of uranium by a 0.1 M solution of
tridecylamine in kerosene on the concentration
of mineral acids in the initial solutions.
The maximum selectivity is shown by tertiary alkyl-
amines, since they hardly extract bi- and trivalent iron,
aluminum, copper, calcium and other impurities found with
uranium. From sulfate solutions with pH = 1-1.5 molybdenum
is extracted together with uranium; however they can readily
be separated during reextraction, when the uranium is reextract-
ed with a 1 M solution of common salt at pH = 1 and molyb-
denum by a 510 solution of soda. Figure 8 shows the flow
sheet for the extraction of uranium and molybdenum by tri-
n-octylamine from a sulfate percolation solution.
For the reextraction of uranium from alkylamines the
most efficient are solutions of chlorides, nitrates, soda and
ammonium carbonate. The reextraction is entirely finished
in 3-5 stages with a ratio of volumes of organic phase to
aqueous phase equal to 6-8. During extraction from sulfate
solutions the reextraction of uranium is best carried out with
a 2.5-3 M solution of ammonium sulfate at pH. 4.5.
Extraction from solutions can therefore be accomplished
by a large number of solutions with varying and often complex
salt compositions. Due to their high selectivity, extraction
processes have been used in the first place to purify uranium
and also in those cases where the use of sorption processes
causes difficulties, for example, when extracting uranium from
concentrated solutions of phosphoric acid. When treating a
number of solutions with low uranium content it is economi-
cally advisable to use sorption, based on the fact that losses of
extraction agents are proportional to the volume of the solutions
being processed. Methods have now been developed for the
extraction of uranium from practically all types of solutions
found in the uranium industry.
Extraction of Uranium from Thick Ore Pulps
The extraction of uranium from solutions is applicable mainly in those cases where colored solutions are
obtained comparatively simply: in percolation, acid filtration or countercurrent decantation.
The extraction of uranium directly from pulps has a number of important advantages: it eliminates the
need for classification of sands, the need to thicken slurries, repulping and filtration [7). The most promising
method for processing uranium ores is the extraction of uranium from thick ore pulps or pastes. The losses of
the extraction agent depend to a large extent on the concentration of the solid material in the pulp (Fig. 9).
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Ore
1/279
Percolation
leaching
Solution
Saturated ex-
traction a?e
Mother
li'uor rejected
Extraction agent
(0.1 M
trioctyl;
amine in
Kerosene
Ma CO3
Regenerated
NHJ Precipitation
or uranium
Filtration
Mother liquor rejected
Diuranate
Fig. 8. Flow sheet for the extraction of uranium
from percolation solutions by a 0.1 M solution of
tri-n-octylamine.
10
Ore
round to Acid
40-60 mesh
III IIII
? 11V6.41X-1.' !!!!
ISM
2
To process
11\
Fresh
1-, extraction
agent
To precipialm
tation of
chemical 19
concrete
200
90 80 70 60 50 40
Content of solid material in pulp, cria
Fig. 9. Losses of extraction agent in the extrac-
tion of uranium from thick pulps.
It can be seen from Fig. 9 that the minimum losses
of extraction agent are found in very thick ore pulps or
pastes. ?The flow sheet for the extraction of uranium from
ore pastes is given in Fig. 10. The non-aqueous extrac-
tion leaching of uranium is very promising for certain
types of ores.
f4 Vapors of
Itextraction agent
17\
411-1"
4
l'Ste
_1 I
Condensate Spent
petont
- orew gate
Water
To cooling
_ J
and return
Fig. 10. Flow sheet for the extraction of uranium from ore pastes; 1) feeder; 2) tank; 3) apparatus for
mixing; 4) apparatus for drying and cooling; 5) extraction fan; 6), 10) header tanks; 7), 9), 13), 16),
19) pumps; 8) filter press; 11) reextraction column; 12) contact vessel; 14) cooler; 15) tank; 17) appa-
ratus for distillation of extraction agent; 18) collector.
441
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TABLE 4
Extraction of Uranium by 0.1 M Solutions of Long Chain Amines
Amines
Chemical formula
Molecular
weight
Solvent ?
K of hexavalent uranium
1 g/ liter of
uranium
(pH= 1)
1g/ liter uran-
nium and
50 g/ liter
SO4 ions
(pH = 1)
Primary Amines
n-Decylamine
n-Dodecylamine
Secondary Amines
Di -n-hexylamine
Di -n-octylamine
Di -(2-ethylhexyl)amine
Di-n-decylamine
Methylstearylamine
N-benzylheptylamine
N-benzyloctylamine
N -benzylhexyl-( 2-ethyl-
amine)
N-benzyl-n-nonylamine
N-benzyl-n-decylamine
N-benzyl-n-dodecyl-
amine
N-benzy1-1-(3-ethyl-
penty1)-4-ethyloctyl-
amine ")
Tertiary Amine
Triisoamylamine
Tri-n-hexylamine
Tri-n-octylamine
Tri-(2-ethylhexyl)amine
Tri-n-decylamine
Octyldimethylamine
Decyldimethylamine
Stearyldimethylamine
N-benzyl-n-dioctyl-
amine
N-benzyl-n-dodecyl-
amine
C 10H2INH2
C 121-123N H2
(C61-113)2NH
(C81-117)2NH
(C81117)2NH
(c10H22)2NH
Ci8H37NHC H3
C6H5C H2NHC His
C6H3C H2NHC8H1.7
C6H5CH2NHC8H17
C6H5C H2NHC9HD
C6H5CH2NHCi0 H21
C6H5CH2NHCl2H25
C6H5CH2NHC1IH35
i(C31-111)3N
(C 6H13)3N
(C8H17)3N
(C 81-103N
(C 101121)3N
C 8HIAC H3)2
C 10H2IN(C H3)2
C 181137N(C H3)2
C H2N C2H17)2
C6H5C112MC 10H21)2
157
185
185
241
241
297
283
205
219
219
233
247
275
345
227
269
353
353
437
167
195
297
331
387
}
50% alcohol,
50% kerosene
5 % alcohol,
95% kerosene
50% alcohol,
50% kerosene
90% kerosene,
10% alcohol
90% kerosene,
10% alcohol
kerosene
5% alcohol,
95% kerosene
100% alcohol
5% alcohol,
95% kerosene
61.6
22.8
2.55
172.0
110.0
64.0
10.0
12.1
154.0
36.4 '
44.7
41.0
23.0
8000
0.048
114.0
200.0
1.0
172.0
0.05
0.38
5.10
23.0
23.0
8.6
22.7
21.7
-^
63.0
154.0
36.4
11.5
0.147
88.0
180.0
1.0
96.0
0.07
1.12
3.55
26.4
8.6
*The solvent was sulphona
? ?From the data of [6].
ted kerosene and isooctyl alcohol.
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In this method ore ground to 40-60 mesh with a moisture content of 5-10% is subjected to acid leaching by
the batch method. The consumption of acid, the duration of leaching and the necessity for heating depend on
the character of the ore and the required degree of extraction. After acid treatment, the pulp usually contains
15-20% moisture. The leaching process is best carried out with the extraction, which leads to a considerable
reduction in the consumption of acid and to a reduction in gas evolution in a number of cases. Leaching by the
method of acid mixing with extraction is carried out in blade mixers with a countercurrent arrangement. When
using nitric acid for the leaching it is possible to use extraction agents such as isoamyl acetate, dimethyl
phthalate, dibutyl ether or 5-10% solution of TBP in kerosene, which gives high extraction of the metal with a
minimum consumption of acids in the leaching.
Further extraction of the extraction agent from the ore cakes can be carried out by the following methods:
1) distillation with heating; 2) washing the cakes with organic solvents (kerosene, hexane) and 3) repulping with
water.
The extraction agent saturated with uranium is subjected to reextraction, water being used to reextract
uranium from ethers and esters of carboxylic acids; uranium can be reextracted from TBP by a 3-5% solution
of sulfuric acid, a 5-10% solution of ammonium sulfate or by ammonium carbonate solutions.
LITERATURE CITED
1. M. Branica, et al., Croatica Chem. Acta, 28, (1956).
2. V. B. Shevchenko, et al., Atomnaya Energ. 7, 3, 236 (1959).?
3. Blake, et al., Proceedings of the Second International Conference on the Peaceful Uses of Atomic Energy
(Geneva, 1958). Selected Reports of Foreign Scientists. The Technology of Atomic Raw Materials.
(Atomizdat, Moscow. 1959) Vol. 7. p. 393.
4. Brown, et al., Proceedings of the Second International Conference on the Peaceful Uses of Atomic Energy
(Geneva, 1958). Selected Reports of Foreign Scientists. The Technology of Atomic Raw Materials. (Ato-
mizdat, Moscow, 1959) Vol. 7, p. 324.
5. Kraus and Nelson, Materials of the International Conference on the Peaceful Use of Atomic Energy (Geneva,
1955) (Metallurgizdat, Moscow, 1958) Vol. 7, p. 144.
6. Colman, et al., Proceedings of the Second International Conference on the Peaceful Use of Atomic Energy
(Geneva, 1958). Selected Reports of Foreign Scientists. The Technology of Atomic Raw Materials. (Ato-
mizdat, Moscow, 1959) Vol. 7, p. 352.
7. F. Grunstedt, Materials of the International Conference on the Peaceful Use of Atomic Energy (Geneva,
1955) (Metallurgizdat, Moscow, 1958) Vol. 8, p. 90.
'Original Russian pagination. See C. B. translation.
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Declassified and Approved For Release 2013/02/19: CIA-RDP10-02196R000100050006-3
INTERACTION OF URANIUM HEXAFLUORIDE WITH AMMONIA
N. P. Galkin, B. N. Sudarikov, and V. A. Zaitsev
Translated from Atomnaya Energiya, Vol. 8, No. 6, pp. 530-534
June, 1960
Original article submitted July 15, 1959
The interaction of uraniumhexafluoride with ammonia in the temperature range from -50 to +200?C has been
studied. The total equations of the reactions are proposed:
1) for the temperature range -50 to -30?C
6UF6+8NH3 ?> 6UF5+6NH4F+N2;
2) for the temperature range 0-25?C
41JF6+8NF3 ---> 2UN5+2NH4UF5+4NH4F?N2;
3) for the temperature range 100-200?C
3UF6-1-8N1-13 3NH413F5+3NH4F+ N2.
The rate of the reaction in the temperature range between -20 and +20?C has been evaluated. The thermal
effect of the reaction in the temperature interval between -50 and -30?C varied between 50.8 and 83.6 kcal/ mole,
and, at -40?C it was in satisfactory agreement with the calculated value obtained from the proposed equation.
A whole series of papers was devoted to the interaction of uranium hexafluoride with various reducing agents.
The reduction of uranium hexafluoride by hydrogen [1-3], by hydrogen chloride and bromide [4, 5], by carbon
tetrachloride [6], by trichloroethylene [3], by ethylene and propane [7], etc., were studied. It has been pointed
out [8] that uranium hexafluoride interacts with ammonia already at the temperature of dry ice (-72?C); in gas
phase uranium hexafluoride is reduced by ammonia at a temperature of 300?C with formation of solid product
containing 98% NH4UF5 and 2% of NH4F. No data concerning the details of this problem are available.
In the present work we studied the interaction of uranium hexafluoride with ammonia in the temperature
range from -50 to 200?C, with the purpose of determining the total equations of the reactions at different tem-
peratures, and also the interaction rates of the reagents and the thermal effect of the reaction.
The interaction of solid uranium hexafluoride with liquid and gaseous ammonia was studied on an apparatus
shown in Fig. 1.
Solid uranium hexafluoride (2.0-4.0 g), liquid ammonia in a quartz ampoule [0.01-0.5 g for the reaction
UF6(sol) + NH3(g), and 0.5-2.0 g for the reaction UF6(501)+ iN H3 ( were placed in a stainless steel bomb under
thermostatization. After the required temperature had been reached, the ampoule containing the ammonia
was broken by a special device. The pressure of the system was controlled by means of a manometer. The
solid and gaseous phases were analyzed. In some cases the reaction products were subjected to x-ray and ther-
mal analysis.
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Liquid
nitrogen'
vacuum
Fig. 1. Scheme of the apparatus for studying the
interaction of solid uranium hexafluoride with
liquid and gaseous ammonia; 1) breaking device;
2) reaction bomb; 3) solid uranium hexafluoride;
4) quartz ampoule with liquid ammonia; 5) base
for the ampoule; 6) manometer; 7) Dewar vessel;
8) solution of the salts NHIC1, NaCl, ZnSO4,
CaC12, etc.
The interaction of uranium hexafluoride with
ammonia in the gas phase was studied with the apparatus
whose scheme is given in Fig. 2.
The reacting gases were previously heated at the
given temperature; uranium hexafluoride was either
diluted with argon or placed in the reactor without addi-
tions.In the first case the reaction proceeded more slow-
ly, without clogging the connection of the apparatus and
thus making the investigation simpler. The solid and
gaseous products of the reaction were analyzed.
The thermal effect of the reaction between solid
uranium hexafluoride and liquid or gaseous ammonia de-
pending upon the temperature of the reaction was deter-
mined by a direct calorimetric method with a special
apparatus (Fig. 3). The weight of uranium hexafluoride
was 0.1-2g, the weight of ammonia was 0.01-0,2g for the
reaction UF6(S01) + NH3( g), and 0.5-2g for the reaction
UF6sol) + NH30)? The temperature of the calorimetric (
liquid (ethyl alcohol) increased by 0.2-1?C at the expense
of the heat of reaction.
The investigations carried out showed that uranium
hexafluoride, at temperatures between -50 and +200?C
reacts with ammonia and forms a mixture of solid products.
On the basis of visual observations it was found that in the
temperature range from -50 to -30?C a white solid product
was formed, which rapidly became green in air. At a
temperature of -20 to 0?C a mixture of a white, a gray, and
a green product was obtained; they took a stable green
color by standing in the air. Finally, at temperatures
above 0?C, a green product was mainly obtained. However, it should be noted that in all cases, the composition
of the solid phase included a white product becoming green in the air (uranium pentafluoride) when the experi-
ments were done in an appropriate manner (rapid interruption of the reaction by cooling the reaction appa-
ratus with liquid nitrogen). Already these observations permitted us to conclude that the reduction of uranium
hexafluoride by ammonia is a complicated reaction, one of whose steps is the formation of uranium pentafluo-
ride.
The results of the chemical analysis of the solid products obtained upon interaction of uranium hexafluo-
ride with ammonia in the temperature range under study are given in Table 1.
As can be seen from these data, in the temperature range from -50 to -30?C a solid product having a rough-
ly constant chemical composition is formed. According to the results of x-ray analysis, uranium tetrafluoride
was not present in this product. It follows that in the reaction of uranium hexafluoride with ammonia in the tem-
perature range from -50 to -30?C, only uranium pentafluoride was formed, and the fluorine ion separated form-
ing with ammonia, ammonium fluoride. It should be noted that the ammonia in these reaction products was
somewhat more than required by the stoichiometry of the formation of ammonium fluoride; the latter circum-
stance can be explained both by the formation of uranium pentafluoride ammoniates, and by the mechanical
capture of ammonia.
.The total reaction of interaction of uranium hexafluoride with ammonia in the temperature range from
-50 to -30?C is described by the following equation;
6UF6 + (84- 6n)N1i3
6UF0NI-13 61\114F I\12,
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Fig. 2. Scheme of the apparatus for the study of the in-
teraction of uranium hexafluoride with ammonia in gas
phase: 1) Argon cylinder; 2) monostat; 3) flowmeter;
4) water thermostat; 5) uranium hexafluoride vaporizer;
6) air thermostat; 7) preliminary heater of uranium hex-
afluoride; 8) reactor; 9) ammonia cylinder; 10) drying
columns with KOH; 11) preliminary heater of ammonia;
12) condenser; 13) trap.
Liquid
nitroger1-1-?
/4
13
vacuum
- vacuum
Fig. 3. Scheme of the calorimetric apparatus; 1) Dewar
vessel; 2) external wall of the calorimeter; 3) internal
wall of the calorimeter; 4) stirrer; 5) heater; 6) stoppers
of a thermal insulator; 7) breaking device (of ebonite);
8) quartz ampoule with liquid ammonia; 9) solid uranium
hexafluoride; 10) resistance thermometer; 11) reaction
bomb; 12) thermal insulator; 13) liquid nitrogen vapo-
rizer; 14) solution of the salts NH4C1, NaCl, ZnSO4,
CaCl2, etc.
446
where n = 0.73. The calculated and experimental
data (mean values in lo) concerning the composition
of the solid phase are reported below:
Data
Utot
NH3
Calculated
50.0
62.2
29.8
7.6
Experimental
50.6
61.2
31.0
7.6
*In all such results we show the percent contents
of U+4 with respect to its total amount.
4
For higher temperatures (-20 to 0?C) the re-
duction reaction of uranium hexafluoride by ammo-
nia did not stop at the formation of uranium penta-
fluoride; it proceeds further; in the reaction pro-
ducts uranium ammonium pentafluoride was detec-
ted (from data of differential-thermal analysis and
x-ray structure analysis). In the temperature range
from 0 to +25?C, a product having a roughly constant
chemical composition was formed; this product cor-
responded to a5010 reduction of the uranium penta-
fluoride formed to uranium ammonium pentafluo-
ride. The total reaction is described by the follow-
ing equation
4UF0 8NH3
- - 2UF5 2NH4UF5 4N144F -I- N2.
The calculated and experimental data (mean values
in /0) concerning the composition of the solid phase
are given below
Data
U+4
UtOt
F
NH3
Calculated
75.0
62.8
30.0
6.6
Experimental
77.6
63.1?
29.0
6.7
For still higher temperatures (100-200?C) the
uranium pentafluoride was completely reduced to
uranium ammonium pentafluoride. The total reac-
tion of reduction of uranium hexafluoride by ammo-
nia in this temperature range is described by the
equation
31.1F6 ? 8NH3 ---> 3N114UF3 3N114F N2*
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TABLE 1
Composition of the solid products of the reaction between
uranium hexafluoride and ammonia.
Reaction
temperature
?C
Degree of
reduction
of ora-
nium, %
Contents of the individual
components in the reac-
tion products, 70
u
F
NH3
-50
50,9
61,3
30,5
7,8
-40
50,3
60,8
31,3
7,6
-30
50,6
61,6
30,4
7,7
-20
59,5
60,2
32,5
6,9
-15
64,4
58,7
33,9
7,2
-10
71,4
59,8
33,1
7,1
-5
73,8
60,0
32,8
6,8
0
77,5
62,9
30,2
6,7
+15
77,6
62,2
31,0
6,6
+25
77,7
63,2
29,7
6,7
+100
98,7
61,6
29,1
9,2
+150
99,1
62,0
29,3
8,9
J-200
99,5
62,2
29,0
8,7
Note; The gas
phase contained in all cases nitrogen; a
quantitative analysis of the gas phase relating to the
content of the nitrogen was carried out only for the
reactions performed in the temperature range 100-200?C;
the data obtained in these experiments agreed well with
the calculated data referring to the contents of nitrogen
computed from the equation of the high-temperature
reaction between uranium hexafluoride and ammonia
given below.
The calculated and experimental data (average
values in /0) concerning the composition of the
solid phase are reported below;
Data
U+4
Utot
F
NH3
Calculated
100.0
61.3
29.4
8.7
Experimental
99.1
61.9
29.1
8.9
The rate of the reaction between uranium
hexafluoride and ammonia in the temperature range
from -20 to +20?C was determined from the change
of the pressure of ammonia in a closed volume.
This could be done due to the fact that at low tem-
peratures the reduction reaction of uranium hexa-
fluoride by ammonia leads to the formation mainly
of solid products; as to the nitrogen produced, its
quantity was comparatively unimportant with res-
pect to the amount of ammonia that had reacted,
the more so because it could always be taken into
account. Naturally it was also necessary to take
into account the influence of the thermal effect on
the change of the pressure of ammonia. The results
of the determination of the rate of the reaction
between solid uranium hexafluoride and gaseous
ammonia are presented in Fig. 4. The values of the
pressure changes for ammonia were calculated in
moles according to Planck's [9] formula. As can be
seen from Fig. 4, the rate of the reaction, as could
be expected, decreased when the temperature de-
creased. Even at a temperature of -20?C the reaction
stopped in 3-5 min.
In the experiments carried out it was not
possible to keep strictly constant the surface of the
solid uranium hexafluoride, and also to avoid a
certain increase of the temperature within the bomb, due to the thermal effect of the reaction; therefore we did
not carry out a mathematical elaboration of the results obtained. However, preliminary calculations showed that
the order of the reaction decreased from 2 in the first 5-10 sec to 0.5 after 3-5 min. This circumstance was
apparently due to surface effects (diffusion of ammonia, absorption of ammonia by solid uranium hexafluoride,
and by the reaction products with subsequent interaction). The experiments carried out at temperatures of
100-200?C showed that uranium hexafluoride in these conditions interacted practically instantaneously with
ammonia.
By direct calorimetric measurements we established that the reaction of uranium hexafluoride with ammo-
nia took place with a large production of heat. The values of the thermal effect of this reaction in the temper-
ature range from -50 to -30?C are shown in Table 2.
The thermal effect of the reaction calculated according to the equation of low temperature interaction of
uranium hexafluoride with ammonia (at a temperature of -40?C), appears to be equal to 65 kcal/mole, in
satisfactory agreement with the value found experimentally for the thermal effect of the reaction at this tem-
perature. This circumstance supports again the correctness of the total equation proposed above for the reaction
between uranium hexafluoride and ammonia at low temperatures.
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(125, 250
e 200
150
(T)
o 50
1
2
3
0 1 2 3 4 5 8 7 89 tO
Time, min
Fig. 4. Dependence of the rate of the reaction between
TABLE 2
Thermal effect of the reaction of interaction of
solid uranium hexafluoride with liquid ammonia.
Reaction
temperature, ?C
Thermal effect of the
reaction kcal/ mole
?50
50,8+1,5
?45
59,1?1,8
?40
(i7,01-2,(
?35
75,4?2,3
?30
83,6?2,5
*Average of 3-5 experiments.
ammonia and solid uranium hexafluoride upon time. Preliminary experiments showed that the ther-
Reaction temperature, ?C: 1) -20; 2) 0; 3) +20. mal effect of the reaction between solid uranium
hexafluoride and gaseous ammonia increased rapidly,
starting with temperatures of -30 to -25?C; this fact was apparently related, first, to the increase of the heat of
reaction due to the complete reduction of uranium pentafluoride, and, second, to the formation of uranium
ammonium pentafluoride. The values of the thermal effects for this temperature range are being refined at
present.
LITERATURE. CITED
1. 0. Ruff and A. Heinzelmann, Z. anorgan. und allgem. Chem. 72, 82 (1911).
2. J. Dawson, D. Ingram, and L. Bircumsnaw, J. Chem. Soc. (Dec.), 1421 (1950).
3. Smiley and Breiter, Transactions of the Second Conference on the Peaceful Uses of Atomic Energy
(Geneva, 1958). Selected Communications of Foreign Scientists, Technology of Atomic Raw Materials.
[Russian translation] (Atomizdat, Moscow, 1959) Vol. VII, p. 561.
4. E. Gladrow and P. Chiotty. Report CK-1498 (1944).
5. K. David and G.Hugh. U.S. Patent, No. 2768872 (Oct. 30, 1956).
6. Nern, Collins, and Taylor. Proceedings of the Second International Conference on the Peaceful Uses of
Atomic Energy (Geneva, 1958), Reports of Foreign Scientists. Atomic Technology. Series [Russian trans-
lation] (Atomizdat, Moscow, 1959) p. 553.
7. J. Katz and E. Rabinovich, The Chemistry of Uranium [Russian translation] (IL, Moscow, 1954) Vol. 1,
p. 356.
8. B. Ayers, Report CC-1504 (1944).
9. R. Planck, Z.Teclui. Phys. 5, 9, 132 (1924).
448
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Declassified and Approved For Release 2013/02/19: CIA-RDP10-02196R000100050006-3
THE FLOCCULATION OF PULP AND POLYACRYLAMIDE-TYPE
FLOCC ULEN TS
I. A. Yakubovich
Translated from Atomnaya Energiya, Vol. 8, No. 6, pp. 535-541,
June, 1960
Original article submitted November 6, 1959
The processes of dividing hydrometallurgical pulps into solid and liquid phases with subsequent washing out of
the dispersed solid particles from the solute are very important in the technology of processing ores and con-
centrates of uranium, lithium, zirconium, and other metals. Flocculation of pulp, i.e., the formation of more
or less large and strong aggregates of particles, is accomplished by special flocculating reagents. Their action
facilitates dividing and washing out the solid phase from the solute. This article describes the properties and
methods of obtaining the most effective flocculents and gives the results of an investigation of the flocculation
processes of aqueous, acid, and carbonaceous hydrometallurgical pulps. The article is of interest for a wide
circle of specialists working in industrial enterprises, in factory laboratories, and in research institutes studying
the concentration of ores and the hydrometallurgy of uranium and other metals.
High-molecular weight or surface -active substances used as flocculents must satisfy the following basic
requirements. The molecules of a flocculating reagent must contain active groups capable of being adsorbed
on the surface of solid particles and must be linear and of definite length in order to form bridges between the
particles. Flocculating reagents must be soluble in water; the reagents must be capable of reacting even when
only very low concentrations of them are present in the pulp. The floccules must be strong enough to form
porous cakes during filtration of the pulp and to retain as little liquid as possible during coagulation by settling.
The intensity of the formation of floccules, their shape, size, and strength depend on the physicochemical
properties of the suspension being flocculated and on the flocculating reagent. Here decisive importance is
attached to the ratio of the liquid phase to the solid phase in suspension, the size and composition of the particles
of the solid phase, the ion composition of the liquid phase, the concentration of the flocculent in the working
solution introduced into the suspension and the method of mixing this solution with the suspension. The presence
in the pulp of precipitates of metal hydroxides suppresses the action of the flocculents. On the other hand the
effectiveness of their action increases in diluted pulps containing small amounts of electrolytes and also with a
decrease in the concentration of the working solution of the flocculent. The latter requires the use of solutions
with a concentration of 0.5-10 g/liter in practice. Circulating solutions can be used for the preparation and
dilution of the flocculents in conformity with the technological pattern of the production cycle.
When introducing the flocculent into the pulp, it is necessary to see that it is rapidly and uniformly distri-
buted throughout the entire mass of the pulp. However,any intensive mechanical mixing of the pulp, during
which the floccules forming can be irreparably destroyed, is completely inadmissible. Under production con-
ditions the most acceptable method is to introduce the flocculent into a stream of pulp flowing in a pipe, trough,
or channel. In many cases the effectiveness of flocculation is increased if the flocculent is introduced into the
pulp not all at once but in individual portions. In this case the pulp should be well mixed with the flocculent
after the introduction of each portion.
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The addition of inorganic compounds (lime, iron and aluminum salts, acids, alkalis) and organic substances
(animal glue, starch, flour, etc.) has been used for many years to accelerate coagulation and filtration of pulps.
We proposed new flocculents of natural origin: VL on a base of seaweed, KLZh on a base of linseed cake, M-42
on a base of potato pulp, etc. [1]. The flocculating power of these reagents is due to the fact that they contain
high-molecular weight compounds belonging to the classes of polysaccharides and proteins. As a result of ex-
perimental investigations it was established that the new flocculents are more effective than the reagents pre-
viously used during coagulation of hydrometallurgical [2-4] and coal slime pulps [5].
In recent years synthetic flocculents which are high-molecular weight water-soluble substances have ac-
quired increasing importance. In spite of the comparatively high cost of synthetic flocculents, their use is prof-
itable,since in effectiveness of action they considerably excel all flocculents of natural origin [3-8]. A list of
foreign high-molecular weight synthetic flocculents and starting materials for their production was published
in [9]. In our country and abroad synthetic flocculents are produced under various names given to these re-
agents by the inventors of the methods producing them [1, 9-12]. The results of laboratory investigations [3, 5,6]
and the experience of industrial enterprises of the coal [7, 13, 14] and metallurgical branches of industry [15-17]
make it possible to consider that the most effective and universal flocculents are reagents of the polyacrylamide
type, among which we may cite Separan 2610, which is produced in the USA by the Dow Chemical Company,
the polyacrylamide produced by the Institute of High-Molecular Weight Compounds of the Academy of Sciences,
USSR and the Institute of Halurgy, and a polyacrylamide flocculent (AMF), which is produced by our method at
one of the factories in the Gorkyi economic region.*
The starting material for the synthesis of the polyacrylamide-type compounds is acrylonitrile (AC). Seve-
ral industrial methods for synthesizing acrylonitrile are known. The most widely used methods are those based
on the reaction of acetylene or ethylene oxide with prussic acid according to the formulas
CH CH + HCN CH, = CH C N;
CH,OH
CH, = CH-1-02 H,C ? CH, + HCN
\ / CH,CN
_ 0
-1120
CH, = CH C N.
Other technological methods of synthesizing acrylonitrile and the kinetics of its polymerization are described
in [18]. Acrylonitrile is easily polymerized and forms polyacrylonitrile (PAC), which can be converted into
water-soluble compounds by hydrolysis with sulfuric acid or alkali solutions. The process of hydrolysis of PAC,
depending on the temperature and the concentration of the sulfuric acid, leads to the formation of products of
various chemical compositions. After hydrolysis of PAC with 75-95% cold sulfuric acid for 4 hrs, a water-sol-
uble polymer is produced which contains imide and amide groups. The reaction of PAC with 65-95% acid when
heated leads to the formation of polymers which dissolve only in dimethyl formamide. A reduction in the con-
centration of the sulfuric acid to 50% leads to an increase of polyacrylic acid in the hydrolysis products [19].
We also obtained water-soluble products by treating PAC with solutions of sodium hydroxide at 80-90?C for
10 hrs.
The products of the hydrolysis of PAC with sulfuric acid and an alkali can be used as flocculents. How-
ever, as our experiments showed, the effectiveness of these reagents is usually lower than the effectiveness of
Separan 2610 and the flocculent AMF obtained by the polymerization of acrylamide.
Of the many known methods of producing acrylamide the most important are those based on hydrolysis
of acrylonitrile in the presence of sulfuric acid and differing only in the method of separating the acrylamide
by diluting the reaction mixture with ice water, by neutralizing with lime, filtering off the calcium sulfate,
and concentrating the aqueous solution in a vacuum, by neutralizing the solution with ammonia, and many other
methods of separation and polymerization of acrylamide [20-26].
The method was developed together with engineers M. P. Vilyanskyi and N. P. Pashkin. Industrial application
of the method was accomplished with the active participation of factory workers E. A. Kulev and R. Z. Khantsis.
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Water
Thermocoupl
WaterN.
?4Th
Vapor.
\Titer
Water
The hydrolysis of acrylonitrile with sulfuric acid with
subsequent polymerization of the reaction products was the
basis of the method of producing AMF. Synthesis of the
reagent AMF was accomplished on an apparatus, the diagram
of which is shown in Fig. 1. The reaction between the
acrylonitrile and sulfuric acid proceeds very violently and
is accompanied by considerable generation of heat. An in-
crease in the temperature of the reaction mixture to 130?C
can cause dangerous splashing and explosions of the reaction
vessels. The possibility of dangerous complications arising
at this stage is eliminated if the acrylonitrile is added gra-
dually and uniformly to the reaction vessel containing the
sulfuric acid and polymerization inhibitor (powdered
copper, elementary sulfur), while the mixture is stirred
intensively. The behavior of the process at a temperature
of about 100?C enables the reaction to go to completion
in 40-50 min.
9 10 11 12 13 The following optimum conditions for realizing the
process were established by numerous experiments: 84.5%
sulfuric acid for a calculation of 1 mole H2SO4 and 1 mole
acrylonitrile was poured into the reaction vessel and ele-
Fig. 1. Diagram of the apparatus for producing mentary powdered sulfur or a solution of methylene blue
AMF flocculents: 1) reaction vessel; 2) service . added. The mixture was heated to 70-80?C and then the
tank for sulfuric acid; 3) service tank for acryl- acrylonitrile was added to the reaction vessel slowly and
onitrile; 4) reactor; 5) service tank for ammonia uniformly with the mixer operating. As a result of the
water; 6) reflux condenser; 7) vacuum filter; exothermal reaction the temperature of the mixture in-
8) service tank for ammonium persulfate; 9-13) creased to 100?C. If cold water (with a temperature of
polymerization chambers. 17-18?C) is circulated through the jacket surrounding the
reaction vessel, then a further increase in the temperature
is not observed. The entire portion of the acrylonitrile was added to the vessel in one hour, after which the re-
action mixture was held at 90?C for 45 min.
Neutralization of the solution after dilution with water can be accomplished by means of one of the follow-
ing reagents: calcium carbonate, lime, sodium hydroxide, sodium carbonate, water and gaseous ammonia. As
an initiator of polymerization we used ammonium persulfate activated by sodium sulfite. The solutions obtained
after neutralization of the acrylamide sulfate by ammonia without separation of the ammonium sulfate were
subjected to polymerization. The pH of the solution after polymerization fluctuated between 9 and 8.
The polymer forming under these conditions is a viscous, gelatinous mass which, during a more or less
long time, depending on the thickness of the layer, the temperature, and the exchange ratio, dries in the air
and is transformed into a light, brittle material containing from 35 to 50% active substances. The AMF floccu-
lent obtained readily dissolves in water during intensive mixing and forms homogeneous solutions.
Comparative investigations of the effectiveness of the reagent AMF and other flocculents of natural and
artificial origin were conducted on different kinds of slime pulps obtained by the metallurgical, coal, and
chemical industries. It should be noted that the presence in the pulp of precipitates of metal hydroxides and
colloidal silica considerably reduces the effectiveness of a flocculent's action.
The effect of additions of flocculents on the sedimentation of pulps was studied in measuring tanks. For
this purpose the change in time of the height of the layer of clarified liquid was recorded. In order to keep the
duration and the intensity of mixing the pulp with the flocculent solution constant, the tanks with the pulp were
attached in a special stand in which they were simultaneously turned over in order to mix the pulp with the floc-
culent solution added to it.
Vacuum
The results of processing the sedimentation curves made it possible for us to obtain not only a qualitative
but also a quantitative characteristic of the effectiveness of the flocculent's action in the form of value ex-
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pressing the height of the layer of the clarified liquid for a definite interval of time from the start of the ex-
periment, or in the form of an index of the requisite specific area of pulp coagulation.
Figure 2 contains curves describing the effect of the additions of various flocculents on the sedimentation
rate of pulp obtained after sulfuric acid leaching of ore containing 35% quartz and 55% feldspar.* The ratio of
the liquid phase to the solid phase in the initial pulp, which is the overflow of the classifiers, was 7.3 : 1; the
PH of the solution was 1 : I. The temperature of the experiments was 20?C. The solid phase of the pulp did
not contain particles larger than 0.074 mm. In this series of experiments the expenditure of the flocculents
was 100 g/m3. The shape of the curves did not change when the flocculents had other rates of expenditure.
From a comparison of the results obtained it can be seen that the effectiveness of the AMF synthetic flocculents
and Separan 2610 are approximately the same and considerably higher than flocculents of natural origin.
250
bA
0
ct
(1.)
Emio
E
4. .0
6,7.
10 20 30 4,0 50 60
Sedimentation time, min
Fig. 2. The effect of additions of flocculents on the
sedimentation rate of sulfuric acid pulp. The expen-
diture of the flocculents is 100 g/m3; 1) starting pulp;
2) carboxymethylcellulose; 3) M-42; 4) KLZh; 5) VL;
6) AMF; 7) Separan 2610.
The next series of experiments was made with
aqueous pulp, which was the overflow of spiral classi-
fiers operating in a closed cycle with ball mills at the
pulverization stage. The ore material contained up
to 70% clays(montmorillonite-kaolinite). The ratio of
the liquid phase to the solid phase in the pulp was
6 : 1, the solid phase of the pulp did not contain par-
ticles larger than 0.3 mm. Figures 3 and 4 contain
curves describing the sedimentation rate of the aqueous
ore at various rates of expenditure of the flocculents.
In this series of experiments the flocculent AMF was
a somewhat more effective reagent than Separan 2610.
The flocculating power of the reagent AMF is
manifested most clearly when it is introduced into
pulps obtained after soda leaching of the metals from
the ores. Figure 5 contains curves describing the pro-
cess of the sedimentation of pulp obtained after soda
leaching of ores containing about 25% carbonates and
65%aluminosilicates. The content of Na2CO3 in the li-
quid phase of the pulp was 25 g/liter; the ratio of the
liquid phase to the solid phase was 3 : 1; the solid phase
did not contain particles greater than 0.1 mm; the tem-
perature of the pulp was 20?C. The experiments were
made at various rates of expenditure of the AMF floc-
culents.
Figure 6 represents the results of the experiments showing the effect of the value of the ratio of the liquid
phase to the solid phase in carbonate pulp on the sedimentation rate. These dependences were obtained while
investigating an entire series of soda pulps obtained after leaching ores of various compositions. It was estab-
lished that the flocculent AMF at a rate of 50-150 g/m3 very effectively flocculates not only diluted but also
rather dense soda hydrometallurgical pulps containing up to 40% solid phase. In this case a high productivity
of coagulation exceeding 10 tons per m2 per day and the production of coagulated pulp with a ratio of the liquid
to the solid phase of about 0.8 : 1 are assumed. These circumstances are especially important when organizing
the technological process of treating the ores with the use of continuous counterflow decantation washing of the
pulp after leaching with the product of the clarified metalliferous solutions for their further processing by extrac-
tion, sorption, or other methods.
The effectiveness of continuous counterflow decantation washing can be calculated by the following for-
mula:
Tvin+l?NE
E ? 100%,
mn+l_i
V. N. 15alagina performed the experiments.
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ow
Ia
b0
0
C.)
4.4
0
0
C.1
(1)
0
Ia
IDA
250
200
150
100
50
0
Sedimentation time, min
Fig. 3. The effect of additions of flocculents on the
sedimentation rate of aqueous ore pulp. The expen-
diture of the flocculents is 20 g/m3: 1) Starting pulp;
2) seaweed grit BVLK; 3) Separan 2610; 4) AMF.
E 250
04200
-0
2
cd
t% 150
co
c.)
4.
o 100
iso
bo
10 20 30 40 50
X Sedimentation time, min
Fig. 5. The effect of the expenditure of the reagent
AMF on the sedimentation time of the carbonate pulp;
1) starting pulp. Expenditure of AMF (g/m3); 2) 20;
3) 50; 4) 100; 5) 200.
60
E 250
c'" 200
150
b0
o^ 100
50
0
bO
10 20 30 40 50
Sedimentation time, min
60
Fig. 4. The effect of addition of flocculents on the
sedimentation rate of aqueous ore pulp. The expen-
diture of the flocculents is 50 g/m3; 1) starting pulp;
2) M-42; 3) KLZh; 4) VL; 5) BVLK; 6) Separan 2610;
7) AMF.
E 250
ct
200
-ci
tio 150
cd
4-4
0
100
^ 50
0
??01.
10 20 30 40 50
Sedimentation time, min
2
3
4,
60
Fig. 6. The effect of the ratio of the liquid phase to
the solid phase in carbonate pulp on the effectiveness
of flocculation with the addition of 100 g/m3 AMF.
Ratio of the liquid phase to solid phase in initial pulp:
1) 1.7 : 1; 2) 3.1 : 1; 3)4.4 :1; 4) 5.0 : 1. Ratio of
the liquid phase to solid phase after 24 hrs: 1) 0.7 ; 1;
2) 0.91 : 1; 3) 0.94 : 1; 4) 0.96 : 1.
where E is the effectiveness of washing (10); M is the modulus of the washing expressing the ratio of the quantity
of the liquid removed in overflow to the quantity of liquid remaining in the coagulated pulp at each stage of
coagulation; n is the number of coagulation stages in the process. Figure 7 depicts this dependence graphically.
It is seen from the figure that with one and the same number of coagulation stages the effectiveness of washing
sharply increases with an increase in the modulus of washing or, what amounts to the same thing, with an in-
crease in the density of the coagulated pulp.
Thus the use of the flocculent AMF can assure high technical-economic indexes of the process of con-
tinuous counterflow decantation washing of metal. In this case acomparatively small yield of commercial
metalliferous solutions is assured per ton of ore when their concentration is sufficiently high.
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100
99
98
96
95
94
93
92
91
90
2 3 5 5n
Fig. 7. The dependence of the effectiveness of con-
tinuous counterflow decantation washing on the value
of the washing modulus M and the number of stages
of washing n.
8. Chem. Age 77, 1969, 604 (1957).
9. A. K. Livshits and L. G. Obreimova, Flotation. A Collection of Articles of the Scientific-Technical
Society of Nonferrous Metallurgists [in Russian] (Metallurgizdat, Moscow, 1956); A. K. Livshits, Khim.
Nauka i Prom. 4, 5, 622 (1959).
10. D. Pye, Eng. and Mining J. 156, 11, 94 (1955).
11. M. McCarty and R. Olson, Mining Eng. 11, 1, 57 (1959).
12. A. K. Livshits and L. I. Gabrielova, Byull. Tsentr. Inst. Informatsii Tsvetnoi Metallurgii, 21, 12 (1957);
Gornyi Zh. 35, 2, 67 (1960).
13. H. Charmbury, Mechanization 21, 9, 60 (1957).
14. B. Franke, Schligel und Eisen, No. 12, 899 (1957).
15. Mining J. 244, 6237, 239 (1955); No. 6238, 271 (1955).
16. E. Brocke, Gliickauf 95, 22, 1365 (1959).
17. M. Clement, Z. Erzbergbau und Metallhiittenwesen 10, 9, 421 (1957).
18. K. Sennewald and K. Steil. Chem-Ingr.-Tech. 30, 7, 440 (1958); Chem. Eng. 66, 4, 55 (1959); M. I.
Yakushkin, Khim. Prom.,No. 7, 575 (1959); Chemik 12, 7, 316 (1959); A. Parts, J. Polymer Sci. 37, 131
(1959).
19. E. A. Sokolova-Vasil'eva, G. I. Kudryavtsev, and A. S. Strepikheev, Zhur. Priklad. Khim. 31, 5, 785
(1958).?
20. E. Carpenter and H. Davis, J. Appl. Chem. 7, 12, 671 (1957).
21. M. N. Savitskaya, Zhur. Priklad. Khim. 32,8, 1797 (1959). ?
22. A. N. Angelov and T. P. Tishchenko, Khim. Prom. 6, 537 (1959).
23. Ts. Shindelarzh and V. Kolarzh, Tsvetnye Metally 32, 8, 5 (1959).
24. I. N. Plaksin and V. I. Zelenov, Byull. Tsentr. Inst. Informatsii Tsvetnoi Metallurgii, 20, 12 (1959).
25. H. Morawetz and T. Fadner.Macromolek. Chem. 34, 10, 162 (1959).
26. T. Suen, Yun-Jen and J. Lockwood, J. Polymer. Sic. 31, 481 (1959).
The introduction of flocculents into industry
will make it possible to improve considerably the quan-
titative and qualitative indices of the process of coa-
gulation and sedimentation of pulp.
LITERATURE CITED
1. I. A. Yakubovich,Tsvetnye Metally 30, 12, 9
(1957).
2. S. F. Kuz'kin and V. P. Nebera, Byull. Tsentr.
Inst. Informatsii Tsvetnoi Metallurgii 13, 10
(1957); Izv. VUZ. Tsvetnaya Metallurgiya
3, 44 (1959).
3. A. D. Mayants and F. I. Barotitskaya, Tsvetnye
Metally 21, 1, 44 (1958).
4. N. R. Romanov, Izv. VUZ. Tsvetnaya Metallur-
giya 2, 2, 32 (1959).
5. G. I.?Preigerzon.Koks i Khimiya 28, 1, 17 (1959).
6. V. La Mer, R. Smellie, and Pui-Kum Lee, Colloid
Sci. 12, 2, 230 (1957).
7. F. Drexler, Gliickauf 92, 35-36, 1023 (1956).
*Original Russian pagination. See C. B. translation.
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DETERMINATION OF ABSORBED DOSES IN ORGANISMS
EXPOSED TO EMANATIONS AND THEIR DAUGHTER PRODUCTS
L. S. Ruzer
?
Translated from Atomnaya Energiya, Vol. 8, No. 6, pp. 542-548,
June, 1960
Original article submitted September 11, 1959
Earlier, we considered the problem of absorbed doses created by short-lived a emitters in inhaling radon [3].
The present paper describes certain methods for determining absorbed doses due to radon itself, short-lived RaB
and RaC 8 -emitters, and long-lived elements of the radon family. Similar calculations were performed for the
thoron and actinon families, as a results of which new values for the maximum allowable concentration of
thoron and actinon in air are recommended. The proposed method for determining absorbed doses can be used
in the case where elements of any radioactive chain have penetrated the organism. It is demonstrated that per-
sonnel health monitoring with respect to the y radiation of emanation daughter products that have settled in the
respiratory system is a problem which can be solved in a practical manner.
The following peculiarities are encountered in determining the amount of absorbed energy when the or-
ganism is exposed to the radiation of emanations and their daughter products.
1. Due to the presence of a radioactive chain, the actual irradiation is due to the isotopes which directly
penetrated the organism as well as those which are formed in the organism during the delay of each member
of the radioactive family;
2. Among the isotopes participating in irradiation, there are a emitters as well as 8 and y emitters;
3. These isotopes have different half-lives [from a few microseconds (RaC) to several tens of years
(RaD)) and different biological half-lives;
4. The energy per single decay is very large (for radon and thoron, it is equal to ^,20 Mev due to the
a emitters alone);
5. All the emitters of the radon family can be divided into three fractions with respect to the relation-
ship between the biological half-life and their half-life (Xb and Xh, respectively): a) radon, for which Xh >> Xh
(Xeff Xb), b) short-lived radon decay products (RaA, RaB, and RaC), for which, according to data from [1 and
2], X.h 1)monoatomic gaseous retarder.
1. Let us assume that the retarding medium is in the form of a collection of alternating plates of thick-
ness 2a and 2b with uniformly distributed sources of monoenergetic neutrons, but with nonuniform absorption:
The plates with thickness 2a contain a uniform mixture of retarder nuclei and nuclei that resonate in the super-
thermal region of neutron energies, while the plates with thickness 2b are made up of the pure retarder. It is
assumed that the presence of resonance-absorbent nuclei does not change the retarding properties of the medium.
Then if(r, x) represents the spectrum of neutrons in the retarder with a constant lifetime T? [2] (T? corresponds
to absorption by the retarder nuclei only), which satisfies the equation
d21Y (x) , dir
dx2 I , (x))?
(x)
Q (x) = 0.
Here Xs is the mean free path relative to scattering, Q (x) =A? (x--x0) is the neutron source density. Assuming
that k 8.
As an example, the stability region is determined for a ring cyclotron having the following parameters:
N = 30, n ,41 10. In this case, there exists a sufficiently wide region of stability at 1.21 < vilvs < 1.33. De-
pending on the choice of operating point, the gain of the machine will vary from 7 to 10. Note that the mean
orbital radius is determined by the finite energy of the accelerator.
*The index n representing the growth of the magnetic field along the vertical is obviously dependent on the
choice of the origin z of the frame of reference.
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In examining the-spatial orbits, we might hope to obtain (by analogy with the "symmetric" ring synchro-
cyclotron (4, 5)) a "symmetric" ring cyclotron in which acceleration of electrons in opposite directions and
head-on collisions are possible simultaneously.
Because of the radiation emitted by the circulating electrons, there exists in the ring cyclotron, for an
even phase distribution of particles, a set of critical orbits determined by the relation
V :15 n cos cri,10114; V =
o
(8)
where the energy increment per revolution eVo cos vo is measured in electron volts, and the boundary values (Po
are determined by the tolerances of the magnetic field. By bunching the electrons around one particular phase
(e.g., utilizing a preliminary phase bunching operation or introducing the phase dependence of the frequency
of revolution), it is possible to effect a significant reduction in the energy spread of the accelerated beam, and
to stack up the particles in a practical manner.
The authors express their gratitude to A. A. Kolomenskii for his kind discussion of the experimental work.
LITERATURE CITED
1. T. Ohkawa, Bull. Am. Phys. Soc. 30, 7, 19 (1955).
2. J. Teichman, Czechosl. J. Phys. 9. 262 (1959).
3., B. N. Rodimov, Atomnaya Energ. 6, 2, 200 (1959)."
4. A. A. Kolomenskii, Zhur. Eksp. i ?Teoret. Fiz. 33, 298 (1957).
5. T. Ohkawa, Rev. Sci. Inst. 29, 108 (1958).
'Original Russian pagination. ?See C. B. translation.
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SOME PROPERTIES OF ACCELERATOR ORBITS
WHERE SIMILITUDE IS OBSERVED
A. A. Kolomenskii and A. N. Lebedev
Translated from Atomnaya Energiya, Vol. 8, No. 6, pp. 553-555,
June, 1960
Original article submitted January 5, 1960
The condition of dynamic similitude is the term used to describe the requirement that the characteristic
numbers of betatron oscillations be independent of the energy of the particles being accelerated. The need for
dynamic similitude is dictated by the fact that the frequencies of the betatron oscillations must not take on the
?dangerous resonance values during an acceleration cycle. It is especially important that the similitude condition
be satisfied in fixed-field alternating-gradient accelerators, where the orbit parameters undergo considerable
variations during acceleration.
As indicated in an earlier paper, the condition of dynamic similitude for planeorbits is the special config-
uration of the magnetic field
II:f(0)r1?,
(1)
where Hz is the component of the field at right angles to the orbit plane; f(0) is an arbitrary periodic function;
r and 0 are the cylindrical coordinates of the point on the orbit; no = const. In this note, we consider some of
the general properties of the motion of particles in such systems, on which research and development work has
been much intensified of late, in connection with the problem of particle storage and beam collisions.
Let there correspond to some particle momentum p a plane closed orbit r(a) (where a is a length of arc).
The linearized equations for deviations along the normal (p) and along the binormal (z) from this orbit exhibit,
as we know, the familiar form
e" -17 K2R2 (1? n) =0;1
z" K2R2nz = 0,
(2)
where the field index n is measured along the normal to the orbit; K is the orbit curvature, and differentiation
is performed with respect to a generalized azimuth 0. = a/R (R being the mean radius of the orbit). The radius
vector of another orbit, the latter in this case corresponding to a momentum p + p, may be represented as
(3)
where n is a unit vector normal to the orbit; 0 (.9.) is a periodic solution of the equation
Ar K2R2 (1-- n) (4)
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(differentiation with respect to .9.).
The invariance condition for equations (2) is obviously
( OK R ? =0; ( ?O. (5)
\ op }fr=const P 0=const
Taking in account the fact that the length of an arc of the curve r1( 0) is
a
Klpda ,
P .
0
-(6)
differentiating Eq. (3), and using the Serret-Frenet formulas, the first condition in Eq. (5) may be readily reduced
to the form
0
K (a? 1) ? nK2t ? K' IS K ?a01=0,
o
2n
(7)
where a=-1 Kip rill is the factor of instantaneous orbits. In this formula, n may be conveniently expressed
2a
in terms of no with the aid of the identity
r OJfzr (011, dQ da
no = =
ar Hz ae aa dr )
= ? K (rn)
(8)
(where 7 is the tangent vector).
For this orbit configuration, Eqs. (4) and (7) constitute an inhomogeneous system with periodic coefficients,
and linear in unknowns no and 0. As we see from the general theorems in [2], such a system may have only one
periodic solution in the nonresonant case, a solution having the same period as the coefficients. Direct verifi-
cation readily convinces the reader that this solution is
no =Const;
1
1?no
(9)
in which case the solution is not bound to any concrete shape of orbit, i.e., it remains valid for an arbitrary
azimuthal dependence of the magnetic field.
The condition of dynamic similitude makes itself felt not only on the orbit geometry, but also exerts a
material effect on the dynamics of the particles undergoing acceleration. In a number of cases, similar orbits
reveal an analogy with circular orbits. This finds particularly clear reflection in the problem concerning radia-
tion effects in fixed-field accelerators.
As has been shown in several previous communications on the subject, the radiation reaction leads to po-
sitive or negative damping of the betatron and synchrotron oscillations with a damping constant [3]:
1'
= [i ?(1 ?201014 (radial oscillationsi
= (vertical oscillations)
s=72_r 12+ (,t 2n) Klm (soyscnfpraottiroonsn)
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where a bar over a symbol denotes averaging over a periodicity element, and r is the dimensionless intensity of
radiation. The sum of the damping constants is always equal to 2T", independently of the form of the magnetic
system.
If the orbit consists of arcs of constant curvature, as in conventional weak-focusing and strong-focusing
synchrotrons, then we obtain the familiar formulas
a I
2 r;
4?a-
2'
?
9
from which, in particular, we learn that, in a strong-focusing magnet (a 1), radial betatron oscillations be-
come unstable [3]. Applied to accelerators with similar orbits of alternating curvature, this conclusion is not
quite so obvious, but retains its validity, as demonstrated below. The basic importance of this assertion must
be stressed, since it signifies that the damping systems developed for conventional synchrotrons cannot be made
effective under the similitude conditions [3, 4]. The proof reduces to substitution of Eq. (9) for the function 11.)
in Eq. (10). We then have:
Bearing in mind that I' ^, K2, we have
-2 [1-1-Ka(rn)---2a (n?1- (rv)
Etc
2 0-0= -- (1-cr
RK = (rn).
Substitution of Eq. (13) into (12) yields
on F
1--no 2
(12)
(13)
(14)
The expressions for the remaining damping constants also agree with Eq. (11). It is important to note that these
expressions are in no way connected to the azimuthal dependence of the magnetic field, which may be arbitrary,
and are derived solely from the similitude condition.
Now, when similitude is preserved it is impossible to set up a damping system operating by the coupling
between radial oscillations and synchrotron oscillations, so that a possible technique for suppressing radiation
instability is still available in the artificial coupling of those oscillations with the vertical oscillations. Using
the formulas derived in a previous paper [3], we find the damping constants of the bound oscillations in a similar
system;
r.a = ? (2 a)
z
02+A2j1/2} (15) '
where 6 is the distance of the operating point from the difference resonance of the coupling; A is a parameter
characterizing the coupling between the two modes of oscillation. Numerical estimates show that, in order to
bring about efficient damping in an electron ring synchrocyclotron of the type described in [5], a magnetic field
comprising ^, 10% of the guide field must be introduced.
LITERATURE CITED
1. A. A. Kolomenskii, Atomnaya Energ. 3, 12, 492 (1957).?
2. J. Sansone, Ordinary Differential Equations [Russian translations] (IL; Moscow, 1954).
3. A. A. Kolomenskii and A. N. Lebedev, Atornnaya Energ. 5, 5, 554 (1958).
4. Yu. F. Orlov, E. K. Tarasov, and S. Z. Kheifets, Pribory i Tekh. Eksp., 1, 17 (1959).?
5. V. Kanunnikov, et al., Symposium CERN (1959) pp. 89-99.
*Original Russian pagination. See C. B. translation.
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MEASUREMENT OF THE RADIATIVE-CAPTURE
y -EMISSION SPECTRA OF NEUTRONS IN SOME ROCKS
A. A. Fedorov, M. M. Sokolov, and A. P. Ochkur
Translated from Atomnaya Energiya, Vol. 8, No. 6, pp. 555-556.
June, 1960
Original article submitted December 12, 1958
Gamma emission with a spectrum consisting of y -lines characteristic Of each nuclear species is the result
when the nuclei of chemical elements are bombarded by thermal neutrons [1, 2]. Analysis of the spectra of y -
lines accompanying neutron capture sometimes allows us to infer the presence of certain elements in substances
of complex chemical makeup [3]. In 1956, the authors completed experiments, in a bore hole, confirming the
possibility of utilizing this method for detecting certain chemical elements in various rock species.
S Mev
60
70
80
Pulse height, v
Fig. 1. y -Emission spectra, for y -radiation in response to neutron irradiation of cherts
(1) and diorites (2).
The neutron source was a Po +Be preparation with a yield of 2 ? 106 neutrons/sec. Gamma emission was
recorded by means of a scintillation spectrometer. The spectrometer resolution for the Cs137 y -line (0.66 Mev)
was 12%. A NaI crystal, FEU-19M phototube multiplier, and an amplifier were placed in a logging toolsuch as is
used in borehole logging. To minimize y -radiation background due to neutron capture in the material of the
logging tool the portion of the latter in which the neutron source, crystal, and phototube multiplier were housed
was made of textolite. The crystal and photomultiplier were specially shielded by a layer of boron carbide. A
bismuth screen was inserted between the source and crystal (spacing of 10 cm). The source was surrounded by
paraffin to improve conditions favorable to slowing down the neutrons.
The y -emission spectra were measured in the energy region from 4.5 Mev and higher, since the spectrum
is distorted at lower energies on account of the scattering of the y -radiation of the neutron source proper by the
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surrounding rocks and drilling mud. Fig. 1 shows the pulse-height distribution of pulses corresponding to the y -
emission spectra, as a result of neutron irradiation of two rock species differing in composition: cherts, which
consist basically of oxygen and silicon (curve 1), and diorites, which consist for the most part of oxygen, silicon,
sodium, calcium, aluminum, and iron (curve 2). Transverse cross sections of thermal-neutron capture (a) and
the energy of the principal y -lines, (Ey > 4.5 Mev) for the elements mentioned are given in tabular form.
0
Na
Al
SI
Ca
Fc
cr , barns
2-10-4
0,50
0,22
0,13
0,43
2,43
E' Mev
v
?
6,41
7,70
4,95
6,42 7,64 ?
?1C
X X
Ia
1 2 3
h,m
190
200
210
4.
5 64 6 8 1012!
Since oxygen absorbs virtually no neutrons, the
y -emission spectrum of the cherts in the energy region
around 5 Mev shows a peak corresponding to the silicon
? y -line (4.95 Mev).* The y -emission spectrum of the
diorites is observed to peak at some other points, corres-
ponding to the y -lines of sodium and calcium (6.4
Mev), aluminum and iron (7.6 Mev).
As we see from Fig. 1, the largest difference
discernible in the y -emission spectra of the cherts and
the diorites is observed at energy ^, 7.6 Mev, corres-
ponding to the y -lines of aluminum and iron. By con-
tinuously recording y-emission intensity from 7.6 Mev
energy on, these elements can be readily identified in
a bore-hole section.
Fig. 2. Logging charts: I) diorites; II) cherts; 1) y -y--
logging with Co" source 2, 3) neutron-y -logging using
Po + Be source (integral count taken at discrimination
levels 0.1 and 4.5 Mev, respectively); 4) neutron-y -
logging with Po + Be source.
In Fig. 2 we have the logging charts obtained from
the same sample. Curve 2 corresponds to an integral ,
count at a discrimination level of 0.1 Mev and repeats
the logging density trace obtained with the Co" source
(curve 1) on which we discern cherts of low density.
On the diagram corresponding to the integral count at
discrimination level 4.5 Mev (curve 3), the cherts are
conspicuous with their shallow minimum due to the
absence of aluminum and iron in that species. A par-
ticularly sharp difference in the makeup of elements in
cherts and diorites stands out in the diagram correspond-
ing to a differential count over the energy range 7.3-9
Mev (curve 4).
The results so obtained provide confirmation of
the possibility of determinations of individual chemical
elements in rocks by the y -emission resulting from ra-
diative capture of neutrons.
LITERATURE CITED
1. G. Bartholow and B. Kinsey, Canad. J. Phys. 31, 1025 (1953).
2. L. V. Groshev, et al., Atomnaya Energ. 3, 9, 187 (1957),*
3. P. Baker, J. Petrol. Technol. 9, 3, 97 (19-57).
? This is due to the high silicon content in cherts, which acts to offset the relatively small capture cross section.
? ?Original Russian pagination. See C. B. translation.
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SLOWING DOWN OF NEUTRONS IN STEEL-WATER MIXTURES
L. A. Geraseva and V. V. Vavilov
Translated from Atomnaya Energiya, Vol. 8, No. 6, pp. 556-557,
June, 1960
Original article submitted January 7, 1960
Measurements of the second spatial moment of the slowing-down density of fission neutrons in water and
steel-water mixtures was carried out in a steel tank measuring 74 cm by 74 cm by 100 cm, flooded with water
and containing slabs of $t-3 steel 71.5 cm by 71.5 cm by 0.3 cm.
To forestall corrosive attack on the slabs and tank walls, these components were finished with a bakelite
resin. The slabs were placed in the tank at right angles to the direction of measurement of the slowing-down
density distribution, and were fastened in the required position by means of duralumin and plexiglas racks placed
on the bottom and walls of the tanks.
(p -
Measurements were performed for three concentrations by specific volume of iron and water
iron volume
), equal to 0.14, 0.26, and 0.43. A control experiment was staged to
iron volume plus water volume
measure the neutron age in water. The fission-neutron source used was a converter which converted thermal
neutrons from the pile into neutrons of the U235 fission spectrum, and was made of uranyl uranate 75% enriched
with U235. The spatial distribution of slowing-down neutrons was measured by means of cadmium-plated indium
foils (mean thickness 40 mg/cm).
The relatively weak flux of thermal neutrons, and consequently of fast neutrons as well, emerging from the
converter, was not sufficient to carry out measurements at distances greater than 56 cm from the source, as
required to determine the age of the neutrons. It is a known fact that the slowing-down density at large distances
from the source falls off in obedience to the law ",(ke-rA)/r2, where X is the relaxation length. This circum-
stance was used to advantage in extrapolating the distributions to infinity.
To compute the neutron age r, we used the familiar formula
C AO dr
1. 6
Ar2 dr
co
The values of y Ara dr and y Ar2.dr were determined empirically, and the values of c AO dr and
4) 6
CO
Ar2 dr
were obtained analytically by extrapolating in accord with the law A (ke-1A)/r2. The value of
k was arrived at by choosing a function ke-rA such that the extrapolated portion could be "tacked on* to the ex-
perimental portion; the value given to X was taken from the last points of the experimental distribution.
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The values obtained were;
TH20=30.2+1,5 cm2
tFe 1-IT20=31M + 2,7 cm2 (e =- 14);
= 39,7 ? 2,0 cm2 (Q=0,26);
tFe+1120= 50,4 ? 2,5 cm2 (Q=0,43).
The neutron age in a mixture of several substances,each of which has a known slowing-down length is
calculated from the formula
Q7 2 QiQi.,
, X
Ti V titi,
2
X it
2
.3/11
2 Ti
2 )
(1)
where pi is the specific concentration by volume of the i-th substance present in the mixture.
On the diagram, the curve corresponds to predicted values of T up to energy 1.46 ev for various concentra-
tions of the mixture of iron and water laminations, computed by Eq. (1). In the calculations, the following values
of r were used; for iron 7 = 743 cm2 (calculated)for water T = 30.5 cm2 (empirical). As we see from the
accompanying diagram, the values arrived at empirically for the age of neutrons in steel-water laminations
show good fit with predicted values.
T, Cm2
50
40
30
200
01 0,2 0,3 0/t f
Dependence of neutron age in steel-water lamina-
tions on concentration. ) plotted by
-r
computation; 0) empirically derived data points.
In conclusion, the authors would like to avail them-
selves of this opportunity to express their acknowledgment
to B. G. Dubovskii, Yu. A. Sergeev for formulating the
problem and for their kind participation in the discussion of
the results, and to our co-workers V. K. Labuzov, Yu. S.
Ziryukin, M. M. Kuzichkina, A. T. Anfilatov, who took
part in the measurements.
LITERATURE CITED
1. L. Roberts, et al., J. Appl. Phys. 26, 8, 1018 (1955).
2. A. D. Galanin, Theory of Thermal Nuclear Reactors.
Supplements No. 2-3 to Atomnaya Energiya [in
Russian] (Moscow, Atomizdat, 1957) p. 42.
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DETERMINATION OF DEGREES OF EQUILIBRIUM
OF SHORT-LIVED RADON DAUGHTERS IN AIR
L. S. Ruzer
Translated from Atomnaya Energiya, Vol. 8, No. 6, IT. 557-559,
June, 1960
Original article submitted January 9, 1960
It was demonstrated earlier [1-3] that the overwhelming portion of the absorbed dose in the case of radon
inhalation is due to the short-lived daughter products of radon decay.
The degree of equilibrium of the short-lived daughter products of radon decay ? was determined by the
usual method of filtering the air with subsequent measurement of the filters for a-radiation [4]. The amount of
radon present was determined by sampling the air in an ionization chamber (known in this case as an emanation
chamber) and measuring the amount of ionization current.
Below, we suggest a method for determining the concentration of short-lived daughter products of radon
decay which is based solely on measurements of the amount of ionization current flowing in the emanation
chamber.
We shall deal only with the ionization current due to a-emitting isotopes (RaA and RaC'), on the basis that
the energy of the a- particles in such a case is much higher than the energy of the 8 -radiation.
We use the term q to denote the radon concentration, in curies/liter, and ri 7IR, and ric to denote the
degrees of equilibrium of RaA, RaB, and RaC, respectively.
The activity due to RaA alone, i.e., the RaA present in a unit volume of the chamber, is qnAe-Xlit (XA,
XB, and Xc being the decay constants for RaA, RaB, and RaC, respectively), and the RaA activity for RaA formed
from radon in a unit volume of the chamber is -q (1-e-XAt). The total activity due to RaA is
A
-2 ki eit
[1]Ae +1?e - 1,
(1)
where v is the chamber volume.
The activity due to RaC' in the chamber at any instant is Ac, (t) = Ac (t), and will be a sum of the activ-
ities due to RaC (4), RaB (A), RaA , (AIcII.) and Rn (AIT). Using the solution for a chain of radioactive elements
[5], we obtain
Arc (t)= gice-Xci ;
Ac ii (I) =__ grin XC
-
C B
The "degree of equilibrium" of a given radon daughter refers to the ratio of the amount of daughter product to
the equilibrium ratio in air (translator's note).
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API (t)= qrlAknkn
e-XAt
(B??LA) kf.)
e-XBt e-4Ct
?
+
(kA?k13) QT.-2LO
,4Icv (t)-=q[1?
(X.A.-20(kB?kc)
kBkCe-kiLl
(B?kA)('C?XA)
Bt13e-2'Ct
kAXCe
(kA-4)(Xc? X13) ()LA? XC) Q13-20_
Turning now to the amount of ionization current due to the entire volume of the chamber,
I=k (EA
I + I ,
RRn A ERn C
we must bear in mind that the amounts of energies of ct-particles from RaA and RaC' are different from ERn,
in consequence of which the ratios EA/ERn and Ect/ERn appear in the' expressions for the ionization current
(and k in the above formula is a conversion factor from activity to current). Finally, we have
I= kg? URn (1)+11A1A (t)-1-T1B/B (1)+1ICIC =
=kqvF (t).
The graphs plotted for functions f Rn (t), f A (t), and f (t)
fR.fg,fsufc
1,0
0,8
0,6
0,4
0,2
20 40 60 80 100
200
300
t, min
Fig. 1. Graphs of functions f Rn (t), f A (t), f B(t),
and fc (t) (curves 1-4, respectively).
Fa) 0:1:11
2,5
1,5
1,0
0,8
0,0
0,4
42
10, 08
(10
01"----
I
Ariii
Pr
r
(8:0:0)
III
III
?
20 40 50 80 100
200
300
t, min
Fig. 2. Function F (t) at different values of ?IA,
71B, C. Parentheses enclose the ratios of the
degrees of equilibrium, as n A: fl B:
(4)
(5)
(6)
(7)
give some idea of the contributions of radon, RaA, RaB, and RaC, respectively, to the amount of ionization cur-
rent, and may be seen in Fig. 1.
The ratios nA , R. and 71C may be found from Eq. (7), if three values of the ionization current taken at
different instants are used. The values of f Rn (t), 1A (t), f (t), and f (t) for any instant of time t are read
off from the graphs in Fig. 1.
Fig. 2 shows a graph of the function F(t), which gives us some idea of the nature of the increase in ioniza-
tion current flowing in the chamber at different values of n A, ng, and nc . As we see from Fig. 2, the ionization
current values in the chamber corresponding to different ratios of the degrees of equilibrium /IA, nil, and nc
differ markedly from each other in the first 60-80 min following sampling of the air.
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The low curve corresponds to the increase in current attributable to radon without decay products. In the
case where the decay products are in equilibrium with radon (nA = 7113 = nc = 1), F(t) = const.
The method described above for determining the concentration of short-lived daughter products of radon
decay is quite simple and requires no equipment other than conventional electrometers used for emanations
(SG-IM). This method is suitable for measuring /IA, nB, and rIc in rooms of fairly small volume.
A similar method may be employed for determining the concenirations of daughter products of other ema-
nations.
LITERATURE CITED
1. S. Cohn, R. Skow, and J. Gong. Arch. of Indust. Hyg. and Occupat. Med. 7, 6, 508 (1953).
2. V. Hultqvist, Ionizing radiation from natural sources [Russian translation] (IL, Moscow, 1959).
3. L.S. Ruzer, Atomnaya Energ. 4, 2, 144 (1958).*
4. E. Tsivoglou, H. Ayer, and D. Holaday, Nucleonics 11, 9, 40 (1953).
5. G. Friedlander and J. Kennedy, Nuclear and RadiocheMistry (J. Wiley, New York, 1955).
?Original Russian pagination. See C. B. translation.
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LUMINESCENT DOSIMETERS BASED ON THE CaSO4. Mn
PHOSPHOR FOR THE DETECTION OF GAMMAS, BETAS,
AND NEUTRONS
V. A. Arkhangel'skaya, B. I. Vainberg, V. M. Kodyukov,
and T. K. Razumova
Translated from Atomnaya Energiya, Vol. 8, No. 6, pp. 559-561,
June, 1960
Original article submitted September 11, 1959
In 1951, Watanabe [1], in an investigation of the thermoluminescent behavior of CaSO4.Mn, noted the
ability of this phosphor to store up energy when subjected to y -emission from radium. The possibility of using
CaSO4.Mn in the dosimetry of ionizing radiations was discussed in several subsequent papers [2-4], but detailed
quantitative data on the dosimetric properties of this phosphor are not available from any source. The present
contribution seeks to fill that gap.
/relative units
1,0
0.8
D.4
0,
0440
480 520
560 A ,nw
Fig. 1. Thermoluminescence spectrum
of the phosphor CaSO4.Mn.
The energy stored by the CaSO4.Mn phosphor during the irradia-
tion process (and known as the light sum) may be obtained in the
form of visible radiation in response to heating of the phosphor. The
peak in the thermoluminescence spectrum of the phosphor falls in the
region of 500 mp (Fig. 1). The dependence of the degree of glow
intensity on the temperature to which the excited phosphor is heated
is shown in Fig. 2. The lone peak on the thermoluminescent glow
curve in the temperature region higher than room temperature is in-
dependent of the mode of excitation. It might thus be inferred that
the thermal properties of dosimeters using the CaSO4.Mn phosphor
will be identical for different modes of excitation.
The response of CaSO4.Mn to x-radiation and soft y -radiation
is appreciably higher than its response to harder y -radiation (curve 1
in Fig. 3). Using a lead filter of predetermined thickness, the do-
simeter response was made to equal the response of the phosphor over
a rather broad range (0.1-2.6 Mev) of energies (curve 2 in Fig. 3).
The response of the phosphor to y -radiation in the energy region of
interest is so high that dosimeters having a luminescent surface area
of 2 cm2 are capable of measuring doses starting as low as 0.001 r,
with unsophisticated photoelectric equipment to aid in the measure-
ments. The upper range of measurable doses D is bounded by a break
in the linearity of the relation between the value of the light sum L and the irradiation dose. As we learn from
Fig. 4a, this boundary lies in the region of irradiation doses of the order of several hundred roentgens; the sublinearity
of the L(D) excitation curve stays within 3010 even at D a-. 1000 r, a fact which may be taken into account in the
measurements.
By utilizing the same sensing equipment as in y -dosimetry, and the same luminescent surface area of the
dosimeters, we were able to record doses of 8-radiation (from Sr25Y20) ranging from 1 ? 105 to 1 ? 108 particles/
/cm2. No break in the linearity of the L(D) relationship was observed, practically speaking (Fig. 4b).
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1,relative units
100
00
60
20
40
60 eo 100
120 140 tc; C
Fig. 2. Thermoluminescent glow curve of the phosphor CaSO4.Mn in
response to 13 -particles (0), UV radiation (x), x-rays (A), and gamma
radiation (0).
According to preliminary data, the response of CaSO4.Mn in the region of x-radiation from energy ^, 15
key (source: an x-ray tube, 15 key, copper anticathode) makes it possible to measure irradiation doses at the
microroentgen level. The L(D) plot for this case is seen in Fig. 4c. The plot of CaSO4.Mn response vs hardness
of radiation in the region of soft x-radiation and intermediate-range radiation, as well as the possibility of com-
pensating the variation with hardness with the aid of filters, will be the subject of a subsequent paper.
Despite the fact that the peak on the thermoluminescence glow curve of the phosphor is found at 90-100?C,
prolonged shelf storage of the irradiated phosphor even at room temperature results in partial loss of the light
sum stored by the phosphor (decay curve, curve 1 in Fig. 5). In the case of the CaSO4.Mn phosphor, the degra-
dation of the light sum with time depends
neither on the magnitude nor on the dose rate
of the irradiation received, in contrast to the
40
light sum of the SrSEu.Sm phosphor, also used
L, relative units
t
Gel
I
/i2203
CS137
cam ea
The "
2 nuNd
_...,--
in individual dosimetry [5]. With increase in
30
temperature, the decay in light sum with time
is speeded up (curves 2 and 3 in Fig. 5). If the
operating temperature of the dosimeter does not
20 exceed 25?C and readings are taken daily, light-
sum losses do not exceed 25% for a working day.
However, with the increase in temperature and
10
0 0,4 0,8
1,2 1,6
2,0
2,4
Ey Mev
Fig. 3. Response of CaSO4.Mn phosphor (1) and dosimeter (2)
longer storage time for the dosimeter, losses
may reach 50% and higher. As investigations
have shown, the initial rapid decay L (t) pro-
ceeds via luminescent emission of the light
sum stored up at shallow trapping levels of the
CaSO4.Mn phosphor. By using a part of the
to y -emission over an energy range. light sum accumulated at deeper electron trap-
ping levels for dose measurements, it would ob-
viously be possible to improve the thermal characteristic of the dosimeter. The light sum to be measured then
consists, as shown by calculations, of not less than 20% of the total (if the excitation temperature and the tem-
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L, relative units
a
L, relative units
L, relative units
80 160 ll,iose
Fig. 4. Dependence of the magnitude of the stored light sum L on the dose of ionizing radiation for the
cases of excitation by y -radiation (a), x-radiation (b), and beta particles (c).
perature at which the dosimeters are left unexposed is not above 35?C) and may be retained for several days in
the dosimeter. In practice, the thermally unstable portion of the light sum may be removed by pre-heating
the dosimeter. This heating may constitute part of the over-all process of heating the dosimeter in measuring
dosage.
The technology involved in fabrication of the phosphor is quite simple and yields readily reproducible
results. The original materials are inexpensive and require no special purification, since impurities of heavy
metals have no effect, even in relatively high concentration, on the response of the phosphor to ionizing radia-
tions.
Repeated bombardments by radiation and repeated heating of the phosphor fail to produce any appreciable
alteration of its properties. A dosimeter based on the CaSO4.Mn phosphor may function without being recalib-
rated for several years.
L, relative units
1,04
0,0
40 15-6 120 160 200 240 t,hrs
Fig. 5. Decay curve of light sum L with dosimeter left unex-
posed, as a function of temperature, amount and dose rate of
irradiation; 1) 22?C; 2) 37?C; 3) 57?C; 20-r dose, dose rate
in r/hr; 49) 6.7; X) 1040; 0) 154.
In contrast to the SrSEu.Sm used in indi-
vidual luminescent dosimetry [5], the CaSO4.Mn
phosphor is stable to moisture attack, and does
not require any special leakproof container.
Total lack of response to visible and ultraviolet
radiation right up to wavelength X = 1500 A
may be counted as one of the additional advan-
tages of this type of luminescent dosimeter.
High-density emission in the region 2600-1800
A (not less than 1 mw/cm2) with prolonged
irradiation of the phosphor leads to partial radia-
tionless loss in light sum. However, neither
direct radiation from the sun at the level pre-
vailing at the earth's surface nor, a fortiori, the
light from an incandescent lamp, have any
effect on the light sum accumulated.
The possibility of growing luminescent
CaSO4.Mn single crystals of modest size has
been reported [4] in the literature. Such single
crystals without supplementary crystal holders
may find application in beta dosimetry. The
use of single crystals in gamma dosimetry has
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obviously made it possible to improve the sensitivity of the method by increasing the thickness of the layer of
phosphor transparent to luminescent emission.*
The CaSO4.Mn phosphor may also be employed to record thermal and fast neutrons. In the first case, the
lead filter is replaced by a filter made of a thin cadmium wafer. To record fast neutrons, polymethyl methac-
rylate is introduced into the composition of the phosphor after the latter has been fabricated.
The totality of all the properties of the CaSO4.Mn phosphor discussed above justify us in viewing it as one
of the most promising developments in individual luminescent dosimetry.
LITERATURE CITED
1. K. Watanabe, Phys. Rev. 83, 785 (1951).
2. U. Mayer, Naturwissenschaften 43, 79 (1956).
3. B. M. Nosenko, L.S. Revzin, and V. Ya. Yaskolko, Optika i Spektroskopiya 3, 4, 345 (1957).
4. V. A. Arkhangel'skaya, B. I. Vainberg, and T. K. Razumova, Optika i Spektroskopiya 4, 5, 681 (1958).
5. V. V. Antonov-Romanovskii, Session of the Academy of Sciences of the USSR on Peaceful Uses of Atomic
Energy. (Session of the Division of Physical and Mathematical Sciences) [in Russian] (Izd. AN SSSR,
Moscow, 1955) p. 342.
? 'Measurements have shown that, for the usual grain size of CaSO4.Mn powder ranging from 1-5?, an increase of over
0.5 mm in the thickness of the layer of powder would not be effective.
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SCIENCE AND ENGINEERING NEWS
LETTER FROM A READER
L. A. Artsimovich
(On the article "Entropy Trapping of a Plasma by a Reversal
of the Magnetic Bottle Configuration")
Translated from Atomnaya Energiya, Vol. 8, No. 6, p. 562,
June, 1960
Original article submitted April 15, 1960
In the third issue of Atomnaya Energiya for the present year, there appeared an abstract in the section
"Science and Engineering News" under the signature of G. B., dealing with several articles by J. Tuck. It is
stated therein that "J. Tuck . . . proposed a new method for injecting a plasma into a magnetic trap with an in-
version of the magnetic bottle configuration . . ." In writing this, the author of the abstract was in error, since
a technique of injection of plasma globs into magnetic traps having the described field configuration was
studied much earlier by S. Yu. Luktya.nov and I. M. Podgornyi [1].
In this connection, I should like to note that this method had already been developed experimentally by
our group at the Institute of Atomic Energy during the past few years. Eight months ago, even. before the arrival
of J. Tuck's first published article on the subject, the article by S. Yu. Luktyanov and I. M. Podgornyi dealt with
the problem of the possibility of trapping a plasma in a trap having a magnetic field of the configuration referred
to, which the author of the abstract terms an anti-bottle configuration. The fact that theoretical research and
experimental investigations have been underway in the Soviet Union along these trends has received mention
earlier yet, in particularly in the papers presented by Soviet physicists at the second Geneva conference on the
peaceful uses of atomic energy [2, 3].
It is difficult for me to judge whether or not J. Tuck was aware of the papers of the Soviet reseachers in
this field, and in particular of this paper by S. Yu. Luk'yanov and I. M. Podgornyi, of which J. Tuck's work is a
further development.
LITERATURE CITED
1. S. Yu. Luktyanov and I. M. Podgornyi, Zhur. Eksp. i Teoret. Fiz. 37, 1(7)27 (1959).
2. L. A. Artsimovich, Proceedings of the Second International Conference on the Peaceful Uses of Atomic
Energy (Geneva, 1958) Paper 2298.
3. 0. B. Firsov, Plasma physics and the problem of controlled thermonuclear reactions [in Russian] (lzd.
AN SSSR, 1958) Vol. ifi, p. 327.
4. J. Tuck; Phys. Rev. Letters 3, 7, 313 (1959).
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GENERATION OF AE--HYPERON BY NEGATIVE PIONS
WITH A MOMENTUM OF 8.3 Bev/c
M. I. Solov'ev
Translated from Atomnaya Energiya, Vol. 8, No. 6, pp. 562-563,
June, 1960
A team of physicists working at the High Energies Laboratory of the Joint Institute for Nuclear Research
(Dubna), integrated by V. I. Veksler, N. M. Viryasov, E. N. Kladnitska, A. A. Kuznetsov, A. V. Nikitin, M. I.
Solev'ev (USSR). Wang Hang-Chang, Wang Tsu-Ren, Ting Ta-Tsao (China), Nguyen Dien Thu (Vietnam),
A. Michula (Rumania), Kim Hi Un (Korea), Jiri Vran (Czechoslovakia), are conducting research with beams of
negative ir -mesons in a 24-liter propane bubble chamber. The bubble chamber is mounted inside a magnet
configuration, the magnet having a constant field of 13,700 strength. The chamber has been in operation for
over a year; during that time a large number of photographs have been amassed.
In an analysis of 40,000 photographs obtained in
the beam of negative 'Tr -mesons with momentum 8.3
? 0.6 Bev/c, one event of generation and decay of a
2' -hyperon was detected (see accompanying color
photo and explanatory diagram).
The Ir-meson primary (track 1) interacts with a
carbon nucleus to form four charged high-energy
particles (tracks 2, 5, 7, 16), two'K?-mesons (tracks
4, 5 and 14, 15), one low-energy particle (the short
13j\5 track 17), and the recoil nucleus.
The decay of particle 2 at point A to particle 3
and a neutral particle in the direction AB is in ex-
cellent agreement with the kinematics of E-hyperon
decay. Track 3 is the track of a ir+-meson. The
neutral particle at a distance of 7.7 mm from the point
where the decay event occurred forms, at point B, a
high-energy six-pronged star (tracks 8-13).
The energy contributed solely by the charged
particles (1483 ? 60 Mev) is much higher than the
kinetic energy of the neutral particle (940 ? 100 Mev).
The neutral particle was thereby determined to be an
antineutron. Under the assumption that a fraction of
Diagram of the principal tracks showing on the photog-
the energy was contributed by neutrons and 11?-mesons,
.
the energy in the star, after taking the binding energy
of nucleons within the nucleus into account, was determined to be higher than 2300 Mev. This energy is close
to the annihilation energy of an antineutron.
0 17
14 15
486
The most probable reaction at point B would be;
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o 0.7
0
0 t,
o
?
?? 0- ?
0{j
; 8
: ? .
0 ?
0oC
? ?
0
?
4
0
o
c ?
t
0
0
?
00
PomAeme pacnaA
ruiepo
0
00
Generation and Decay of a 'f-Hyperon 4527
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This implies that the primary 11---meson formed, at point 0, a hitherto unobserved particle, an antisigma
minus hyperon, decaying at point A in accord with the decay scheme E- IT + n. The presence of two neutral
K-mesons allows us to write the reaction for the generation of the E -hyperon as follows:
31--Fc?> +...
The negative particle (track 6) is determined to be a K- (by considerations of strangeness conservation).
For the lifetime of this E -hyperon, the following value was arrived at:
/Tr = (1,18 ? 0,07)?10-1? sec.
This detection of the first charged antihyperonis a fact of enormous scientific significance. Our concepts
of the microcosmos are further enriched thereby. One more stage has beenreached inthe knowledge of the nature
of elementary particles and their transformations.
The discovery of the new particle is a great success for the entire staff of the High Energies Laboratory
of the Institute.
HUNGARIAN EXHIBIT OF INSTRUMENTS FOR EXPERIMENTAL
NUCLEAR PHYSICS RESEARCH
Translated from Atomnaya Energiya, Vol. 8, No. 6, pp. 563-564,
June, 1960
In April 1960, the Hungarian exposition of instru-
ments for experimental nuclear physics, organized by
the Atomic Energy Commission of the Hungarian
Peoples Republic and the Hungarian METRIMPEX fo-
reign trade organization, was held at the Dubna Joint
Institute for Nuclear Research. The exposition was de-
dicated to the 15th anniversary of the liberation of
Hungary from the fascist aggressors.
In recent years, instrument design, and in par-
ticular the manufacture of instruments for nuclear
physics applications, has developed in rapid stride in
Hungary. It suffices to state that at the present time the
volume of instrument production is 13 times that in
1950, and will have increased by 28 times in 1965.
Manufacture of mass-production instruments for
experimental nuclear physics is being stepped up. To
improve the quality and lower costs of items manu-
factured in enterprises engaged in mass production of
electronic equipment, small research teams have been
set up.
Instrument design is the concern of such large-
scale scientific centers as the Central Scientific Re-
search Institute for Physics attached to the Hungarian
At the Hungarian exposition on instruments for experi- Academy of Sciences, and the research departments in
the universities and the various institutes.
mental nuclear physics. (Photo by M. Pyatkin).
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Instruments, diagram:, and wiring layouts were exhibited in the two halls occupied by the exposition (see
accompanying photo). On demonstration here were a thickness gauge, automatic sample changer, timing con-
trols, scalers, counting rate meters, a differential discriminator with resolving time of 5 ?sec, stabilized high-
voltage power supplies, general-purpose scintillation detectors, dosimeters, y -scintillation recording charts with
image size 297 by 420 mm, affording physicians the possibility of studying a tissue by enriching the tissue with
a radioactive isotope. Other counting and measuring instrumentation was also on display.
The electronic equipment is used in Hungary not only for scientific work in the field of the physics of the
atomic nucleus, but also finds applications on a broad scale in medicine, agriculture, ferrous metallurgy, the food
processing industry, and the plastics industry, and others.
The exhibition organized at Dubna showed the significant achievements of the Hungarian Peoples Republic
in the peaceful uses of the gains of science and technology.
RECENT DATA ON C14 CONCENTRATION IN THE ATMOSPHERE
Yu. V. Sivintsev
Translated from Atomnaya Energiya, Vol. 8, No. 6, pp. 573-575,
June, 1960
One of the scientific problems subjected to broadest discussion in recent years is the question of radiation
hazards due to testing of nuclear weapons [1-4]. It is a familiar fact that an enormous quantity of neutrons is
liberated in the explosion of any nuclear bomb, the neutrons going on to interact with nitrogen contained in the
air, forming the radioactive isotope of carbon CR:
CIA in the atmosphere has been steadily increasing since 1953.
7 N14 + 0 n1
6C14+1H1.
Being a pure 8 -emitter, C14 decays with a half life of about 5,600 years, thanks to which the quantity of
In contrast to atomic and uranium fission bombs, where the basic radiation hazard is associated with the
concentration in one area of long-lived radioactive fragments, Sr90 in the first instance, taken into the human
organism with the resultant formation of malignant neoplasms, the pure hydrogen (deuterium-tritium) fusion
bomb is dangerous to both contemporary and future generations because of the formation of mutant genes in
response to irradiation of the gonads by 8 -particles emitted by the isotope C14. Assuming a linear dose depen-
dence of the genetic effect, and also assuming that the spontaneous rate of mutation in humans is 10/0 due to
natural irradiation, 0. I. Leipunskii reached the conclusion [1] that the total number of genetic victims from the
explosion of a 10-megaton pure hydrogen bomb comes to 49,000 persons, as against 41,000 persons for an ex-
plosion of a conventional nuclear bomb. In line with these data, it is interesting to quote U. S. statistics on the
total equivalent of the nuclear test explosions carried out, which totals 174 megatons up to December 31, 1958,
according to [5].
This reference [5] also gives the results of measurements of Cu concentration in atmospheric carbon diox-
ide gas over the past 4 years (1956-1959). The procedure used in the study consisted in direct mass-spectro-
metric determinations of the CU/C12 ratio in samples of atmospheric carbon dioxide and plant substances which
had assimilated carbon not long prior to the sampling measurements (ring growths on trees of several years age).
For purposes of comparison, the results were normalized to the usual C13/01 ratio, which successfully eliminated
the differences associated with a possible dilution of the isotopes in the process of laboratory treatment of the
samples, or during photosynthesis (in the case of plant samples). The "background" content of Cm in wood was
also taken into account in the normalization. The final result of measurement of a sample ACIA thus showed
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the amount of radioactive carbon formed solely in the wake of nuclear weapons tests [6]. The results of the
measurements for the Northern hemisphere [5], expressed in tenths of a percent, are diagrammed. These data
indicate that the total annual increase in the long-lived radioactive CIA isotope present in atmospheric carbon
dioxide during the last three years was about 5%.
The data compiled increase the results of preceding investigations published in prior years. For example,
one report [7] stated an annual increase of 4.40 for the Northern hemisphere, while another gave figures of
3.2% for Central Europe and 2% for South Africa [8].
The authors of [5] indicate some possible reasons for the low results in preceding reports. One is related
to the dilution of atmospheric carbon dioxide by the enormous wastes of combustion products from fossil fuel
in large-scale industrial regions. Because the CIA of natural origin present in those substances had undergone
considerable decay, the air in large cities and their environs has a lower CIA content than in agricultural regions
(this is termed the Suess effect [9]). To take this into account, the authors of [5] plotted a curve (see diagram)
for samples collected over areas remote from industrial regions, above the Atlantic ocean (in the Northern
hemisphere), and in the Mediterranean. It is interesting to note that the data reported in [8] for plant samples
from industrial regions in West Germany showing a smaller annual increase in the amount of C14 display excellent
agreement with samples taken in New York, Rome and Kearny (state of Nevada, USA) when plotted on the graph.
180
120
s' 80
C
U "
4
0
-20
-40
-80
-
'
1.938 1953 1954 1955 1956 1957 1958 1959
Date of growth
CIA concentration in carbon dioxide (Northern hemisphere) [5, 7-9];
1) Atlantic (atm. CO2); 2) Mediterranean (atm. CO2); 3) Great Plains
(plants); 4) New York (plants); 5) Rome (plants); 6) Kearny, Nevada,
USA (plants).
A second factor responsible for the harvest of low results in preceding years is failure to trek the effect of car-
bon dioxide carried by the soil. Since plants growing on high-yield acreage may obtain a considerable quantity
of carbon dioxide from the soil, and since the latter probably formed through decay of organic substances bio-
synthesized prior to nuclear weapons testing, the difference between the CIA concentrations in soil and atmos-
pheric carbon dioxide widens its span rapidly with the passage of time. Also of interest is the fact that, in the
words of the authors of [5], the points used to plot the curve were obtained from sparsely sown areas and are
therefore free from effects of soil carbon dioxide.
Comparing their results with the data presented in [10] for the Southern hemisphere, the authors of [5]
stated that, although the overwhelming bulk of the weapons tests of recent years were carried out in the Northern
hemisphere, the C14 concentration in the atmosphere of different hemispheres differs by at most 36%. This last
result confirms the supposition advanced in [11] on the high rate of mixing of the atmosphere of the two hemis-
pheres (mixing period of about two years). In conclusion, the authors of [5], after discussing the various possible
mixing models for the stratosphere and troposphere of the two hemispheres, estimated the total quantity of bomb-
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produced radioactive C accumulated up to March 1958 (see Table). One fact of extreme interest is that the
most probable result which they arrived at coincides beautifully with the theoretical predictions put forth in [12].
As we see from the data presented in the table, the principal source of the uncertainty is the estimate of
the amount of Cm accumulated in the stratosphere. Reference [5] is successfully supplemented by the investi-
gations of stratosphere-borne C14 concentrations reported on in [13]. The authors, following a detailed descrip-
tion of the procedure used in sampling air at different altitudes (from 14 to 28.5 km) and measuring activities,
present their empirical data on the quantity of A-bomb C14 present for four altitude belts and five sampling
points, covering the period from 1958 to May 1959. The sampling points were spaced in a manner aimed at en-
Amount of Carbon-14 of A-bomb Origin (and H-Bomb
Origin) Accumulated by March 1958, in 1027 C14 atoms.
Tropo-
sphere
Earth
bio-
sphere
Ocean
Strato-
sphere
Total
Minimum
Most probable
value
Maximum
? 3.6
3.6
3.6
0.2
0.2
0.2
0.6
1.0
1.5
0.7
7.0
28.5
5.1
11.8
27.8
compassing a broad latitude belt (from 45?N to 25?S)
and thus to obtain the material needed for resolving
the question of the rate of mixing of the atmosphere
between the two hemispheres. On the basis of these
measurements, the authors asserted that the amount
of C14 in the stratosphere due to bomb testing, over
the period from 1955 to 1958, remained virtually
constant (to an accuracy of f 30%) and amounted to
an average of 7.2 ? 1027 atoms of C14 with fluctuations
from 5.6 ? 1027 to 8.6 ? 1027. However, Hagemann
and associates [13] feel that since 85% of the nuclear
explosions in recent years were conducted in the air,
the fraction of the C14 formation and build-up refer-
able to each megaton exploded should be scaled upwards. On this basis, they arrived at the conclusion that the
total quantity of C14 formed as of October 31, 1958 as a result of the detonation of nuclear weapons amounted to
25 ? 1027 atoms of Cm. If we assume this carbon to be evenly distributed over the entire atmosphere, then the
total tropospheric concentration of Cm will be found to increase 1.75 times over the background estimate. It
should be particularly stressed that the infiltration of atmospheric Cm into ocean waters, on which great hopes
had been laid in earlier years, from the standpoint of a relatively rapid cleansing of the atmosphere, is a rather
slow process, as seen in the light of the latest data. In particular, the authors of [5], basing their views on ex-
perimental determinations of the amount of Cm found in the surface waters of the Northern and Equatorial
Atlantic, reported that only 10/0 of the C14 generated by bomb tests by March 1958 had found its way into ocean
waters.
In conclusion, it should be mentioned that reference [14] gives the first three measurements of C14 concen-
tration in tissues of the human organism (lungs, blood, respiratory acid) as a result of which they found that the
C14 concentration in human tissues lags behind the C14 content of the atmosphere by as much as 1.1-1.8 years.
LITERATURE CITED
1. 0. I. Leipunskii, Atomnaya nerg. 3, 12, 530 (1957)."
2. 0. I. Leipunskii, Atomnaya nerg. 4, 1, 63 (1958).?
3. A. D. Sakharov, Atomnaya Energ. 4, 6, 576 (1958)?
4. J. Totter, M. Zelle, and H. Hollister. Science 128, 3337, 1490 (1958).
5. W. Broecker and A. Walton, Science 130, 3371, 309 (1959).
6. E. Anderson and W. Libby, Phys. Rev. 81, 64 (1951).
7. H. de Vries and H. Waterbolk, Science 128, 3338, 1550 (1958).
8. K. Munnich and J. Vogel, Naturwissenschaften 45, 14, 327 (1958).
9. H. Suess, Science 122, 3166, 415 (1955).
10. T. Raefter and G. Fergusson, Conference on the Peaceful Uses of Atomic Energy (Geneva, 1958) paper 2128.
11. G. Fergusson, Proc. Roy. Soc. A 243, 561 (1958).
12. Libby, Proc. Nat. Acad. Sci. V.S.A. 44, 816 (1958).
13. F. Hagemann, et al., Science 130, 3375, 542 (1959).
14. V. Broecker, A. Schulert and E. Olson. Science 130, 3371, 331 (1959).
?Original Russian pagination. See C. B. translation.
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APPLICATIONS OF ALPHA RADIATION FROM RADIOACTIVE
ISOTOPES FOR QUALITY CONTROL IN GRINDING OPERATIONS
V. V. Kondashevskii, A. N. Chertovskii,
V. S. Pogorelyi, and A. M. Gutkin
Translated from Atomnaya gnergiya, Vol. 8, No. 6, pp. 576-578,
June, 1960
A transducer operating on the basis of the number of particles reaching the counter as a function of the
transverse cross section seen by the rays of particles has been developed by the authors and has been studied un-
der laboratory and production conditions. The application for which the transducer is intended is to provide new
automatic controls for the grinding machine operation, with improved precision and reliability in performance.
The basic layout of the radiation transducer is shown in Fig. 1. The rod 2 of the transducer is held against
the measuring rod 1 of the gage, which comes in contact with the part being monitored. Resting on the rod 2, ?
and loaded by a spring 3, is an angle lever 4 connecting to a slide valve 5, which is inserted between the isotope
6 and a MST-17 type end-window particle counter 7. The radiation source is housed in a casing 8. Displace-
ment of rod 1 results in rotation of the cranked lever and slide Valve. The intensity of the a-radiation arriving
at the end-window counter depends on the extent to which the slit opening in the valve orifice 9 is open.
Fig. 1. Mechanical layout and electric circuitry of the radioisotope transducer.
The counter, whose function is to record the intensity of radiation from the isotope, is included in the
electric circuit with a data-indication dial 10 and relay 11.
Any change in the size of the machined part being monitored can be sensed by reference to the dial; at
the instant when the part is reduced to a predetermined size, the relay sends a command to automatically halt
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Fig. 2. Diagram of leverage design in radioisotope trans-
ducer for quality control of splined shafts: 1) connecting
levers (of instrument); 2) arc-shaped measuring adapter;
3) part being machined; 4) grinding wheel; 5) link ful-
crum; 6) axis of rotation; 7) faceplate for electrical
assembly; 8) data indicator; 9) transducer support bracket;
10) radioisotope transducer; 11) spring.
Fig. 3. Indicator-dial snap gauge withmulticommand
radioisotope transducer.
the machining process. Intensive research and
development work bore fruit in optimization of the
design parameters of the transducer components:
slit cross section (in orifice) of 0.4 by 15 mm; trans-
fer ratio of mechanical portion of transducer with-
in the range 4 : 1 to 10 : 1.
The emitter selected was a thorium isotope
for which the body of the transducer provided ade-
quate shielding for the operator from emitted alphas.
Only one paper* is known to the authors where an
alpha-emitting radioisotope transducer has been
mentioned.
The radiation transducer was mounted in a
triple-contact snap gauge6(see Fig. 3) to monitor
smooth cylindrical shafts during grinding, and was
tested under laboratory conditions, followed up by
later testing under production conditions on a grin-
der in the machine shop of the Omsk Sibzavod
plant. Working under a tolerance of 23 microns, the
spread in dimensioning of the machined shafts was
found to be 13 microns, whereas use of a conven-
tional gauge meets with a hard time in keeping with-
in tolerances in shaft work.
In work with a single-point gauge (Fig. 2)
coupled to the radioisotope transducer used to moni-
tor grinding of splined shafts to an 0. D. of 72 mm
and to a tolerance of 17 microns, the actual scatter
in the dimensions of the ground shafts was found to
be 17 microns. Without the aid of an instrument,
a skilled operator would encounter difficulty in ma-
chining the parts to within tolerance specifications.
The electrical circuitry for the radioisotope
transducer, described above, requires only one com-
mand in feeding work to the grinder, the command
to retract the wheel from working position. In some
cases, this one command turns out to be inadequate.
The radioisotope transducer 5 shown in Fig. 31s built
to give three commands to a machine tool:
1. Command to switch from rough feed to
finish feed, when the machined part has been di-
mensioned down to 30-60 microns within specifica-
tions. This command switches on an annunciator
light bulb 1.
2. Command to stop finish feed when the part is machined to 10-15 microns of specification. Light bulb
2 switches on at this point and the final finish machining is then initiated without feeding the wheel into the
work ("coasting").
3. A command to instantaneously retract the grinding wheel when the final dimensioning is completed.
Light bulb 3 flashes on at this point.
'M. B. Neiman. Use of radioactive isotopes in machine design [in Russian). Symposium "Automation of manu-
facturing processes in machine building (control applications)". Moscow, Academy of Sciences Press, 1955.
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The indicator snap gauge 4has two reaching scales: one coarse scale 7 and a precision scale 8. The scale
divisions of the precision scale run from 0.5 to 2 microns (set during adjustment), and the divisions of the coarse
scale run from 2.5 to 10 microns. The dial scales are switched on and off automatically. The presence of two
scales facilitates observation and monitoring of changes in part size.
Comparative tests of radioisotopes, inductive, pneumatic, and electric-contact transducers have shown
that the, precision of the radioactive transducers matches the levels of the best variable inductors. The cost of the
radioisotope transducer and the complexity of its electrical circuitry do not exceed the cost and complexity of
variable inductors.
BRIEF COMMUNICATIONS
Translated from Atomnaya Energiya, Vol. 8, No. 6, p. 578
June, 1960
USSR. In Minsk, construction work was completed on the main building of the 2000 kw (th) research reac-
tor of the Academy of Sciences of the Byelorussian SSR. Assembly of the reactor core and equipment for beams
of radiation is in progress. Electrical engineering equipment, control and measuring instrumentation and auto-
matic controls are being put in place.
The reactor is designed for biological and miscellaneous research projects, production of radioactive iso-
topes, and studies of the behavior of various materials under exposure to neutron and gamma-ray bombardment.
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BIBLIOGRAPHY
NEW LITERATURE
Translated from Atomnaya Energiya, Vol. 8, No. 6, pp. 581-583,
June, 1960
BOOKS AND SYMPOSIA
N. F. Nelipa, Coupling Between Photoproduction of 71. -mesons and Scattering. Moscow, Atomizdat, 1959
88 pages, 4 rubles.
This text gives a systematic review of papers on coupling between photoproduction of ir -mesons on nuc-
leons and scattering, and the application of the results obtained to an analysis of experimental data. The book
is supplemented with Tables of Clebsch-Gordon, Racah, and Z coefficients.
The book is written for scientific research workers engaged in the study of nuclear reactions at high ener-
gies, and for students taking nuclear physics programs.
Controlled Thermonuclear Reactions. A symposium of translated materials. Moscow. Atomizdat, 1960.
319 pages. (Main control board on atomic energy. Directors of science and engineering information and exhib-
its. No. 26). 14 rubles, 80 kopeks.
This symposium contains 20 articles reflecting the level of work on controlled fusion in Britain and West
Germany at the start of 1957. The burden of the symposium centers on two series of articles dealing with the
work of a British group at Harwell and a German group at G8ttingen. Work by American physicists was not
readily available at that time, and is represented in only very limited degree. The symposium concentrates
mainly on theoretical research, and experimental work is represented by several brief communications.
The symposium will prove useful for persons interested in plasma physics and the physics of controlled
thermonuclear reactions.
J. L. Synge, Relativistic Gas. Translated from the English, edited by D. A. Frank-Kamenetskii. Moscow,
Atomizdat, 1960, 140 pages, 4 rubles, 20 kopeks.
Operation "Argus". Translated from the English. Moscow, Atomizdat, 1960, 160 pages, 6 rubles.
This book constitutes a symposium of papers presented at the special symposium on operation "Argus",
held in April 1959.
Light is shed on the results of observations of the behavior of electrons trapped by geomagnetic fields.
The observations were conducted with the aid of artificial earth satellites during a series of nuclear explosions
carried out at heights of 480 km by the USA, in 1958.
The book will be of interest to physicists, astrophysicists, and meteorologists in the first instance. However,
it is also within reach of a broader readership.
D. Ya. Surazhskii, Techniques in Prospecting and Exploration of Uranium Deposits. Moscow, Atomizdat,
1960, 240 pages, 8 rubles, 70 kopeks.
This book is a handbook of techniques on one of the most important branches of prospecting and explo-
rative geology, compiled on the basis of the experience accumulated in prospecting and exploring uranium
deposits in the USSR and abroad.
495
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The book also provides a general picture of the types of uranium deposits which are of interest, and the
main criteria governing prospecting activities are treated. Methods for prospecting uranium beds by radiation,
gas, and salt haloes are described, as well as methods of preliminary prospecting, detailed exploration, and
sampling of ore bodies.
The book is intended for geologist-engineers, technicians, and students in institutes teaching mining, geo-
logy and exploration geology.
Kh. B. Mezhiborskaya, V. L. Shashkin, and I. P. Shumilin, Analysis of Radioactive Ores by the beta-gamma-
Technique. Moscow, Atomizdat, 1960, 64 pages, 1 ruble, 90 kopeks.
This text is devoted to radiometric analysis of samples of uraniferous and uranium-thorium ores with
disturbed radioactive equilibrium.
Topics discussed include the theory of the 8 y -method, the equipment used, special problems in mea-
surement of 8 -y -emission from samples, methods for determining the coefficient in computations and formulas.
Recommendations are given for estimating the accuracy of analyses made by the 8 - y -technique.
A short description is given of methods of radiometric analysis of samples in complex radioactive ores.
The book is written for physicists and geophysicists working in the field of the analysis of radioactive ores.
It may also be found useful by students in the corresponding specialties, as a manual for use in a course on
radiometry.
Extraction and Purification of Exotic Metals. Translated from the English, edited by 0. P.Kolchin,
Moscow, Atomizdat, 1960, 512 pages, 24 rubles, 35 kopeks.
This book consists of a collection of papers presented at the 1956 symposium of the London Institute of
Mining and Metallurgy.
The 22 papers give the results of laboratory research, and in some cases of industrial research, on the tech-
nology of uranium, thorium, beryllium, zirconium, hafnium, niobium, vanadium, titanium, selenium, and
several other rare metals.
The book will be of interest to metallurgical engineers, chemical engineers, ore processors, and scientific
research workers engaged in the field of the production and application of radioactive and exotic metals.
S. V. Elinson and K. I. Petrov, Analytical Chemistry of Zirconium. Moscow, Atomizdat, 1960, 212 pages.
7 rubles, 80 kopeks.
The chemical and physical-chemical properties of zirconium and zirconium compounds are discussed.
The most important analytical reactions and methods for detecting zirconium in other materials are described.
Methods for isolating zirconium from the other elements are discussed. A detailed exposition is given of the
volumetric, gravimetric, calorimetric, and spectral techniques for zirconium assay in alloys, salts,and other
materials. Techniques for determining gas-forming elements and carbon present in zirconium are dealt with
extensively, and the chemical and spectral techniques for determining other trace impurities in zirconium and
components present in zirconium alloys are also discussed.
This text may be used as a practical handbook for workers in plant laboratories and research institutes,
and also as a textbook manual for students in chemical and metallurgical institutes.
M. I. Shal'nov, Neutron Tissue Dose. Edited by B. M. Lsaev, Atomizdat, 1960, 218 pages, 8 rubles, 20
kopeks.
This book represents an attempt to generalize the data culled from the literature, as well as the materials
obtained by the book's author in his own experimental work with neutrons, on all of the principal questions
involved in tissue dosimetry. The first two chapters are devOted to the basic properties and sources of neutrons,
and to the interaction of neutrons with matter. Subsequent chapters discuss the theoretical and experimental
results on depth distribution of absorbed neutron dose in tissue-simulating media; the principal problems asso-
ciated with the study of the relative biological effectiveness of nuclear radiations, the problem of the maximum
tolerable dose, etc. are discussed. The last chapter contains a concise review of the instruments and techniques
of neutron dosimetry.
496
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This book is of interest to biologists and medical workers engaged in the field of radiation biology and
radiation medicine.
Chemical Protection of the Organism from Ionizing Radiations. Edited by V. S. Balabukh. Moscow, Ato-
mizdat, 1960, 152 pages, 4 rubles, 30 kopeks.
The first part of this symposium contains a brief review of the problems encountered in chemical protec-
tion from ionizing radiations, experimental data on the synthesis of, and biological testing of the protective
properties of a number of chemical compounds.
The second portion deals with the results of experimental research on isolation of radioactive isotopes
from the organism. The characteristics of the state of several radioactive isotopes in the blood and bone tissue
are discussed. The physical-chemical and biological evaluation of the effectiveness of complexing agents used
to remove isotopes from the organism are discussed.
The symposium is intended for chemists engaged in the field of searching for means of chemical protec-
tion, and for complexing agents capable of serving that function, as well as for those biologists and other spe-
cialists concerned with radiation biology problems.
G. 0. Davidson, Biological Consequences of Total-Body Irradiation in Humans. Translation from the
English. Edited by M. F. Popova. Moscow, Atomizdat, 1960, 108 pages, 4 rubles, 70 kopeks.
The chief task of this book is an attempt to draw conclusions of practical interest on shielding against the
effects of radioactive precipitates, based on biological and medical data on the effects of ionizing radiations on
animal and human organisms.
Experimental material on the relationship of lethal doses in instantaneous, intermittent, and chronic irra-
diation is discussed. The theory of injury and recovery following radiation injury is developed, and attention
is given briefly to the immediate sequelae of irradiation. The experimental material given points up the path-
ways of possible future biological research having a direct bearing on problems of shielding against radioactive
ionizations.
The book is written for a broad readership of biologists and physicians interested in problems of radiation
shielding.
G. Shreiber, Biophysical radiology. Translated from the German. Moscow, Atomizdat, 1960, 368 pages,
18 rubles, 50 kopeks.
This book is an outline of a course of lectures presented by the author at the Humboldt University in
Berlin, as an introduction to radiology. The book discusses the fundamental concepts of the physics of ionizing
radiations; presents data on the most important physical, physico-chemical, and biological effects occurring
in response to the interaction of radiation with matter; and sheds light on questions concerning dosimetry.
The book is written primarily for biologists and medical workers working in the fields of radiology
radiation selection, etc.
L. S. Kozyreva-Adelsandrova, and N. I. Temnikova, The Radioactive Isotope Iodine-131. Moscow,
Atomizdat, 1960, 24 pages, 50 kopeks.
This brochure describes the properties of radioactive iodine and discusses the various fields of application
for the isotope. Concrete examples of the use of I131 in medical practice for diagnostic procedures and thera-
peutic applications, in chemistry for the study of chemical processes, etc. are given. Recommendations on
rules for handling 1131 and health physics questions are treated. The introductory part of the brochure cites
briefly the characteristics of the isotopes and techniques used in its production.
The brochure is written for a broad audience.
N. P. Galin, A. A. Maiorov, and U. D. Veryatin, Technology of the Processing of Uranium Concentrates.
Moscow, Atomizdat, 1960, 162 pages, 6 rubles, 50 kopeks.
The book gives a short outline of the development of the uranium industry. Fundamental information on
hydrometallurgical processes for isolating uranium from the raw ore material, on reserves of uranium ores,
497
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scales of production, and fields of application for uranium then follow. The bulk of the attention is given over
the technology of processing of uranium concentrates to pure salts and uranium metal.
Methods for producing the most important uranium compounds are described, and their physico-chemical
properties are cited. Production flow charts used in various countries for uranium metal production processes
are compared. A special chapter deals with problems of safety and sanitation in the process of purification of
uranium chemical concentrates.
The book is written for workers in the uranium industry, scientific research organizations,and may be used
for the training of corresponding specialists in advanced institutions.
S. V. Rumyantsev, Radioactive Isotope Applications in Nondestructive Testing. Moscow, Atomizdat, 1960,
296 pages, 11 rubles, 25 kopeks.
The physical and engineering fundamentals concerned in applications of radioactive isotopes in nondes-
tructive testing for flaws in materials are outlined in this book. Characteristics of the isotopes and their fields
of application are cited. The procedure used in radiographing manufactured parts and handling the associated
equipment is discussed. Problems related to the effect of metallurgical flaws on the strength of weld joints are
investigated. A presentation is made of the ionization technique, of xeroradiography, etc. The problems of
safe handling of isotopes, shielding from ionizing radiations,are elucidated.
The book is written for engineers and technicians engaged in flaw detection and nondestructive testing.
L. K. Tatochenko, Radioactive Isotopes in Instrument Design. Moscow, Atomizdat, 1960, 368 pages,
13 rubles, 20 kopeks.
The theoretical and practical problems related to the use of radioactive isotopes in industry are discussed
in this book. The procedure followed in the engineering design of instruments based on the use of radiations is
discussed. Light is shed on problems concerning organization of work with and handling of radioactive isotopes,
concerning the dosimetry of ionizing radiations, and radiation safety practices.
The book is written for engineers and scientists working in various fields of industry.
English-Spanish-Russian-French Dictionary of Scientific and Engineering Terms on Nuclear Energy.
Moscow, Atomizdat, 1959, 215 pages, 16 rubles.
The dictionary is reproduced from the fourth edition of the English-Spanish-Russian-French dictionary of
scientific and technical terms on atomic energy published by the Terminology Section of the UN in 1958. The
dictionary contains about 6,000 terms in four languages. The material is arranged in alphabetical order accord-
ing to English entries. To facilitate searching for Spanish, Russian, and French terms, the dictionary has special
index sections appended.
The dictionary was compiled for the benefit of scientists, engineers, translators, and students in the area
of interest.
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INDEX
THE SOVIET JOURNAL OF ATOMIC ENERGY
Volume 8, Numbers 1-6
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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 ATOMNAY A ENERGIY A,
a publication of the Academy of Sciences of the USSR
(Russian original dated January, 1960)
Vol. 8, No. 1 April, 1961
CONTENTS
Influence of the Reactor Temperature Characteristics Upon the Choice of the Optimum
PAGE
RUSS.
PAGE
Thermodynamic Cycle of an Atomic-Electrical Generating Station. D. D. Kalafati
1
5
Number of Neutrons Emitted by Individual Fission Fragments of U. V. F. Apalin,
10
15
Yu. P. Dobrynin, V.P. Zakharova, I. E. Kutikov, and L. A. Mikaelyan
Method of Estimating the Critical Parameters of a Body of Arbitrary Shape Made from
Fissionable Material. V. G. Zagrafov
17
23
Removal of Oxides from Sodium and Tests for the Oxide Content. P. L. Kirillov, F. A.
Kozlov, V. I. Subbotin, and N. M. Turchin
23
30
On the Change in the Color and Transparency of Glasses when Bombarded by Gamma Rays
from a Co69 Source and in a Nuclear Reactor. S. M. Brekhovskikh
29
37
LETTERS TO THE EDITOR
Mass-Spectrometric and Spectroscopic Studies of Hydrogen Discharge of an Ion Source.
A. I. Nastyukha, A. R. Striganov, I. I. Afanas'ev, L. N. Mikhailov, and M. N. Oganov
35
44
New Isotopes of Holmium and Erbium. N. S. Dneprovskii
38
46
Fission Cross Section of Th229 for Monochromatic Neutrons in the 0.02-0.8 ev Region.
Yu. Ya. Konakhovich and M. I. Pevzner
39
47
Mean Number of Prompt Neutrons per Spontaneous Fission of U238. E. K. Gerling and
Yu. A. Shukolyukov
41
49
The Effect of Boron-Containing Layers on the Yield of Secondary Gamma Radiation.
D. L. Broder, A. P. Kondrashov, A. A. Kutuzov, and A. I. Lashuk
42
49
Critical Heat Flows in the Forced Flow of Liquids in Channels. A. A. Ivashkevich
44
51
Investigation of Heat Transfer in the Turbulent Flow of Liquid Metals in Tubes. M. Kh.
Ibragimov, V. I. Subbotin, and P. A. Ushakov
48
54
Determination of Melting Points of Binary Mixtures of Uranium Oxides with Other Oxides
in Air. S. G. Tresvyatskii and V. I. Kushakovskii
51
56
The Distribution of Iron in Microvolumes of Zirconium Alloys. P. L. Gruzin, G. G.Ryabova,
53
58
and G. B. Fedorov
Reactions of Nitrogen Dissolved in Water, by the Action of Ionizing Radiations.
M. T. Dmitriev and S. Ya. Pshezhetskii
56
59
Method of Calculating Dos age Field of Powerful Isotopic Units. N. I. Leshchinskii
59
62
Integrating Detector of Penetrating Radiation. 0. A. Myazdrikov
62
64
Measurement of Co69 y -Ray Dose Close to the .Boundary between Two Bodies.
V. I. Kukhtevich, B. P. Shemetenko, and B. I. Sinitsyn
64
66
On the Efficiency of Gas-Discharge Counters. V. P. Bovin
67
68
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)
PAGE
RUSS.
PAGE
Airborne Radiometer-Analyzer. V. V. Matveev and A. D. Sokolov
70
70
Investigation of the Production of an Electromotive Force in a System of Semiconductors
with Uranium during Irradiation in a Reactor. Yu. K. Gus'kov, A. V. Zvonarev,
73
72
and V. P. Klychkova
NEWS OF SCIENCE AND TECHNOLOGY
International Symposium on the Metrology of Radioactive Isotopes. K. K. Aglintsev and
V. V. Bochkarev
76
76
International Conference on Accelerators. A. N. Lebedev
78
78
At the Institute for Physical Methods of Separation (German Democratic Republic).
N. M. Zhavoronkov and K. I. Sakodynskii
80
81
[Uranium Production in Canada during 1958
82]
[Use of Ammonium Molybdophosphate in Treating Fission Waste Solutions
84]
Building and Designing of Atomic Powered Vessels in Western and Eastern Countries.
A. V. Klement'ev
82
85
Brief Communications
84
86
BIBLIOGRAPHY
New Literature
85
88
NOTE
The Table of Contents lists all material that appears in AtomnayaEnergiya. Those items
that originated in the English language are not included in the translation and are shown en-
closed in brackets. Whenever possible, the English-language source containing the omitted
reports will be given.
Consultants Bureau Enterprises, Inc.
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Declassified and Approved For Release 2013/02/19: CIA-RDP10-02196R000100050006-3
EDITORIAL BOARD OF
ATOMNAYA ENERGIYA
A. I. Alikhanov
A. A. Bochvar
N. A. DollezhaV
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 February, 1960)
Vol. 8, No. 2 May, 1961
CONTENTS
Thermal Stresses in Reactor Constructions. A. Ya. Kramerov, Ya. B. Fridman, and
PAGE
91
100
RUSS.
PAGE
101
112
? S. A. Ivanov
The Deformation of Uranium Under the Influence of Thermal Cycles During the Simultaneous
Action of an External Tensile Load. A. A. Bochvar, G. Ya. Sergeev, and V. A. Davydov. .
The Separation of Pa233 Without a Carrier from Thorium Nitrate Preparations Irradiated by
Slow Neutrons. V. I. Spitsyn and M. M. Golutvina
105
117
Determination of the Optimum Yield of Enriched Ore in Radiometric Enrichment of Uranium
Ores. E. D. Mal'tsev
108
121
Strong Focusing in a Linear Accelerator. P. M. Zeidlits, L. I. Bolotin, E. I. Revutskii, and
114
127
V. A. Suprunenko
LETTERS TO THE EDITOR ?
Stability of Plasma Bunches in a Waveguide. M. L. Levin
120
134
Self-Reproducing Solutions of the Plasma Equations. B. N. Kozlov
121
135
Complex Fission of Uranium by 2.5-Mev Neutrons. Z. I. Solov'eva
124
137
Fission Cross Sections for Th222, Pu246, Pu241, and Am 241 by Neutrons with Energies of 2.5 and ?
14.6 Mev. M. I. Kazarinova, Yu. S. Zamyatin, and V. M. Gorbachev
125
139
Analysis of Neutron Interactions with Hes', C12, and 016 Nuclei Using an Optical Nuclear
Model. f. Ya. Milthlin and V S Stavinskii_
127
141
Experimental Investigation of Heat Transfer in Slit-Type Ducts with High Heat-Transfer
Rates. Yu. P. Shlykov
130
144
An Investigation of the Alloys of the Uranium-Germanium System. V. S. Lyashenko and
132
146
V. N. Bykov
Coprecipitation of Pu (IV) with Organic Coprecipitants. V. I. Kuznetsov, and T. G. Akimova.
135
148
Contribution to the Problem of Electron Injection to a Betatron. V. P. Yashukov
137
150
Some Data on the Distribution of Radiations Emanating from the Synchrocyclotron of the.
Joint Institute for Nuclear Research. M. M. Komochkov and V. N. Mekhedov
138
152
Dose Field of a Linear Source. V. S. Grammatikati, U. Ya. Margulis, and V. G. Khrushchev . .
140,
154
Experimental Investigation of Scintillation Counter Efficiency. V. P. Bovin
142
155
A Mobile Neutron Multiplier Unit. T. A. Lopovok
145
158
NEWS OF SCIENCE AND TECHNOLOGY
The Production and Use of Stable Isotopes in the USSR
147
160
Conference on the Uses of Large Radiation Sources in Industry and Particularly in Chemical
Processes
151
164
Annual subscriptions 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.
111
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CONTENTS (continued)
PAGE
RUSS.
PAGE
Tashkent Conference on the Peaceful Uses of Atomic Energy. A. Kiv, and E. Parilis
154
167
[Atomic Energy in Italy
169]
[Experiments on Doppler Broadening of Resonance Levels in Uranium and Thorium
171]
[Shielding Design Nomograms
172]
[Uranium Prospecting Methods in France
172]
Standards. Thin? Gamma Sources
156
177
Brief Notes
157
174
INFORMATION AND BIBLIOGRAPHY
New Literature
158
178
A Message from the Central Committee of the Communist Party of the Soviet Union and the
Council of Ministers of the USSR
163
Insert
Mikhail Mikhailovich Konstantinov
166
NOTE
The Table of Contents lists all material that appears in Atomnaya fnergiya. Those items
that originated in the English language are not included in the translation and are shown enclosed
in brackets. Whenever possible, the English-language source containing the omitted reports will
be given.
iv
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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 March, 1960)
Vol. 8, No. 3 May, 1961
CONTENTS
PAGE
RUSS.
PAGE
The Late Frederic Joliot-Curie (On the Occasion of his Sixtieth Birthday) ?
167
A Cyclotron With a Spatially Varying Magnetic Field. D. P. Vasilevskaya, A.A. Glazov,
V. I. Danilov, Yu. N. Denisov, V. P. Dzhelepov, V. P. Dmitrievskii,B. I. Zamolodchikov,
N. L. Zaplatin, V. V. Kol'ga, A. A. Kropin, Liu Nei-ch'uan, V. S. Rybalko,
168
189
A. L. Savenkov, and L. A. Sarkisyan
Acceleration of Ions in a Cyclotron with an Azimuthally Varying Magnetic Field.
R. A. Meshcherov, E. S. Mironov, L. M. Nemenov, S. N. Rybin, and Yu. A. Kholmovskii. .
179
201
Method of Obtaining an Average Value for the Nuclear Constants, Involved in Fast Reactor
Calculations, Taking into Account the Neutron Values. A, I. Novozhilov and
S. B. Shikhov
186
209
The Feasibility of Using Organic Liquids, Heated in Nuclear Reactors, as Working Fluids in
Turbines, from the Thermodynamical Standpoint. P. I. Khristenko
191
214
Some Force and Deformation Characteristics in the Metal Forming of Uranium. I. L. Perlin,
195
219
L. D. Nikitin, V. A. Fedorchenko, A. D. Nikulin, and N. G. Fteshetnikov
Prospecting Criteria for Uranium Deposits. M. M. Konstantinov
203
228
Dosimetry of Intermediate-Energy Neutrons. A. G. Istomina and I. B. Keirim-Markus
212
239
LETTERS TO THE EDITOR
The Neutron-Deficient Isotope Ho155. B. Dalkhsuren, I. Yu. Levenberg, Yu. V. Norseev,
219
248
V. N. Pokrovskii and S. S. Khainatskii
Determination of the Dampness of Dry Granular Substances, by Means of Neutron Moderation.
A. K. Val'ter and M. L. Gol'bin
220
248
Local and Mean Heat-Transfer for a Turbulent Flow of Nonboiling Water in a Tube with High
Heat Loads. V. V. Yakovlev
221
250
On the Question of the Choice of Heat Carriers for Nuclear Reactors. E. I. Siborov
224
252
Turbulent Temperature Pulsations in a Liquid Stream. V. I. Subbotin, M. I. Ibragimov,
226
254
and M. N. Ivanovskii
Electrolytic Preparation of Layers of Uranium Compounds with Densities of 1-3 mg/cm2.
V. F. Titov
229
257
Solubility of Uranium (IV) Hydroxide in Sodium Hydroxide. N. P. Galkin and M. A. Stepanov. .
231
258
Catalytic Effect of Iron Compounds in the Oxidation of Tetravalent Uranium in Acid Media.
Vikt. I. Spitsyn, G. M. Nesmeyanova, and G. M. Alkhazashvili
233
261
Effects of Gamma Radiation on the Electrode Properties of Lithium Glass. N. A. Fedotov
235
262
Measurement of Gamma-Radiation Dose by the Change in Optical Activity of Certain
Carbohydrates. S. V. Starodubtsev, Sh. A. Ablyaev, and V. V. Generalova
237
264
Annual subscription S75.00 @ 1961 Consultants Bureau Enterprises, Inc., 227 West 17th St., New York II, N.Y.
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CONTENTS (continued)
NEWS OF SCIENCE AND TECHNOLOGY
VII Session of the Learned Council of the Joint Institute for Nuclear Research (Dubna)
PAGE
RUSS.
PAGE
M. Lebedenko.
239
266
Conference of Representatives of 12 Governments. M. Lebedenko..
241
267
T II All-Union Technical-School Conference on Electron Accelerators Yu.M. Ado and
242
268
K. A. Belovintsev
Symposium on Extraction Theory. LV. Seryakov
243
269
Development of Nuclear Power in Sweden. M. Sokolov
245
270
[Research Reactors in West Germany
273]
[Start-Up of a BWR in Norway
275]
Plasma Research on the Stellarator
247
zr
[Entropy Trapping of Plasma by a Magnetic Field with Inflation of Magnetic Bottle
281]
[New Electrostatic Accelerator Designs
283]
[American Research in the Area of Nuclear Fuel Processing
285]
New Shielding Materials
252
285
BRIEF NOTES
252
286
BIBLIOGRAPHY
New Literature
253
289
NOTE
The Table of Contents lists all material that appears in Atomnaya E/nergiya. Those items that originated
in the English language are not included in the translation and are shown enclosed in brackets. Whenever
possible , the English-language source containing the omitted reports will be given.
vi
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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 ATOMNAY A ENERGIY A,
a publication of the Academy of Sciences of the USSR
(Russian original dated April, 1960)
Vol. 8 No. 4 June, 1961
CONTENTS
PAGE
RUSS.
PAGE
Lenin on Science and Industry. I. I. Kul'kov
259
301
Design of the VVR-S Research Reactor. V. F. Kozlov and M. G. Zemlyanskii
263
305
Ion Cyclotron Resonance in Dense Plasmas. L. V. Dubovoi, 0. M. Shvets, and
273
316
, S. S. Ovchinnikov
Electrolytic Isolation of Small Amounts of Uranium, Neptunium, Plutonium, and Americium
A. G. Samartseva
279
324
Pole of Oxidation-Reduction Processes in the Solution of Uranium Oxides in Acid Media.
G. M. Nesmeyanova and G. M. Alkhazashvili
284
330
Composite Radiometric Work in Mining. I. M. Tenenbaum
289
336
Heat-Treatment of Uranium. G. Ya. Sergeev, V. V. Titova, Z. P. Nikolaeva, and
A. ?M. Kaptel'tsev
292
340
Investigation of the Internal Friction Increase in Polycrystalline Uranium Specimens Caused
by Temperature Changes. Yu. N. Sokurskii and Yu. V. Bobkov
299
348
Methods of Radioactivity Metrology in USSR. K. K. Aglintsev, V. V. Bochkarev,
304
354
V. N. Grablevskii, and F. M. Karavaev
LETTERS TO THE EDITOR
Cross Section for the Reaction Th232 (n, 2n)Th231 at 14.7 Mev Neutron Energy.
Yu. A. Zysin, A. A. Kovrizhnykh, A. A. Lbov, and L. I. Sel'chenkov
310
311
360
361
y-Radiation Emitted by U238 Under the Action of 14 Mev Neutrons. A. I. Veretennikov,
V. Ya. Averchenkov, M. V. Savin, and Yu. A. Spekhov
A Study of Scintillations in Helium at Liquid Helium Temperatures. B. V. Gavrilovskii . . . .
313
363
Mass-Spectrometric Analysis and the Identification of Technetium. G. M. Kukavadze,
R. N. Ivanov, V. P. Meshcheryakov, Yu. G. Sevast'yanov, B. S. Kir'yanov, V. I. Galkov,
316
365
and A. P. Smirnov-Averin
Heat Transfer to Sodium at Low Re Numbers. M. S. Pirogov
318
367
Separation of Lithium Isotopes on a Simple Ion-Exchange Column. G. M. Panchenkov,
319
368
E. M. Kuznetsova, and L. L. Kozlov
Some Aspects of Aerial 7-Ray Prospecting Over Forested Regions. G. N. Kotel'nikov and
N. L Kalyakin
321
370
On the Accuracy of Calculation of the Build-Up Factor for 7-Rays in Thin Absorbing and
Scattering Media. A. V. Bibergal' and N. I. Leshchinskii
324
372
Radiation Field Due to a Cylindrical Source Placed Behind a Plane Screen. D. P. Osanov
325
374
and E. E. Kovalev
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CONTENTS (continued)
An Investigation of Certain Artificially Radioactive Isotopes and Their Use in Medical
Radiography. I. A. Bochvar, V. E. Busygin, and U. Ya. Margulis
PAGE
327
RUSS.
PAGE
376
NEWS OF SCIENCE AND TECHNOLOGY
Tenth All-Union Conference on Nuclear Spectroscopy. 0. Kraft
330
378
At the Institute of Physics of the Academy of Sciences of the Ukrainian SSR. (A conversa-
tion with the vice-director of the Institute of Physics in charge of scientific research,
0. F. Nemets). V Parkhit'ko
332
380
[Utilization of Nuclear Power in Brazil and Argentina
381]
[Plans for the Development of Nuclear Power in Spain
382]
[Start-Up of a Fast Power Reactor at Dounreay
384]
[The Nuclear Power Station at Latina (Italy)
387]
[The Turret High-Temperature Gas-Cooled Reactor
389]
[Recent Data on Neutron Cross Sections
391]
[Fission Parameters for U235
392]
[New Uradium Deposits Outside of the USSR
392]
[Industrial Unit for Exposure of Materials to Radiation
396]
Brief Communications
333
397
BIBLIOGRAPHY
New Literature
334
398
viii
NOTE ?
The Tables of Contents lists all material that appears in Atomnaya Energlya. Those items
that originated in the English language are not included in the translation and are shown en-
closed in brackets. Whenever possible, the English-language source containing the omitted
reports will be given.
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I ,
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EDITORIAL BOARD OF
ATOMNAYA gNERGIYA
A. I. Alikhanov
A. A. Bochvar
N. A. Dollezhali
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 ATOMNAY A ENERGIY A,
a publication of the Academy of Sciences of the USSR
(Russian Original Dated May, 1960)
Vol. 8, No. 5 June, 1961
CONTENTS
Winners of Lenin Prizes
Determination of the Mean Number of Secondary Fission Neutrons from the Fragment
PAGE
341
RUSS.
PAGE
Mass Distribution. Yu. A. Zysin, A. A. Lbov? and L. I. Sel'chenkov
343
409
An Investigation of the Properties of Metals and Some Steels after Irradiation
by Fast Neutrons. Sh. Sh. Ibragimov, V. S. Lyashenko, and A. I. Zav'yaloy
347
413
Vapor Pressure of T20. M. M. Popov and F. I. Tazetdinov
353
420
Radiometric Analysis of Ores on Conveyers. L. N. Posik, S. I. Babichenko,
358
425
and R. A. Grodko
Angle-Energy Distribution of y-Radiation Scattered in Water and Iron. Yu. A. Kazanskii
364
432
Universal Apparatus with a Co60 y -Ray Source with an Activity of 60,000 g-eq of Ra
for Simulating Radiation-Chemical Apparatuses, and Investigations (The "K-60.000"),
A. Kh. Breger, V. B. Osipov, and V. A. Gordin
371
441
LETTERS TO THE EDITOR
Investigation of the Spent Fuel Element of the First Atomic Power Station. A. P. Smirnov-
Averin, V. I. Galkov, Yu. G. Sevastiyanov, N. N. Krot, V. I. Ivanov, I. G. Sheinker,
375
446
L. A. Stabenova, B. S. Kiriyanov, and A. G. Kozlov
On Improving the Efficiency of Power Station Reactors with Gaseous Coolants.
T. Kh. Margulova and L. S. Sterman
377
448
Measurement of the Fast Neutron Flux Distribution in the Core of the VVR-S Reactor
with Respect to Changes in the Electrical Conductivity of Germanium Specimens.
E. Aleksandrovich and M. Bartenbakh
381
451
Calculation of Thermal Shocks in Reactor Structural Parts. Yu. E. Bagdasarov
383
452
600-key Proton Injector for a Linear Accelerator. Yu. N. Antonov, L. P. Zinov'ev,
386
454
and V. P. Rashevskii
Mean Number of Prompt Neutrons Emitted in Photofission of Th232 and U 2 '3 8 by y -rays
Produced in the F19 (p, 07)016 Reaction. L. I. Prokhorova and G. N. Smirenkin . . . ?
390
457
Electron Acceleration in a Traveling-Wave Cyclical Waveguide Accelerator.
A. A. Vorob'ev, A. N. Didenko, and E. S. Kovalenko
392
459
Use of Scintillation Counters in Gammascopy. V. E. Nesterov
394
461
NEWS OF SCIENCE AND TECHNOLOGY
Atomic Energy at the Soviet Exposition in Havana. L. Kimel', and V. Tsurkov
397
464
Atomic Energy of the All-China Exposition on Industry and Means of Communication.
Shen Chung-po
399
464
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Declassified and Approved For Release 2013/02/19: CIA-RDP10-02196R000100050006-3
CONTENTS (continued)
[Washington Conference of the American Nuclear Physics Society and Atomic Industrial
Forum. Sources; Nucleonics 17, No. 12, 17-23 (1959); Nuclear Power 5, No. 45,
' 111-116 (1960)
PAGE
RUSS.
PAGE
467]
[Development of Nuclear Power in the Countries of South and Central America
467]
[Organic Moderated Reactors for Land-Based and Seagoing Facilities ?
470]
[Reactor as a Neutron Source
472]
Measurement of Magnetic Moment of Li8-
400
473
[New Foreign Articles on Rolling of Uranium
474]
[On the Use of Statistical Analysis Techniques in Explorations for Uranium Deposits.
Source; R. Bates, Econ. Geol. 54, No. 3, 449 (1959)
476]
BIBLIOGRAPHY
New Literature
401
480
NOTE
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 en-
closed in brackets. Whenever possible, the English-language source containing the omitted
reports will be given.
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Declassified and Approved
EDITORIAL BOARD OF
ATOMNAYA gNERGIYA
A. I. Alikhanov
A. A. Bochvar
N. A. Dollezhali
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
For Release 2013/02/19: CIA-RDP10-02196R000100050006-3
THE SOVIET JOURNAL OF
ATOMIC ENERGY
A translation of ATOMNAY A ENERGIY A,
a publication of the Academy of Sciences of the USSR
(Russian original dated June, 1960)
Vol. 8, No. 6 July, 1961
CONTENTS
RUSS.
PAGE PAGE
The 50-Megawatt SM Research Reactor. S. M. F einberg, S. T. Kono
beevskii, N. A. Dollezhal', I. Ya. Emel'yanov, V. A.
Tsykanov, Yu. M. Bulkin, A. D. Zhirnov, A. G. Filippov,
0. L. Shchip,akin, V. P. Perfil'ev, A. G. Samoilov,
and V. I. Ageenkov 409 493
New Ideas in the Structural Design and Layout of Nuclear Reactors.
A. N. Komarovskii 420 505
Mechanical Properties and Microstructure of Certain Construction Materials
After Neutron Irradiation. I. M. Voronin, V. D. Dmitriev, Sh. Sh.
Ibragimov, and V. S. Lyashenko 429 514
Extraction of Uranium from Solutions and Pulps. B. N. Las k or in, A P.
Zefirov, and D. I. Skorovarov 434 519
Interaction of Uranium Hexafluoride with Ammonia. N. P. G a lkin,
B. M. Sudarikov, and V. A. Zaitsev 444 530
The Flocculation of Pulp and Polyacrylamide-Type Flocculents. I. A. Yak ubo v ich 449 535
Determination of Absorbed Doses in Organisms Exposed to Emanations
and Their Daughter Products. L. S. Ruzer 455 542
LETTERS TO THE EDITOR
Absorption Section of Fast Neutrons. T. S. Be 1 an ova 462 549
Convergence of the Series in the Many-Velocity Theory of Neutron Diffusion.
A. V. Stepanov 464 550
A Ring Cyclotron Accelerator with Vertically Growing Magnetic Field.
A. P. Fateev and B. N. Yablokov 468 552
Some Properties of AcCelerator Orbits Where Similitude Is Observed.
A. A. Kolomenskii and A. N. Lebedev 471 553
Measurement of the Radiative-Capture y -Emission Spectra of Neutrons in Some Rocks.
A. A. Fedorov, M. M. Sokolov, and A. P. Ochkur 474 555
Slowing Down of Neutrons in Steel-Water Mixtures. L. A. Geraseva
and V. V. Vavilov 476 556
Determination of Degrees of Equilibrium of Short-Lived Radon Daughters in Air.
L. S. Ruzer 478 557
Luminescent Dosimeters Based on the CaS0 ? Mn Phosphor for the Detection
of Gammas, Betas, and Neutrons. V. A. A rkh an g el ' sk a y a , B. I.
Vainberg, V. M. Kodyukov, and T. K. Razumova 481 559
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Declassified and Approved For Release 2013/02/19: CIA-RDP10-02196R000100050006-3
CONTENTS (continued)
RUSS.
PAGE PAGE
SCIENCE AND ENGINEERING NEWS
Letter from a Reader (on the Article "Entropy Trapping of a Plasma by a Reversal of the
Magnetic Bottle Configuration"). L. A. Art si mo v i ch 485 562
Generation of a I-Hyperon by Negative Pions with a Momentum of 8.3 Bev/c.
M. I. Solov'ev 486 562
Hungarian Exhibit of Instruments for Experimental Nuclear Physics Research 488 563
[Economics of Organic-Cooled Organic-Moderated Low-Power Reactors. V. V. B at ov 564]
[The Piqua Organic-Moderated Reactor. V. V. B at ov 565]
[The New British Research Reactor (Jason). A S el igm an 568]
[Nuclear Power Developments in West Germany. Yu. M it y a ev 570]
[Uranium Production in the Union of South Africa. R. R a f al' ski i 572]
Recent Data on C14 Concentration in the Atmosphere. Yu. V. S i v int s e v 489 573
Applications of Alpha Radiation from Radioactive Isotopes for Quality Control in
Grinding Operations. V. V. Kondashevskii, A. N. Chertovskii,
V. S. Pogorelyi, and A. M. Gutkin 492 576
Brief Communications 494 578
BIBLIOGRAPHY
New Literature ? Books and Symposia 495 581
INDEX FOR JANUARY-JUNE, 1960
Table of Contents, Volume 8
Author Index xiii
NOTE
The Table of Contents lists all material that appears in Atomnaya fnergiya. 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.
Consultants Bureau Enterprises, Inc.
ERRATA
Vol. 8, No. 4, June, 1961
Page Column Line Reads
277 left 7-8 cog /k2c2., 1
278 left 14-18 The increase . . .
xii
Should read
wg/k2c2 5.. 1
The increase in the magnitude of the longitudinal veloc-
ity component v vo to values v II > vo means
that it is necessary to take account of the additional re-
duction of Tc (E) due to the reduction in the time spent
by the ion (1. ) in the region of the heating section:
II c 1 kv11'
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Declassified and Approved For Release 2013/02/19: CIA-RDP10-02196R000100050006-3
AUTHOR INDEX
THE SOVIET JOURNAL OF ATOMIC ENERGY
Volume 8, ,Numbers 1-6
(A translation of Atocnnaya energiya)
Ablyaev. Sh. A.
237
Dmitrievskii, V. P.
- 168
Ivanovskii, M. N.
- 226
Ado, Yu. M.
242
Dneprovskii, N. S.
- 38
Ivashkevich, A. A.
- 44
Afanas'ev, I. I.
- 35
Dobrynin, Yu. P.
- 10
Ageenkov, V. I.
409
Dollezhal', N. A.
- 409
Kalafati, D. D.
- 1
Aglintsev, K. K.
- 76
Dubovoi, L. V.
273
Kalyakin, N. I.
- 321
Aglintsev, K. K.
304
Dzhelepov, V. D.
168
Kaptertsev, A. M.
- 292
Akimova, T. G.
135
Karavaev, F. M.
- 304
Aleksancirovich, E.
381
Emeryanov, I. Ya.
- 409
Kazanskii, Yu. A.
- 364
Alkhazashvili, G. M.
233
Kazarinova, M. I.
- 125
Alkhazashvili, G. M.
284
Fateev, A. P.
468
Keirim-Markus, I. B.
- 212
Antonov, Yu. N.
386
Fedorchenko, V. A.
195
Khainatskii, S. S.
- 219
Apalin, V. F.
- 10
Fedorov, A. A.
- 474
Kholmovskii, Yu. A.
- 179
Arkhangel'skaya, V. A.
- 481
Fedorov, G. B.
- 53
Khristenko, P. I.
- 191
Artsimovich, L. A.
- 485
Fedotov, N. A.
- 235
Khrushchev, V. G.
140
Averchenkov, V. Ya.
- 311
Feinberg, S. M.
- 409
Kimer, L.
- 397
Filippov, A. G.
- 409
Kirillov, P. L.
- 23
Babichenko, S. I.
358
Fridrnan, Ya. B.
91
Kir'yanov, B. S.
- 316
Bagdasarov, Yu. E.
383
Kir'yanov, B. S.
- 375
Bartenbakh, M.
381
Galkin, N. P.
- 231-
Kiv, A.
154
Belanova, T. S.
462
Galkin, N. P.
- 444
Klement'ev, A. V.
- 82
Belovintsev, K. A.
242
Galkov, V. I.
- 316
Klychkova, V. P.
- 73
Bibergal', A. V.
324
Galkqv, V. I.
375
Kodyukov, V. M.
- 481
Bobkov, Yu. V.
- 299
Gavrilovskii, B. V.
- 313
Korga, V. V.
- 168
Bochkarev, V. V.
- 76
Generalova, V. V.
- 237
Ko1omenskii, A. A.
- 471
Bochkarev, V. V.
- 304
Geraseva, L. A.
- 476
Komarovskii, A. N.
- 420
Bochvar, A. A.
100
Gerling, E. K.
- 41
Komochkov, M. M.
- 138
Bochvar, I. A.
- 327
Glazov, A. A.
168
Konakhovich, Yu. Ya.
- 39
Bolotin, L. I.
- 114
Gorbin, M. L.
220
Kondashevskii, V. V.
- 492
Bovin, V. P.
- 67
Gol'din, V. A.
371
Kondrashov, A. P.
- 42
Bovin, V. P.
- 142
Golutvina, M. M.
- 105
Konobeevskii, S. T.
- 409
Breger, A. Kh.
- 371
Gorbachev, V. M.
- 125
Konstantinov, M. M.
- 203
Brekhovskikh, S. M.
- 29
Grablevskii, V. N.
- 304
Koternikov, G. N.
- 321
Broder, D. L.
- 42
Grammatikati, V. S.
- 140
Kovalenko, E. S.
- 392
Bulkin, Yu. M.
409
Grodko, R. A.
- 358
Kovalev, E. E.
- 325
Busygin, V. E.
327
Gruzin, P. L.
- 53
Kovrizhnykh, A. A.
- 310
Bykov, V. N.
132
Gus'kov, Yu. K.
73
Kozlov, A. G.
- 375
Gutkin, A. M.
492
Kozlov,
N. B. , B N
- 121
Chertovskii, A. N.
- 492
Kozlov, F. A.
- 23
Ibragimov, M. I.
226
Kozlov, L. L.
- 319
Dalkhsuren. B.
- 219
Ibragimov, M. Kh.
- 48
Kozlov, V. F.
- 263
Danilov, V. I.
- 168
Ibragimov, Sh. Sh.
- 429
Kraft, 0.
- 330
Davydov, V. A.
- 100
Ibragimov, Sh. Sh.
- 347
Kramerov, A. Ya.
- 91
Denisov, Yu. N.
- 168
Istomina, A. G.
- 212
Kropin, A. A.
- 168
Didenko, A. N.
- 392
Ivanov, R. N.
- 316
Krot, N. N.
- 375
Dmitriev, M. T.
- 56
Ivanov, S. A.
- 91
Kukavadze, G. M.
- 316
Dmitriev, V. D.
- 429
Ivanov, V. I.
375
Kukhtevich, V. I.
- 64
xiii
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Declassified and Approved For Release 2013/02/19: CIA-RDP10-02196R000100050006-3
Kul' kov, I. I.
Kushakovskii, V. I.
Kutikov, I. E.
Kutuzov, A. A.
Kuznetsov, V. I.
Kuznetsova, E. M.
Lashuk, A. I.
Laskorin, B. N.
Lbov, A. A.
Lbov, A. A.
Lebedenko, M.
Lebedev, A. N.
Lebedev, A. N.
Leshchinskii, N. I.
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51
10
42
135
319
- 42
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343
239
78
471
59
324
219
- 120
- 168
- 145-
- 132
- 347
- 429
- 108
- 140
- 327
377
70
138
179
316
10
35
127
179
62
35
179
233
- 284
- 394
- 195
- 292
- 195
- 219
186
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- 35
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124
311
105
233
375
237
127
231
464
377
35
23
48
- 226
- 444
- 114
353
289
229
- 292
- 51
- 397
- 409
- 23
- 48
481
220
168
476
311
392
429
- 468
- 221
- 449
- 137
17
444
10
1-68
125
- 168
- 347
- 434
- 114
- 263
- 80
409
386
73
310
343
xiv
Declassified and Approved For Release 2013/02/19: CIA-RDP10702196R000100050006-3
Declassified and Approved For Release 2013/02/19 : CIA-RDP10-02196R000100050006-3
PROCEEDINGS OF THE ALL-UNION SCIENTIFIC AND TECHNICAL
CONFERENCE ON THE APPLICATION OF RADIOACTIVE ISOTOPES
MOSCOW, 1957
Application of Radioactive Isotopes in Biochemistry and
the Study of Animal Organisms
Jan.-Feb., 1959 heavy paper covers 20 papers',
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Application 'of Radioactive Isotopes in the Food
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Jan.-Feb., 1959 heavy paper covers 16 papers,
illustrated $30.00
,Application of Radioactive Isotopes in Microbiology_
Jan.-Feb., 1959 heavy Paper covers 5 papers,
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Radiobiology
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Individual volumes may be purchased separately
The utilization of isotopes and radiation in biology,
-medicine, and agriculture is covered in 78- reports.
Included in these significant papers are the:latest
? Soviet techniques in the Action of radiation on the,
living organism for the purpose of producing directed
changes in plants and animals, curing of human ill-
nesses and the utilization of isotopes as tagged Atoms
In the studir of vital processes,. Every biologist, chem-
ist, health physicist, and physician employing the
techniques should have access to this outstanding
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Note: Individual reports from each volume are
available at $12.50 each. We will gladly supply
a detailed _table of contents upon request.
CONSULTANTS BUREAU
227 WEST 17TH STREET. NEW YORK 11 N Y
Declassified and Approved For Release 2013/02/19: CIA-RDP10-02196R000100050006-3
Declassified and Approved For Release 2013/02/19: CIA-RDP10-02196R000100050006-3
SOVIET
T,G0
ea
[im
,
ANALYTICAL CHEMISTRY
sCDP UMniKKINg
A collection of ten papers from the Consultants Bureau
translations of the Soviet Journal of Analytical Chemistry
and the famous "Doklady" of the ,Academy of Sciences
(1949-,58) .. This 'collection will acquaint'" the analytical
chemist working in this field with Soviet techniques for
the, determination of uranium in solutions, in ores and the
products of their treatments, and in accessory minerals,
plus methods for the determination of impurities in uranium.
heavy paper covers , illustrated .$10.00
CONTENTS
,
,? Extraction of Uranyl a -Nitros- 8 -naphthokate and Sepa-
ration of Uranium froiri Vanadium and Iron.
? The Composition 9f Uranyl Selenite. A Volumetric Method
of Determining Uranium.
? The Composition of ,the Luminescence Center of Sodium
Fluoride Beads Activated by Uranium.
? Rapid Luminescent Determination of Uranium in Solutions. '\
? Preparation of Slightly Soluble Compounds of Quadrivalent
Uranium Using Rongalite.
?? Investigation ? of Complex Compounds of the Uranyl Ion
Which are of Importance in Analytical Chemistry.
? Uranyl and 'Thorium Selenites.
? The Evaporation Methbd and $ Use for the Determination
of Bpron and Other Impurities in Uranium.
? Spectrographic Determination of Uranium in Ores and the
Products Obtained by Tieatment of These Ores. 1
Determination of Urahium in Accessory Minerals.
CONSULTANTS BUREAU
227 WEST 17TH STREET, NEW YORK 11. N Y
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