SOVIET ATOMIC ENERGY VOL. 42, NO. 6

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• Sequence Number: 6
• Publication Date: December 1, 1977
• Content Type: REPORT

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Russian Original Vol. 42, No. 6, June, 1977
rc
December, 1977
SATEAZ 42(6) 503-598 (1977)
SOVIET
ATOMIC
ENERGY
ATOMHAA 3HEPI1411
(ATOMNAYA iNERGIYA)
-TRANSLATED FROM RUSSIAN
CONSULTANTS BUREAU, NEW YORK
q)}
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Soviet Atomic Energy is a cover-to-cover translation of Atomnaya
Energiya, a publication of the Academy of Sciences of the USSR.
SOVIET
ATOMIC
ENERGY
Soviet Atomic Energy is abstracted or in-
dexed in Applied Mechanics Reviews, Chem-
ical Abstracts, Engineering Index, INSPEC?
Physics Abstracts and Electrical and Elec-
tronics Abstracts, Current Contents, and
Nuclear Science Abstracts.
An agreement with the CopyrightAgeacy of the USSR (VAAP)
makes available both advance copies of the Russian journal and
original glossy photographs and artwork. This serves to decrease
the necessary time lag between publication of the original and
publication of the translation and helps to improve the quality
of the latter. The translation began with the first issue of the
Russian journal.
Editorial Board of Atomnaya Energiya:
Editor: 0. D. Kazachkovskii
Associate Editor: N. A. Vlasov
A. A. Bochvar
N. A. Dollezhal'
V. S. Fursov
I. N. Golovin
V. F. Kalinin
A. K. Krasin
V. V. Matveev
M. G. Meshcheryakov
V. B. Shevchenko
V. I. Smimov
A. P. Zefirov
Copyright ? 1977 Plenum Publishing Corporation, 227 West 17th Street, New York,
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SOVIET ATOMIC ENERGY
A translation of Atomnaya energiya
December, 1977
Volume 42, Number 6
June, 1977
ARTICLES
CONTENTS
Engl./Russ.
Uranium Ore Systems. Experience with Morphogenetic Grouping ? D. Ya. Surazhskii .
503
443
Low-Additive Uranium Alloys for the Fuel Elements of the KS-150 Reactor
? V. F. Zelenskii, A. L Stukalov, V. P. Ashikhmin, and A. V. Azarenko
513
452
Critical Energy in Heat-Transfer Channels of Complex Shape ? L. N. Polyanin
519
457
%Radiation Testing of Materials in Reactors ? B. A. Briskman and E. A. Kramer-Ageev
524
461
`Model Studies of the Stressed State in Pressure Vessels ? N. N. Zorev,
Yu. S. Safarov, V. K. Tutynin, V. N. Sakhelashvili,
and N. L. Narskaya
528
465
Some Problems on the Mechanism of Breakup and Mass Transfer in Pulse Extraction
Columns ? E. L Akharov and S. M. Karpacheva
539
473
Investigation of the Adiabatic Expansion of Water Vapor from the Saturation Line in
Laval Nozzles ? E. K. Karasev, V. V. Vazinger, G. S. Mingaleeva,
and E. I. Trubkin
545
478
The Origin of the Tracks of Fission Fragments in Whitlockite from the
Bjurbole Meteorite ? V. P. Perelygin, S. G. Stetsenko,
and N. Bhandari
550
482
Sputtering and Blistering in the Bombardment of Inconel, SiC+C Alloy, and
Carbon-Pyroceramic by H+ and He+ Ions ? N. P. Busharov,
V. M. Gusev, M. L Guseva, Yu. L. Krasulin, Yu. V. Martynenko, S. V. Mirnov,
and I. A. Rozina
554
486
ABSTRACTS OF PAPERS DEPOSITED AT VINITI
Some Characteristics of the Irradiation of Specimens during Activation Studies
in a Fast Reactor ? L. N. Yurova, A. V. Bushuev, V. N. Ozerkov, V. V. Chachin,
G. I. Gadzhiev, and A. V. Inchagov
560
490
The Possibility of Allowance for the Fine Structure of the Spectrum in
Calculations of Fast Reactors and Installations ? A. G. Morozov, Yu. A. Zverkov,
A. M. Sirotkin, I. S. Slesarev, and V. V. Khromov
562
492
The Calculation of Resonance Absorption in Infinite Multicomponent Media
? A. G. Morozov, Yu. A. Zverkov, A. M. Sirotkin, I. S. Slesarev,
and V. V. Khromov
563
492
The Calculation of the Escape of Photoneutrons from Thick Targets in the Region
of.Giant Resonance ? V. P. Kovaiev and V. I. Isaev
564
493
Neutron Distribution in Aqueous Solutions of a-Active Substances
? A. S. Bespyatykh, E. A. Parfent'ev, V. A. Peregudov, E. M. Tsenter,
and E. V. Chvankin
565
494
Errata
565
494
LETTERS TO THE EDITOR
Specific Sensitivity of an NaI(T1) Scintillator in the Recording of)' Rays from
Radioactive Ores ? G. F. Novikov, A. Ya. Sinitsyn, and Yu. 0. Kozydna
566
495
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Thermodynamic Properties of Liquid Pu +In Alloys - V. A. Lebedev, V. I. Kober,
V. G. Serebryakov, G. N. Kazantsev, I. F. Nichkov, S. P. Raspopin,
CONTENTS
(continued)
Engl./Russ.
and 0. V. Skiba
568
496
Temperature Dependence of Quasielastic Scattering of Slow Neutrons by Water
- A. G. Novikov?and S. M. Iskenderov
571
498
Fine Structure of Combination-Scattering Lines and Laser Plasma Diagnostics
- A._F. Nastoyashchii
574
501
Use of the Calorimetric Method to Measure the Energy Release in a Compensating Rod
- A. S. Zhilkin, V. P. Koroleva, N. P. Kurakov, L. A. Chernov,
and E. V. Shestopalov
576
502
Behavior of Chromel-Alumel Thermocouples during an Emergency Shutdown of the
BR-10 Reactor - A. S. Kruglov, P. V. Vyrodov, and M. L Redchenko
579
504
Total Neutron Cross Sections of 230Th in the Thermal Energy Region and 226Ra Up to 1 keV
- R. N. Ivanov, S. M. Kalebin, V. S. Artamonov, G. V. Rukolaine,
N. G. Kocherygin, S. L Babich, S. N. Nikoltskii, T. S. Belanova.,
and A. G. Kolesov
581
505
Total Neutron Cross Sections and Resonance Parameters of 175,176Lu Up to 200 eV
- S. M. Kalebin, V. S. Artamonov, R. N. Ivanov, G. V. Rukolaine, T. S. Belanova,
A. G. Kolesov, and V. A. Poruchikov
583
506
CONFERENCES
All-Union Conference on the Preparation of Operating Personnel for Atomic Power
Plants - L. M. Voronin, G. N. Ushakov, and V. M. Gordina
587
510
Fifth International Conference on the Peaceful Use of Underground Nuclear Explosions
- I. D. Morokhov, M. P. Grechushkina, and V. N. Rodionov
588
510
NEW APPARATUS AND INSTRUMENTS
A Radiation-Chemical Installation with an EU-0.4 Electron Accelerator for
Obtaining Organosilicon Monomers - B. L Vainshtein,
D. M. Margolin, I. A. Ryakhovskaya, and N. G. Ufimtsev
590
512
The RID-41 Universal Hose-Type 'y Ray Flaw Detector - V. N. Glebov
and V. G. Firstov
592
513
BOOK REVIEWS
A. A. Vantkov, A. L Voropaev and L. N. Yurova. The Analysis of a Reactor-Physics
Experiment - Reviewed by A. D. Zhirnov
594
515
E. A. Stumbur. The Application of Peuturbation Theory to the Physics of
Nuclear Reactors - Reviewed by V. N. Artamkin
595
515
L. S. Tong. Boiling Crises and Critical Heat Flux - Reviewed by N. S. Ithlopkin
596
516
The Russian press date (podpisano k pechati) of this issue was 5/24/1977.
Publication therefore did not occur prior to thisdate, but must be assumed
to have taken place reasonably soon thereafter.
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ARTICLES
URANIUM ORE SYSTEMS.
EXPERIENCE WITH MORPHOGENETIC GROUPING
D. Ya. Surazhskii
An ore system is a collection of ore manifestations which are regularly connected with
one another, have unified structural control, and are enclosed in a space whose boundaries
may vary depending on economic factors [1].
Each ore system is usually a component of another, more complicated system and can be
divided into a number of subordinate systems. The basis of the hierarchical structure so
formed should be taken to be the ore body ? a system which consists of ore aggregates and
rocks containing no ore and which includes useful components in a quantity that justifies
the future costs of ore extraction and processing.
A collection of related ore bodies constitutes an ore bed, a collection of ore beds
constitutes an ore deposit, and a collection of deposits constitutes an ore field. It is
assumed that the shape, dimensions, and spatial position of the components making up ore
systems at a particular level of the hierarchical structure is the result of a certain com-
plex of external-environment factors which are peculiar to this level alone.
If the ore is formed at the same time as the surrounding rocks, then the ore systems
are subject to one form of control ? stratigraphic control (syngenetic sedimentary systems)
or facies control (syngenetic magmatogenic systems). The morphological features of the
mineralization are usually determined by two or more factors. Some of them are determined
by the boundaries of the region (or zone) of ore genesis, others by the boundaries of ore
systems within this region (or zone). The factors of the first group are called ore-control-
ling factors, while those of the second group are called ore-localizing factors. These
terms are conditional and should be used only in connection with the level at which the study
of ore elements is being conducted: the same element of geological structure (e.g., a fissure
developing into a large fracture) may be a localizing factor with respect to the deposit and
a controlling factor with respect to the ore beds making up this deposit.
Types of Uranium Ore Systems. Uranium ore systems can be grouped according to many
criteria. In a survey, an important role is played by the morphological criterion determin-
ing the method used in the survey and the placement of exploratory units; according to this
criterion, we can distinguish:
1) systems with two long dimensions and one short dimension ? vein systems and veinlike
(tabular) systems, which are connected with faults, and stratum systems and stratumlike
systems, which are not related to faults;
2) systems with two short dimensions and one long dimension ? columns and pipes, which
are related to fractures; ribbons and rolls, which are not related to fractures;
3) systems which are isomeric ? stock systems;
4) systems which are transitional between the first and third types ? lenticular systems.
Morphological groupings of this kind are useful but insufficient for searches to find
the optimal solutions of survey problems, since they are constructed according to a purely
geometric principle, not taking account of the genesis of the shape and internal structure
of the objects under study. The lack of a differentiated approach to objects which, although
belonging to the same morphological group, differ in their morphogenetic properties may lead
to serious misunderstandings. Accordingly, uranium ore systems and ore systems in general
should evidently be classified first of all on the basis of combination of natural factors
Translated from Atomnaya Energiya, Vol. 42, No. 6, pp. 443-451, June, 1977. Original
article submitted December 1, 1976.
This material is protected by copyright registered in the name of Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y.
10011. No part of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic,
mechanical, photocopying, microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is
available from the publisher for $7.50.
503
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TABLE 1. Uranium Ore Systems
Type of ore fields
Morphological type and spatial position
deposits
ore beds
ore bodies
Litera-
ture
1. Stratum fields,
in troughs of the
platform mantle,
composed mainly of
nonmetamorphic and
nondislocated rocks
of shelf facies of
epicontinental
seas
2. Stratumlike
fields, in erosion
basins superimposed
on a crystalline
base and filled
chiefly with ter-
rigenous rocks
3. Stratumlike
fields, in gas-
oil cupolas com-
posed of unmeta-
morphized rocks
of marine and
continental
origin
504
1.1. Stratum de-
posits, in black
shales
1.2. Stratum de-
posits, in dark
clays with bone
remnants of
fossil fishes
1.3. Stratum de-
posits, in phos-
phorite-bearing
sandstones
2.1. Stratumlike
deposits, in un-
metamorphosed
gray bituminous-
carbonate carbon-
iferous layers,
contrasting in
permeability
1.1.1.Stratum beds,
bounded by the ero-
sion lines of the
productive stratum
1.1.2. The same
1.1.3. The same
2.1.1. Stratumlike
beds, in permeable
rocks between water-
resistant rocks
2.1.2. Ribbon-
shaped beds, in
paleochannels
2.1.3. Roll-shaped
beds, on sandy-
argillaceous
contacts
2.2. Stratumlike 2.2.1. Ribbon-
deposits, inmeta- shaped beds, in
morphosed conglo- paleochannels
me rates
3.1. Isomeric de-
posits, in vari-
egated and gray
layers (sometimes
limestones),
crumpled in folds
and separated by
faults
.3.1.1 .Stratumlike
beds, in permeable
rocks between water-
resistant rocks
3.1.2.Veinlike
beds, in frac-
tures intersect-
ing the produc-
tive stratum
The ore bodies
are not indi-
vidually dis-
tinguished
2.1.1.1.Stratum-
like and lenti-
cular bodies con-
nected with local
accumulations of
organic matter
2.1.2.1.The same
2.1.3.1. Roll-
shaped and lenti-
cular bodies on
the boundaries
of stratum-oxida-
tion zones
2.2.1.1.Stratum-
like and lenti-
cular bodies in
deeper parts of
channel bottom
3.1.1.1.Stratum-
like and lenti-
cular bodies,
spatially con-
nected with local
accumulations of
bitumens
3.1.1.2. Saddle-
shaped veins in
layering fissures
in the crowns of
anticlines
3.1.2.1. Column-
shaped and pipe-
shaped bodies, in
cave-in funnels
and fissure-inter-
section structures
[2-5,
32]
f3, 8-12,
33, 34]
[8, 9,
19-22,
35]
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TABLE 1. (Continued)
Type of fields
Morphological type and spatial position
deposits
ore beds
ore bodies
Litera-
ture
4. Isomeric fields,
in volcanotectonic
depressions filled
with heterogeneous
sedimentary-effusive
layers
5. Lenticular
fields, in zones
of tectonic con-
tacts of grani-
toid massifs
6. Isomeric
fields, in com-
plex nodes of
faulted-folded
deformations of
sedimentary-meta-
morphic rocks
(with concordant
ore systems)
4.1. Isomeric de-
posits, on the inter-
sections of frac-
tures oriented in
different direc-
tions
4.2. Column-shaped
deposits, in
wedge-shaped blocks
between converging
fractures
4.3.Veinlike de-
posits, in dikes
of acid rocks
5.1.Lenticular de-
posits, in systems
of step faults
which intersect
granitoids and
contact-metamorph4
Osed sedimentary
rocks
6.1. Tabular de-
posits, in cen-.
? troclinal
closures or com-
plications of
the wings of
principal struc-
tures
4.1.1. Stratum-
like beds, of the
stratiform-stock-
work type, in
brittle rocks
separated by inter-
layers of inter-
formation clays
4.1.2. Veinlike
beds, in inter-
secting fractures
4.2.1. Irregularly
shaped blocks
bounded by fissures
of different orders
and different direc-
tions
4.3.1.Irregularly
shaped stockworks
at the bends of
dikes
5.1.1.Veinlike
beds, in bundles
of branching fis-
sures on the con-
jugations of
interlayer frac-
tures and inter-
secting fractures
6.1.1.Tabular beds,
in interstratum
stripping areas
and zones of fine
fissures on the
boundary of
rocks subject to
brittle and
plastic deforma-
tion
4.1.1.1. Vein-
like bodies,
lenticular and
irregularly
shaped stock-
works control-
ling fissures
4.1.2.1. Same
4.2.1.1.Lenti-
cular bodies, in
tectonic contacts
of rocks with dif-
ferent mechanical
properties
4.3.1.1.Lenticular
bodies in load
fissures
4.3.1.2.Column-
shaped bodies, in
tectonic contacts
of dikes with
surrounding rocks
5.1.1.1.Columnar
bodies, on the
intersections of
steep fissures of
interlayers of
"favorable" rocks,
on the conjugations
of main and connect-
ing fissures, etc.
5.1.1.2.Veinlike
and lenticular
bodies under shield-
ing surfaces and in
"structural traps"
on the conjugations
of differently
oriented fissures,
etc.
6.1.1.1:Column-
shaped bodies, at
the flexural bends
of productive strata
6.1.1.2.Stratum
bodies and saddle-
shaped veins, in
layering fissures
on the crowns of
small folds
[7, 15,
23-28]
[3, 5,
10, 25]
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[26,
30]
505
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TABLE 1. (Continued)
Type of fields
Morphological type and spatial position
Literature
deposits
ore beds
ore bodies
7. Isomeric
7.1. Isomeric de-
7.1.1. Vein beds, in
7.1.1.1.Columar
fields, in com-
posits, on up-
break-off fissures
and veinlike
plex nodes of
lifted portions .
oriented normally
bodies, under
fault-fold de-
of anticlinal
to axial planes
shielding planes
formations of
sedimentary-meta-
morphic rocks
crests
,the
and in structural
traps formed by
conjugation of
(with nonconcordant
ore systems)
the fissures
[31]
8. Tabular
8.1. Tabular de-
8.1.1.Columnar de-
8.1.1.1.Columnar
fields, in ex-
posits, in dis-
posits, in detri-
and veinlike bodies
tended frac-
torted faults
tus zones along
in alkaline metaso-
tures of the
basement of
ancient platforms
the boundaries
of faults
matites
[26, 34,
35]
8.2. Columnar de-
posits at con-
jugations of
faults with in-
cipient fractures
8.2.1. Same, in
systems of parallel
fractures
8.2.1.1. Same
which are responsible not only for the geometry of ore objects but also for their dimensions,
their position in space, and the variability and dispersion of the mineralization.
These combinations are not constant but depend on the nature of the geotectonic units
in which the uranium mineralization is situated. The largest uranium ore fields known thus
far are situated: in the troughs of the platform mantle, in erosion basins superimposed on
a crystalline base, in disintegrating gas-oil structures, in volcanotectonic depressions,
in zones of tectonic contacts of granitoid massifs, in complex nodes of fault-fold deforma-
tions (two modifications), and in extensive fractures of the basement of old platforms
(Table 1).
The types of ore fields (see Table 1) are distinguished by the nature of the ore control.
In the first and second types the uranium mineralization is essentially subject to stratifica-
tion control and lithological-facies control. In the third type, for the first time, a
significant role in the ore distribution is also played by fissural structures which are not
usually accompanied by any significant displacement of the walls and, as a rule, do not go
beyond the limits of the productive bands. In the fourth type there are large fractures in
stratified rock masses, which are contrasting in their physical and mechanical properties and
therefore react in different ways to tectonic stresses. In the fifth type the position of
the ore objects is due to a complicated set of natural factors ? systems of step faults, con-
jugation structures of layered and intersecting fractures of different orders, the composi-
tion of the surrounding rocks, etc. In the sixth and seventh types the mineralization is
controlled by a combination of folds and fault deformations genetically related to folding.
In the eighth types the most important factors of the ore control are regional fractures,
within which the localization of the ore object is determined by the combined influence of
the structural properties (fissures at bends) and lithological properties (zones of preore
metasomatosis) of the rocks surrounding the ores.
Within each ore field the complex of ore-control conditions depends on the position of
the ore object in the general hierarchical structure. In the great majority of cases it
becomes more complicated as we go from higher to lower, levels, and it is by no means always
possible to establish which of these conditions are the most important and which are secondary.
For ease of inclusion of ore objects in surveys, in some cases it is desirable to dis-
tinguish: a) ore fields in which the ore system at all levels of the hierarchical structure
belong to the same morphological group and have approximately the same spatial orientation;
506
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b) ore fields in which the included ore systems belong to different morphological groups and
have different spatial orientations. The first will be called "ore fields with concordant
ore systems," and the second "ore fields with nonconcordant ore systems."
The successive stages of geological survey work on ore fields of the first group differ
from one another chiefly in the density of the observation networks. The survey method
and the orientation and geometry of the survey network usually remain unchanged. In ore
fields of the second group, each successive stage of geological survey work differs from the
preceding one not only in the density of the survey network but also in its geometry and
orientation, depending on the country in which the work is done.
Ore Fields in Troughs of the Platform Mantle. The amount of uranium contained in such
fields is related to the undislocated or only slightly dislocated strata of sedimentary
rocks, which consists primarily of shelf facies of epicontinental marine basins and extend
over areas measured in tens of thousands (or sometimes hundreds of thousands) of square
kilometers.
The mineralization is usually of the single-stage or two-stage type, lean, very uniform,
and practically continuous. The continuity of the ore-bearing horizons is sometimes broken
by erosion areas, which determine the boundaries of the individual deposits or ore beds.
More often, however, the concepts of deposits coincides with the concepts of ore beds and
ore bodies.
The most characteristic representatives of ore systems of this type are uranium-bearing
black shales, phosphorites, dark clays with bone remnants of fossil fishes, and similar forma-
tions, in which the uranium has been deposited during the process of sedimentogenesis and
was not subjected to any substantial redistribution as the sediment developed. These have
been described in many studies [2-6].
Survey work in such systems is relatively simple. Drilling at the points of a rectangular
network and gamma logging of the boreholes yield completely reliable information concerning
the boundaries of the productive band and the amounts of useful component contained in it.
The role of main-line intersections is played by lines of boreholes which can be oriented in
any direction. The study of the ore fields is completed at the stage of detailed surveying.
There is usually no need for operational surveying.
Ore Fields in Erosion Basins Superimposed on a Crystalline Base. This is one of the
most widespread types of ore bodies. Its most characteristic features are: localization
near input areas of water-pressure systems, confinement to terrigenous rocks, multistage
structure, sharply fragmented mineralization, and a multiplicity of factors controlling the
position of deposits, ore beds, and
In most cases, the ore systems
direction of an underground stream.
ore bodies.
are represented by stratiform formations extending in the
The deposits are arranged either in gray, bituminous-
carbonate, coal-bearing and multicolored layers of contrasting permeability (type 2.1) or in
ancient metamorphosed conglomerates (type 2.2).
Among the ore beds forming deposits of type 2.1 we may distinguish, depending on the
morphogenesis conditions, the following: stratified, in permeable rocks between water-re-
sistant rocks (type 2.1.1); ribbon-shaped, in paleochannels (type 2.1.2); roll-shaped, on
sandy-argillaceous contacts (type 2.1.3). The ore bodies in the deposits may be stratified
(type 2.1.1.1 and 2.1.2.1) or roll-shaped (type 2.1.3.1). In the first case their position
is controlled by local accumulations of organic matter, and in the second case by the boundar-
ies of the zones of stratum oxidation.
The type described above includes ore systems in which the uranium concentration is re-
presented chiefly by the products of the activity of groundwater circulating in the boundary
region of artesian basins [1, 3, 5, 7-12]. Their dimensions vary over a wide range. The
areas occupied by the deposits sometimes reach several hundred square kilometers, and the
areas of the ore beds are often measured in square kilometers. The distance between beds
within deposits and between ore bodies within beds are highly diverse. The ore-content coef-
ficient* varies from 0.2 to 0.5 for deposits, from 0.4 to 0.6 for ore beds, and from 0.7 to
*This is the ratio of the areas (or volumes) occupied by all the various systems of one
level to the area (or volume) of the higher-level ore system enclosing them.
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0.9 LVL ULU UULLIUS. ine variaoility or Ene mineralization, expressed by the absolute-contrast
coefficient proposed by J. Matheron, is usually in the 0.03-0.128 range.
This group of ore systems includes uranium-bearing conglomerates similar to those known
in the Republic of South Africa (Witwatersrand) and in Canada (Blind River). Morphologically
these are similar to the above-described ore systems, although they differ from them in their
mechanism of formation and in the nature of the intrasystem relationships [3, 5, 10, 13-171.
Survey work is carried on chiefly by means of boreholes, with lines oriented across the
direction of an underground stream. Complications in the surveying process sometimes arise
as a result of a breakdown of the radioactive equilibrium and the consequent need for special
investigations to determine the correction coefficients to be applied to the metal-content
values calculated on the basis of the gamma logging of the boreholes.
Ore fields with disintegrating gas-oil structures are spatially related to eroded anti-
clines and gas-and-oil-bearing cupolas composed of terrigenous red and gray layers (sometimes
limestones), which served in the past as foci for relieving the stresses on ancient water-
pressure systems.
The ore fields consist of a set of deposits made up of ore beds of two morphological
types: stratified, chief metasomatic deposits, controlled by certain horizons or interlayers
of "favorable" rocks (type 3.1.1), and veinlike beds confined to fractures intersecting the
productive layers but not going beyond the limits of the productive bands (type 3.1.2). In
the stratified beds, individual ore bodies of stratified or lenticular form are confined to
local accumulations of bitumens and oriented according to the layering (type 3.1.1.1), while
in the veinlike beds they often have the shape of pipes or columns filling cave-in funnels
or inclined along the curves of intersection of two fissures (type 3.1.2.1). In the struc-
ture of stratified beds we sometimes also find saddle-shaped veins (type 3.1.1.2), whose
shapes are determined either by exfoliation fissures in the crowns of small anticlines or by
combination of these fissures with intersecting faults.
This group includes ore fields within whose limits the uranium concentration consists
chiefly of products of the activity of rising solutions circulating in a region of attenuated
destressing of artesian basins [8, 9, 18-19].
Similar to these ore fields in their morphological properties are some ore systems o
disputed genesis in black shales [20-22].
The ore-content coefficient is 0.01 for deposits, 0.3-0.5 for ore beds, and 0.7-0.8 for
ore bodies. The absolute-contrast coefficient for various ore bodies ranges from 0.18 to
0 54.
The discovery of evaluation of such ore objects is a rather complicated process. In
most cases the drilling of vertical boreholes from the surface makes it possible to obtain
the information necessary and sufficient for drawing the contours of the deposits and clarify-
ing the general picture of the distribution of ore beds in them, i.e., for solving the pro-
blems of evaluation at the preliminary-survey stage. The system of detailed surveying de-
pends on the morphological properties of the mineralization. For stratified beds this, means
a combination of horizontal and ascending operations; in the case of veinlike deposits it
means drifts incombinationwithcross drifts and ascending operations. Operational surveying, as
a rule, consists in the drilling of chamber boreholes arranged in a fan pattern, at various
angles to the horizon.
Ore Fields in Volcanotectonic Depressions. The mineralization develops in heterogeneous
sedimentary-effusive layers, rocks of funnel-type and subeffusive facies, sometimes in the
crystalline basement. The deposits are largely controlled by fault conjugation or intersec-
tion structures. Some of them ? the isomeric deposits ? are combinations of stratified beds
of the stratiform-stockwork type with veinlike beds, sometimes going beyond the boundaries
of the productive horizons (type 4.1), those of a second group have the shape of columns or
wedges bounded by converging faults and including ore beds of different and usually very com-
plicated configuration (type 4.2), and a third group consists of mineralized dikes of acid
rocks of subeffusive facies, at the bends of which there are ore beds having the form of
stockworks of irregular shape (type 4.3).
The complexes of geological-structure elements controlling the position of ore beds and
ore bodies in deposits of all three varieties are highly diverse.
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The boundaries of the stratiform stockworks (type 4.1.1) are chiefly due to the physico-
mechanical properties of the rocks surrounding the ore, reacting in a certain way to tectonic
stresses, and interlayers of interformational clays, which play the role of shields for the
mineralization. It is difficult to distinguish the ore bodies within these. Large mineral-
ized segments within the beds are connected to one another by individual fissures or boundles
of fissures which are mineralized to different degrees. Under such conditions, it is more
correct to speak not of ore bodies but of certain structurally homogeneous blocks which differ
from one another in the shape and character of the ore-controlling elements.
For the most part, these are stockworks of lenticular, veinlike, or irregular shape on
segments of conjugations of fissures of different orders and different directions (types
4.1.1.1 and 4.1.1.2).
In deposits of columnar form the ore beds are most often in the form of irregular blocks
bounded by fissures of second and higher orders (type 4.2.1), while the ore bodies in the
beds are stockworks, also of irregular, sometimes veinlike, shape, controlled by intrusive
dikes or local areas of granulation at the contacts of rocks with different mechanical proper-
ties (type 4.2.1.1).
In deposits connected with dikes of acid rocks, stockwork ore bodies of lenticular and
columnar shape are confined, respectively, to systems of horizontal fissures and tectonic
contacts of dikes with surrounding rocks (type 4.3.1.1 and 4.3.1.2).
Ore fields in volcanotectonic depressions are characterized in many cases by very im-
pressive dimensions. Their cross-sectional areas may be as high as several hundred square
kilometers. Individual deposits stretch out over many kilometers, and the areas of ore
bodies are measured in thousands of square meters, with the depth of mineralization reaching
1500-2000 m. The ore-content coefficients vary from 0.1 to 0.2 for deposits, from 0.3 to
0.5 for ore beds, and from 0.6 to 0.8 for ore bodies. The absolute-contrast coefficient is
usually in the 0.2-0.3 range.
For ore fields in volcanotectonic depressions the most characteristic deposits are those
of uranium-molybdenum formation, a description of which may be found in many published
studies [7, 20-29]. An evaluation of these requires the use of highly diversified methods.
The boundaries of the ore field and the position of individual deposits in it are usually
established by drilling boreholes in the surrounding complexes at several levels. The sub-
sequent stages of the survey work consist chiefly in carrying out mining operations - drifts-
in combinationwithcrossdriftsintersectingthe entirethicknessof the ore-bearing formation.
The resulting system of horizontal and vertical cross sections makes it possible to establish
fairly easily the boundaries of the mineralized areas or to draw roughly the contours of the
ore deposits. The problem of evaluating individual ore bodies is usually solved by means of
chamber drilling, in fans inclined at various angles to the horizon.
Ore Fields in Zones of Tectonic Contacts of Granitoid Massifs. The deposits (type 5.1)
are confined to systems of step faults intersecting a sedimentary-metamorphic layer and
granitoid massifs. The ore beds (type 5.1.4) consist of bundles of branching veins oriented
parallel to these faults. As a rule, they are localized in areas of conjugation of layer-by-
layer and intersecting faults. In surveys and operations involving such deposits, the term
"ore body" is taken to mean the set of plane ore lenses found in one vein and controlled by
a large number of higher diverse factors. An important role is played by lithological con-
trol, under the influence of which the ore bodies often take the shape of columns inclined
along the dip of a stratum or interlayer of "favorable" rocks (type 5.1.1.1). In addition,
the geometry of the ore bodies and their position in space are substantially affected by
various kinds of screening surfaces formed by a combination of fractures following the veins
in various directions, tectonic contacts of rocks with contrasting physicomechanical proper-
ties, intrusive dikes, etc.
Ore fields of this type are particularly characteristic of five-metal formations (uranium-
nickel -cobalt-bismuth-silver), which have been described many times in many investigations
[3, 5, 10, 22]. In some cases they extend over tens of kilometers, with a width of several
hundred meters. In these narrow bands the deposits are arranged in segments measurable in
kilometers. The length of individual deposits does not exceed 200 or 300 m, reaching 1-2 km
only in isolated cases. The depth of mineralization sometimes reaches 2 km, and the "critical
horizon" is usually a surface of granite massifs underlying the productive sedimentary-meta-
morphic layer.
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because or the high degree ot tragmentation and variability of the mineralization, an
evaluation of all the ore systems is possible only on the basis of data obtained by carrying
Out, testing, and mapping mining operations. At the preliminary-survey stage these consist
of major cross-cuts oriented along the contact, i.e., normal to the predominant course of the
ore veins; in detailed surveying it consists of drifts in combination with ascending cleaving
veins on rectangular blocks; at the operational-survey stage it consists of undercut opera-
tions and chamber-drilled boreholes used for the fixation of vein apophyses.
Ore Fields at Complex Nodes of Fault-Told Deformations of Sedimentary-Metamorphic Rocks.
Among these it is expedient to distinguish two types: those with concordant ore systems
(type 6) and those with nonconcordant ore systems (type 7).
In ore fields of the sixth type the deposits are confined either to centroclinal closures
or to complications of the wings of the main structures (type 6.1); the ore beds in them,
usually stratified or veinlike in shape, show a clear connection with interstratum breakoffs
and zones of fine fissures at the contacts between rocks with different mechanical properties
(type 6.1.1). The ore bodies have the shape of columns (type 6.1.1.1), stratified or saddle-
shaped veins (type 6.1.1.2). The columns are most often connected with flexural bend and the
veins with folds of higher orders. The ore-controlling structures are systems of fine fis-
sures arising at the boundary of rocks which have undergone brittle and plastic deformation.
The folded structures of ore fields are often complicated by series of preore fractures,
but in the process of ore formation the predominant role is played not by the filling of open
cavities but by the selective displacement of the material of the surrounding medium.
The most characteristic representatives of ore fields of the above-described type are
those in iron-ore formations of the Precambrian [26, 30]. The largest of these stretch out
over 10-12 km, with a width of 3-4 km and a depth of 1.0-1.5 km.
The deposits themselves, however, occupy no more than 10-15% of this space. They are
represented chiefly by collections of ore beds oriented along the stratification of the
surrounding rocks, parallel to one another. The distances between the deposits are usually
measured in a few tens of meters, while the ore bodies in the beds are arranged fairly com-
pactly, separated from one another by segments of practically ore-free rocks no more than a
few meters wide.
The area ore-content coefficient is 0.4-0.6 for deposits, 0.8-0.9 for ore zones, and
0.9-1.0 for ore bodies. The absolute-contrast coefficient is 0.06-0.08.
For a general evaluation of such deposits at the preliminary-survey stage, boreholes
along curves oriented across the folded structures have been successfully used. The evalua-
tion of individual beds in detailed surveying is carried out exclusively by mining operations ?
drifts in combination with cross drifts intersecting the entire thickness of theore-bearingforma-
tion. At the stage of operational surveying, ascending operations and chamber-drilled bore-
holes are also used.
The seventh type includes ore fields which are also confined to folded structures but
differ in the fact that the deposits in them are represented by bundles of subparallel veins
which exhibit a clear connection with areas of uplift of fold crests (type 7.1). In such
situations these veins can be regarded as ore beds (type 7.1.1). Usually they fill fissures
oriented normally to the axial planes of the anticlines and apparently arise primarily as
breakoff fissures. The term "ore body" in such beds means vein areas containing uranium
minerals in a quantity justifying the expense of extraction and processing of the vein mass.
Their position is determined by the curves of conjugation with incipient fissures or by the
shielding influence of rocks superimposed on the productive formation (type 7.1.1.1).
The length of one such ore field [31] is 10 km, with the distances between individual
deposits measured in kilometers and the distances between individual ore beds measured in
hundreds of meters. The productive veins extend for many hundreds of meters horizontally
and vertically. The ore bodies in the beds are separated from one another by ore-free masses
tens of meters thick. The areas of the ore bodies vary from several tens to several hundreds
of square meters.
Underground mining operations constitute the only source of the information required
for evaluating the parameters of such ore systems. At the preliminary-survey stage
these are chiefly major cross-cuts oriented along the axis of an anticline; at the
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detailed-survey stage they are multiple-level dritts witft ascenaing ana cleaving ULC UeUb
on rectangular blocks; at the operational-survey stage they are undercut drifts and chamber
boreholes arranged in a fan, at various angles to the horizon.
Ore Fields in Zones of Regional Faults of the Old Basement. The deposits in these
are represented by mineralized blocks of rocks included in the composition of tectonic zones
and confined either to distortions of a fracture or to the curves of conjugation of this
fracture with incipient fractures of second and higher orders. In the first case they are
tabular in form (type 8.1); in the second case they take the form of columns inclined along
the line of intersection of two planes (type 8.2). In tabular deposits the ore beds extend
along a limited fracture as chains of flat lenses arranged in coulisse form and exhibiting
a clear connection with alkaline metasomatites (type 8.1.1). In columnar deposits the con-
cept of ore bed is most closely met by bundles of veinlike ore bodies or collections of
lenticular ore bodies parallel to the main structure (type 8.2.1). Ore fields of this type
are most characteristic of areas of activation of Precambrian shields and platforms [4, 14,
15].
The length of the ore-bearing faults measures many kilometers, sometimes tens of kilo-
meters; the deposits extend for several kilometers, the ore beds for hundreds of meters, and
the ore bodies in the beds for tens of meters. The depth of mineralization amounts to 2-2.5
km. The ore-content coefficient is 0.4-0.5 for deposits, 0.7-0.9 for ore beds, and nearly
unity for ore bodies. The absolute-contrast coefficient is 0.08-0.10.
CONCLUSIONS
A systematic approach to the analysis of the natural phenomena involved in the genesis
of forms of uranium ore objects enables us to distinguish among them eight fundamental types
of ore fields, in which we find at least 14 morphogenetic types of deposits, 18 types of ore
beds, and 20 types of ore bodies. Each types arises as a result of a completely definite
combination of external-environment conditions which are characteristic of this type alone.
The proposed grouping does not reflect all of the diverse forms of uranium ore systems.
It should be regarded only as a first step on the difficult path of establishing a morpho-
genetic classification of natural geological objects which is necessary for solving a number
of methodological problems arising in the process of discovering and surveying deposits of
minerals.
The author wishes to express his gratitude to A. N. Eremeev, V. A. Petrov, and S. A.
Deinege for their valuable comments.
LITERATURE CITED
1. D. Ya. Surazhskii, Soy. Geol., No. 2, 3 (1974).
2. M. N. Althausen, in: Metals in Sedimentary Rocks [Russian translation], Vol. 3, Moscow
(1966), p. 102.
3. M. M. Konstantinov and E. Ya. Kulikova, Uranium Provinces [in Russian], Atomizdat, Moscow
(1960).
4. V. McKelvey, D. Everhart, and R. Garrels, in: Problems of Ore Deposits [Russian trans-
lation], Izd. Inostr. Lit., Moscow (1959), p. 428.
5. E. Heinrich, Mineralogy and Geology of Radioactive Mineral Raw Materials [Russian trans-
lation], Izd. Inostr. Lit., Moscow (1962).
6. V. McKelvey and S. Nelson, Econ. Geol., 45, No. 1, 35 (1950).
7. F. I. Vol'fson (editor), Geology and Problems of the Genesis of Endogenous Uranium
Deposits [in Russian], Nauka, Moscow (1968).
8. L. S. Evseeva et al., in: Problems of Applied Radiogeology [in Russian], Atomizdat,
Moscow (1967), p. 326.
9. L. S. Evseeva et al., Geochemistry of Uranium in the Zone of Hypergenesis [in Russian],
Atomizdat, Moscow (1974).
10. D. Ya. Surazhskii, Methods of Discovering and Surveying Uranium Deposits [in Russian],
Atomizdat, Moscow (1960).
11. W. Finch and S. Warren, Geol. Sur. Prof. Papers, No. 538 (1967).
12. D. Shaw and H. Granger, Econ. Geol., 60, No. 2, 240 (1965).
13. U. Libenberg, in: Proceedings of the Second Geneva Conference, 1958. Reports by Foreign
Scientists [Russian translation], Vol. 8, Atomizdat, Moscow (1959), p. 377.
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14. Davidson, Mines Mag., 88, 73 (1953).
15. Davidson, Econ. Geol., 52, No. 6, 668 (1957).
16. Holmes, Econ. Geol., 50, No. 1, 751 (1955).
17. Roscoe and H. Steacy, First International Conference, Geneva (1955), Rep. 224.
18. I. Zubov, Geol. Rudn. Mestorozh., No. 5, 6 (1960).
19. Russel, in: Proceedings of the First Geneva Conference. Reports of Foreign Scientists
[Russian translation], Vol. 8, Atomizdat, Moscow (1959), p. 81.
20. R. V. Getseva, in: Problems of Uranium Geology [in Russian], Atomizdat, Moscow (1957),
p. 20.
21. V. A. Krupennikov, Geol. Rudn. Mestorozhd., No. 4, 74 (1969).
22. B. L. Rybalov, ibid., No. 2, 3 (1965).
23. V. P. Vlasov et al., Geology of Deposits of Uranium-Molybdenum Formation [in Russian],
Atomizdat, Moscow (1966).
24. F. I. Volifson et al., Izv. Akad. Nauk SSSR. Ser. Geol., No. 11, 114 (1967).
25. D. I. Shcherbakov (editor), Geology of Hydrothermal Uranium Deposits [in Russian], Nauka,
Moscow (1966).
26. V. I. Kazanskii and N. P. Laverov, in: Ore Deposits of the USSR [in Russian], Vol. 2,
Nedra, Moscow (1974), p. 319.
27. A. B. Kazhdan et al., in: Problems of Applied Radiogeology [in Russian], No. 2, Atomizdat,
7.
, Tr. IGEM, No. 2, Izd. Akad. Nauk SSSR, Moscow (1962), p. 116.
, Geol. Rudn. Mestorozhd., No. 6, 38 (1964).
Uranium Deposits in Iron-Ore Formations of the Precambrian [in
Moscow (1969).
., in: Uranium Deposits: Zonality and Parageneses [in Russian],
Atomizdat, Moscow (1970), P. 93.
32. I. U. Aizeksen, in: Proceedings of the First Geneva Conference [in Russian], Vol. 6,
Gosgeoltekhizdat, Moscow (1956), p. 413.
33. S. T. Batulin et al., Exogenous Epigenetic Uranium Deposits [in Russian], Atomizdat,
Moscow (1965).
34. T. V. Balibina et al., Geol. Rudn. Mestorozhd., No. 5, 67 (1963).
35. V. I. Kazanskii et al., ibid., No. 1, 3 (1968).
C.
C.
S.
P.
A.
R.
Moscow (1967), p. 28
28. N. P. Laverov et al.
29. N. P. Laverov et al.
30. R. P. Petrov et al.,
Russian], Atomizdat,
31. A. V. Zavarzin et al
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LOW-ADDITIVE URANIUM ALLOYS FOR THE FUEL ELEMENTS OF THE KS-150 REACTOR
V. F. Zelenskii, A. I. Stukalov, UDC 669.822'5:621.785.784/786
V. P. Ashikhmin, and A. V. Azarenko
The cores of the fuel elements used today in the KS-150 reactor are made of uranium of
reactor purity with small admixtures of iron and silicon [(1-2).10-2 and (2-3).10-2% by mass,
respectively]. Earlier, it was shown [1, 2] that the use of these admixtures in such a
proportion promotes the formation, under (3 hardening, of uranium rods with a weak axial
texture of the [100] type. Loop radiation tests showed that such a texture ensures minimal
change of shape in the fuel elements as a result of radiation growth in the 200-350?C tempera-
ture range. In fuel-element segments operating at higher temperatures (350-500?C) the radia-
tion deformation develops chiefly as a result of swelling of the uranium, under loop-test
conditions close to the design conditions of the KS-150 reactor (gas-coolant pressure 607.95
kPa, temperature 400-500?C, specific thermal stress 55 kW/kg) it reaches 15-20% by volume
for a burnup of '\,10,000 MW.day/ton. Local variation in the geometric dimensions of the
cores, resulting from this amount of swelling, may lead to rupture of the jacket and break-
down of the fuel element. Therefore, the limitation of fuel swelling is one of the effective
methods of increasing fuel lifetime at high burnup values.
A method widely used for reducing the swelling of uranium is alloying it with metallic
and nonmetallic admixtures [3, 4]. In the present case the admixtures most acceptable from
the standpoint of minimal parasitic capture of neutrons were found to be aluminum and chromium.
In the present paper we discuss the main results of a metal-physics investigation of
alloys with admixtures of aluminum and chromium (0.1 and 0.2% by mass, respectively) and
radiation tests of fuel elements of the KS type of cores made of these alloys.
Material. The alloys were prepared by the method of vacuum remelting of compact uranium
with aluminum, and also with aluminum and chromium in powdered or shaving form. The amount
of the additives in the base material corresponded to the chemical composition of KS-type
fuel-element cores. Cylindrical castings 40 mm in diameter and having a mass of 2.5 kg were
machined on a lathe and were subsequently used for pressing 6.5-mm diameter rods at 950?C;
the rods were chilled in a vacuum.
Hot Working. The hardening of the pressed rods from the y phase included inductive
heating to 900-950?C, maintenance at this temperature for 10-15 sec, and transverse hardening
in a water shower; these operations were carried out in air, with translational motion of the
rod in the inductor of a high-frequency generator and in a cooler. The hardened rods were
annealed at 550?C for various lengths of time and then subjected to hardening from the 6
phase. This kind of hot working ensured uniform separation of the finely dispersed inclusions
of the second phase of the uranium matrix (UA12, U2Si, U6Fe), and also promoted the formation
of a fine-grained structure (when the time of intermediate annealing in the a phase was 3-5
h) [5, 6].
Results of the Metal-Physics Investigation
Structure. The macrostructure of the y-pressed alloys is essentially large-grained,
with the diameter of the largest grains reaching 1 mm (Fig. la). The intermetallic compounds
and carbides form a granular substructure with discrete (discontinuous) boundaries (Fig. lb).
Such a substructure is typical of low-additive uranium alloys cooled slowly from the y phase.
The density of separation of the second phase from the intermetallic and carbide inclusions
measuring 0.5-1 p was calculated by the well-known method of [7] and found to be 6.109 par-
ticles/cm3. Such a structure is unacceptable for fuel-element cores because of the large
Translated from Atomnaya Energiya, Vol. 42, No. 6, pp. 452-456, June, 1977. Original
article submitted December 22, 1975; revision submitted July 5, 1976.
This material is protected by copyright registered in the name of Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y.
10011. No part of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic,
mechanical, photocopying, microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is
available from the publisher for $7.50.
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Fig. 1. Macrostructure (a) and granular substructure (b) of y pressed.
rods (x8 and x1000 , respectively).
grain size and the coagulated separations of the second phase. Hardening of the pressed
rods from the 13 phase does not lead to any substantial refining of the structure or finer
dispersion of the intermetallides. Therefore, we used hardening of the alloys from the y
phase. The structure of the y-hardened alloys is very fine-grained (Fig. 2a), and the sub-
structure is characterized by a clearly formed dislocation-admixture network (Fig. 2b). The
boundaries of the network cells are continuous lines which, as assumed by the authors of [8,
9], are the boundaries of subgrains of the R phase formed during rapid conversion from y to
13 and decorated with admixture atoms in the preseparation state.
The above-described structure is substantially different from the structure of y-pressed
alloys and is acceptable in principle for fuel-element cores. However, further investigations
showed that the distribution of the texture along the y-hardened rods was not uniform. This
leads to warping of the fuel elements under irradiation and thereby increases the danger of
breakdown.
A further improvement of the alloy structure can be obtained by using intermediate
annealing in the high-temperature region of the a phase and subsequent 0 hardening. Annealing
of y-hardened alloys leads to dissociation of the supersaturated solid solution and more
uniform separation of the UA12 intermetallides, both in the matrix and along the subgrain
boundaries [10]. Figure 3 shows the structure of uranium-aluminium alloys annealed after
y hardening. It can be seen that the cells of the subgrains are formed by particles of urani-
um intermetallides and by admixtures which were present in the y-hardened alloy in the pre-
separation state; within the cells we also noticed a large number of second-phase separations.
The density of 0.5-1 i separations is 6.10" particles/cm'. Thus, y hardening and subsequent
annealing leads to a refining of the structure of the y-pressed metal, increases by a factor
of 100 the density of second-phase separations, and promotes a more uniform distribution of
the intermetallide particles. However, the annealing does not eliminate the nonuniformity of
the texture that arises in y hardening. The formation of the desired texture in the rods and
the uniformization of its distribution along the length, as well as the formation of a uniform
fine-grained structure, are promoted by the concluding operation of the 13 hardening. In
macrostructure and granular substructure the 0-hardened alloys differ little from those shown
in Fig. 3.
The separation of the admixtures from the solid solution during intermediate a annealing
and their subsequent coagulation apparently have a substantial effect on the formation of
the structure during the process of 0-to-a conversion, especially when the material is kept
in the 0 phase only for a short time. Some of the intermetallides which grew larger during
annealing apparently cannot dissolve in the 0 phase during the short time the alloy is kept
in that phase before hardening, and they limit the growth of the 0-uranium grains, which in
turn affect the formation of the a structure. The effect of intermediate a annealing on
grain size in the hardened alloy of uranium with aluminium and chromium is shown in Fig. 4.
The decrease in grain size as the annealing time increases to '1,3 h is easily explained by
the above considerations. The increase in the grain size of the 0-hardened uranium for
longer times of annealing in the a phase before 0 hardening is due to the coagulation of the
intermetallides and the decrease of the separation density.
Texture. The texture of the 0-hardened rods was determined by calculating the orienta-
tion parameters ? the growth indices Gx and Gap ? on the basis of roentgenographic investiga-
tions and also measurements of the coefficients of linear expansion a and electrical resis-
tivity p [11, 12].
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Fig. 2. Macrostructure (a) and granular substructure (b) of y-cooled
alloys of uranium (magnification as in Fig. 1).
Fig.. 3. Macrbstructure (a) and granular-substructure (b) of annealed
alloy .(same magnification).
300
250
0
200
150
100
50
0.10 JO 60 MOM 600 100.
AnneMingtime,rnin
Fig. 4. Grain size in 0-hardened
uranium alloy as a function of the
time. of intermediate annealing..-in.the
a phase before- 13-hardening.
I -I
It was found that in the 0-hardened alloy rods, -as in the case-of unalloyed uranium
rods, a [100]-type texture is formed along the longitudinal axis; how pronounced this texture
is for the selactedconditions of the 0-hardening depends on the duration of the intermediate
a annealing. (before the 0 hardening) (Fig. 5). It was noted,aboye that for the formation
:of a fine-grained-structure the optimal length of time for such annealing is 3-5 11.. In this
case, during the 0 hardening the alloy forms a texture with a value of 10-12% expressed in
terms? of the growth indices. It should-.be noted that the quantitative estimate of the
texture in uranium alloys may be inaccurate, since the Methods used were designed for deter-.
Mining the texture of unalloyed uranium. Nevertheless,-the,use of these methods in the pres-
ent case is completely justified, since it-enables us to investigate the nature of the varia-
tion in the hardening texture during the hot working of the alloys.
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100
80
60
bO40
61 20
10 30 60 180 300 600 1440
Annealing time, min
Fig. 5. Texture of hardened alloys
of uranium with aluminum and chromium
as a function of time of annealing at
550?C.
100
80
60
0
100
80
50
20
400 500 500 0 100
Temperature, ?C
200 300
400 500 600
Fig. 6. Mechanical characteristics of alloys of uranium with
aluminum and chromium (1), uranium with aluminum (3), and the
original uranium (2).
Mechanical Properties. The mechanical tests were conducted on a tensile testing machine
over a temperature range of 20-600?C. It follows from Fig. 6 that the alloy with aluminum
and chromium admixtures has better strength characteristics than other materials over the
entire range of investigated temperatures. The strength of the uranium-aluminum alloy can be
compared with that of the original uranium up to a temperature of 200?C; at higher temperatures
it is somewhat lower.
For all the materials, we observe a considerable growth in plastic characteristics as
the temperature increases to 100?C. At higher temperatures the plasticity of the uranium--
aluminum-chromium alloy remains unchanged, while for the uranium-aluminum alloy and the
original uranium we observe a decrease in the relative elongation. This type of variation
in the plastic properties of materials in this range can be explained [13] by the effect of
the temperature on the number of systems in which twinning takes place during the defomation
process; as the temperature increases (above 100?C), this number decreases for some alloys.
The minimum relative elongation, linked to the transition to grain-boundary deformation,
for the original uranium is found at a higher temperature than for the uranium-aluminium alloy
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22
20
18
8
6
4
2
0
300
400 500
Temperature, eC
600
Fig. 7. Swelling of uranium (A) and
uranium-aluminum-chromium alloy (0)
as functions of temperature.
and is practically absent in the case of the uranium-aluminium-chromium alloy The
minimum plasticity in the unalloyed uranium is caused [14] by the formation of cracks around
the inclusions and cavities along the grain boundaries. The absence of this minimum in the
uranium-aluminum-chromium alloy may be explained by the high density of the UA12 and chromium
separations which apparently strengthen the grain boundaries. An analogous phenomenon has
been observed for an alloy of uranium with beryllium oxide [14].
Radiation Tests. Fuel elements with cores made of the alloys investigated, together with
fuel elements whose cores had a chemical composition matching those used in the 16-150 reac-
tor, were tested to a maximum burnup value of 8040 MW.day/ton in the KS-60 loop under condi-
tions close to reactor conditions. The only significant difference found in the loop tests
was the large number of shutdowns (,120), 5-6 times the number found in reactor tests of
fuel elements at the same burnup value.
Evaluation of the Results. Figure 7 shows how the swelling of the uranium and the
uranium-aluminum-chromium alloy varies with temperature; the swelling was calculated from
measurements of the elongation and change in diameter in various segments of the fuel ele-
ments. It follows from Fig. 7 that the temperature-swelling curve for unalloyed uranium has
two clearly marked maxima: the first at 430?C and the second at 500?C. According to the
cavitation-swelling model proposed in [15], we may assume that the low-temperature peak in
the swelling is due to the appearance and development of voids (cavitation) along the boundar-
ies of the uranium matrix (grain boundaries, twins), and the appearance of the high-tempera-
ture peak is due to the homogeneously degenerating pores of vacancy origin. It can be seen
that the contribution of the low-temperature part to the total swelling constitutes the
largest portion, while the maximum swelling value at a temperature of 430?C is 1,20% in this
case (a diameter increase of '1,8%) and at 500?C the value is 10% (with 3% core thickening).
In fuel elements tested to the same burnup value in the KS-150 reactor under similar condi-
tions, a single swelling maximum was found at 500?C (%9%) [16].
As noted above, the reactor tests differ from the loop tests chiefly in the small number
of shutdowns (20-30 shutdowns during the tests up to burnup values of '1,8000 MW.day/ton).
Thus, a comparison of these data enables us to conclude that the development of the low-
temperature peak in the swelling under the loop-test conditions is due in large measure to
the presence of a large number of thermal cycles. The effect of the thermal cycles on the
increase in the low-temperature swelling may be attributed to the accelerated development
of cracks along the boundaries of the uranium matrix which result from sharp changes in the
stressed state of the fuel-element core. The agreement in temperature and magnitude between
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the high-temperature maxima indicates that sharp changes in temperature, within the investi-
gated ranges, do not have any substantial effect on the vacancy-swelling process.
It follows from Fig. 7 that the addition of aluminum and chromium to uranium will
eliminate the low-temperature swelling almost completely; this can be attributed to strengthen-
ing of the grain boundaries by the second-phase separations. The reduction in the amount of
swelling in the high-temperature range may be due, as assumed in [15], to the strengthening
of the process of recombination of vacancies and displaced atoms through limiting of the
finely dispersed separations in the migration of displaced atoms to dislocations and of
vacancies to voids.
CONCLUSION
The investigations showed a high resistance to swelling in an alloy of uranium with small
amounts of aluminium and chromium at a burnup value of 'NA_ at.%. The hot-working
scheme for the alloy, including y hardening of the solid solution, a annealing (aging), and
0 hardening, ensured the formation of a fine-grained isotropic structure with a finely dis-
persed distribution of second-phase particles, which facilitated the effective utilization of
the alloying additives in limiting the cavitation and vacancy swelling of the uranium.
The results obtained showed that an alloy of uranium with aluminum and chromium will
make it possible to obtain high burnup values in the KS-150 reactor.
LITERATURE CITED
1. V. F. Zelenskii et al., At. Energ., 39, No. 1, 24 (1975).
2. V. E. Ivanov et al., in: Start of Operation of the A-1 Power Station. Part I [in Russian],
Czechoslovak Atomic Energy Commission Press, Bratislava (1974), p. 224.
3. W. McDonell and E. Sturcken, Nucl. Technol.,, 26, No. 4, 420 (1975).
4. J. Lehman et al., in: Proceedings of the IAEA Symposium "Investigation of the Swelling
and Growth of Uranium Alloys under Irradiation," Vienna, June 2-6, 1969 p. 413.
of Uranium Alloys under Irradiation," Vienna, June 2-6, 1969 p. 413.
5. A. I. Stukalov et al., Byull. Izobret., No. 2, 131 (1975).*
6. A. I. Stukalov et al., Byull. Izobret., No. 2, 131 (1975).
7. D. Kramer, J. Nucl. Mater., 27, No. 3, 281 (1962).
8. C. Angerman and R. Huntoon, J. Less-Common Met., 9, No. 5, 338 (1965).
9. A. A. Bochvar et al., At. Energ., 27, 193 (1965).
10. A. Smith, J. Nucl. Mater., 27, No. 2, 194 (1968).
11. E. Sturcken and W. McDonell, J. Nucl. Mater., 7, No. 1, 85 (1962).
12. J. Stobo and B. Pawelski, J. Nucl. Mater., 4, No. 1, 109 (1961).
13. T. Khan, I. Brun, and I. Decours, J. Nucl., Mater., 37, No. 1, 27 (1970).
14. A. I. Voloshchuk et al., At. Energ., 29, No. 6, 416 (1970).
15. W. McDonell, "Model of cavitational swelling of uranium," International Conference on
Physical Metallurgy of Reactor Fuel Elements, Berkeley, England, Sept. 2-7, 1973, p.266.
16. I. Novak, F. Slancar, and K. Splichal, Jad. Energ., 21, No. 5, 172 (1975).
*References 6 and 6 are identical as in the Russian original ? Publisher.
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CRITICAL ENERGY IN HEAT-TRANSFER CHANNELS OF COMPLEX SHAPE
L. N. Polyanin UDC 621.039.524.44:621.039.51
The theoretical analysis of the sharp deterioration of heat transfer as a local phenome-
non depends upon estimates of the detailed velocity and temperature fields for the heat
carrier over the total channel volume. The critical situation in the establishment of steady
flow should then arise out of a phenomenological consideration of the general combined problem
of the hydrodynamic and thermal stabilization of flow in specific flow conditions (channel
shape, characteristics of the heated surface, initial flow conditions). In practice, however,
empirical correlations based on calculations of the mean or local parameters are used for
the estimates of the critical heat fluxes in heated channels.
Because there are no sufficiently rigorous equations of heterogeneous two-phase flow, the
local heat-carrier parameters in complex channels and, in particular, in bundles of rods [1,
2] cannot be determined with sufficient accuracy, which in a number of cases leads to pro-
nounced lack of agreement between the experimental and calculated values of the critical
energy [2, 3]. However, this lack of agreement is not simply the consequence of the in-
adequate rigor of the expressions of the determination of the local flow parameters. Other
factors ? the effect of low-frequency pulsations on the thermocouple readings, systematic
and random errors of the measuring instruments, and deviations of the geometric characteris-
tics of the channel from the nominal values ? mean that the maximum possible accuracy of the
empirical correlation should be assumed to correspond to a mean square error of q6-6% [4].
As a result of the indeterminacy that exists at present in calculations of the distribu-
tion of the heat-carrier parameters over the channel volume, the local approach to the cal-
culation of the critical heat fluxes has no advantage, in terms of accuracy, over the so-
called global approach. Moreover, global estimates of critical channel energy in terms of
the inlet parameters of the heat carrier, taking into account the real energy distribution
and the channel geometry, are more convenient for practical engineering calculations.
The present article outlines a method of calculating the critical channel energy; the
method rests on the generalization of experimental material in a sufficiently wide range
of flow and geometry parameters, so that it can be used to estimate the critical power in
various conditions of forced and natural circulation of heat carrier.
As an example, consider an assembly of fuel-element rods (Fig. 1). The channel is
divided into arbitrary hydraulic cells and the thermohydraulic diameter is introduced
dth.f (1)
wheref ? is the through cross section of the j-th cell; also
3
1 ?t
1t01,maX E nrplop
rj
(2)
nii is the part of the perimeter of the l-th heat-transfer surface belonging to the j-th
cell; K1. is the heat liberated in the l-th rod; Kmax = max KL .
rj rj 7. rj
The cell which corresponds to the minimum thermohydraulic diameter is called the maxi-
mally heat-stressed cell.
The equation for the critical heat flux can be written in the form
=a?bx,
(3)
Translated from Atomnaya Energiya, Vol. 42, No. 6, pp. 457-460, June 1977. Original
article submitted August 30, 1976.
This material is protected by copyright registered in the name of Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y.
10011. No part of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic,
mechanical, photocopying, microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is
available from the publisher for $7.50.
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Fig. 1. Cross section of channel arbi-
trarily divided into hydraulic cells.
3
Fig. 2. Cases in which Eq. (11) is and
is not satisfied for various energy dis-
tributions.
where-the parameters a andb are functions of thecoordinatesof the point of heat-transfer crisis
Eo = Z0/14 and, in particular, the shape of the energy distribution Kz(E) = q5()/Fi5 (Cis is
the mean over the channel height of the specific heat flux).
The mass vapor content x for the maximally heat-stressed cell is written in the form
go
X =(flcr/Q) K (t) 4 ?-.(Aiin/r), (4)
where kr is the critical energy of a Channel with a nonuniform energy distribution over the
-height; also
Q==rMV445;
(5., min if dth
Here Wo is the mean mass velocity of heat carrier in the channel; Ho is the total heated
perimeter; r is the latent heat-of vaporization.
At the critical-point
520
Since
(5)
(6)
ger (o) = qs(o). (7)
qa (to) moil) eaxkz
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(8)
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it follows from Eqs. (3), (4) and (8) that
kr
where
II0H [a +b (Min /r)]
max to
K2 ()d
x =b11011.1Q
Also, when the curve of K(F) has a continuous derivative for 0.5..a.51:
dqcAo) dq. (o)
--
Then Eqs. (3), (4), and (11) give
dK
K:Inax Ar-z (tO == 0.
In Fig. 2, curve 1 corresponds to a case when Eq. (11) is satisfied and curves 2 and 3
to cases where it is not.
(9)
(10)
(12)
Consider the example of the linear field
and assume that Eq. (11) is satisfied at the channel outlet (Eo = 1); neglecting
dependence of the parameters a and b on the shape of the energy distribution, the result
obtained is
(13)
the
0-2x/(2el3x-I-x), (14)
i.e., the gradient of the vertical field may be fairly significant (tct-,' 1) , and the crisis
will nevertheless occur at the channel outlet.
Experiment confirms this result and indicates that when the vertical energy distribution
is only slightly nonuniform, the critical energy is largely independent of its shape. The
critical energy of a channel with a nonuniform energy distribution over the height is deter-
mined in terms of the corresponding value of the critical energy for a channel with a con-
stant energy distribution as follows
the inlet enthalpy
the excess heating
cell when the real
parameter Kz(E0).
1VCI
kr 01'11)4/4
to
(Ko/Ko) 3 K. (E)
0
(15)
iin in the calculation of kr being reduced by an amount corresponding to
of the heat carrier over the length Eo for the maximally heat-stressed
energy distribution is replaced by a constant distribution with relative
To compensate the excess energy leading to an overestimate of the heating
Kzo /Kzo is included in the denominator of Eq. (15); IC" =-- Kz (to);
ence of the
assembly
K." = (1/E) 1111rz(t)4. The pres-
o
integral up .to the limit in the denominator means that the upper part of the
? Eo)hasno effect on the critical energy of the lower part.
The expression for the critical energy of the channel with constant energy distribution
over the height is written in a form corresponding to Eq. (3)
where 0 = a/b; x is the outlet equilibrium vapor content; and
(17)
As a result of this analysis and treatment of a large quantity of experimental data
(in particular, [5-9]), the parameters $ and c may be written as follows
(18)
(19)
(16)
= (op 1K.Tax)
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Here EH = H/H*, and hence H* = 5 in.
The expressions for A(p), B(X.), and C(X), where
Pcr
are
(20)
--=-Wp6/(Wp6)*; (21)
= 225.6 kgf/cm2; (Wpd)* = 0.707 kgf/m?sec, are independent of the type of channel and
universal functions of the heat-carrier flow parameters, as follows
A=I
(22)
I. 2--ft, p, 0 of the
absorbed dose rate Pc, and of their ratio as
a function of separation from the edge of the
core of a VVR-Ts reactor.
Using the VVR-Ts reactor as an example, Fig. 2 shows the variation of the neutron flux
density TE > 0.1 mei and of the absorbed dose rate in graphite Pc as a function of the separa-
tion from the center or edge of the core. We assume that irradiation of some object, the
sensitivity of which is roughly the same for neutrons and y rays, is performed in the core
at 63 cm from the edge of the core until identical fluences are reached, i.e., the identical
radiation effect. Figure 2 indicates that the absorbed dose will be 100 times less in the
first case than in the second. Therefore extrapolation of the results obtained in the first
case to the irradiation conditions in an actual situation corresponding to the second case
will lead to an overestimate of the radiation resistance of the object.
Thermal Neutron Contribution to Radiation Effect. Formal addition to dosimetric informa-
tion of the values of thermal-neutron influence does not always clear up the situation in
radiation tests. It is well known that the addition of boron reduces the radiation resistance
of materials during irradiation in thermal reactors [8]. For example, the additionof 5 wt.%
of boron to natural rubber decreased its radiation resistance by a factor of 25. Without
additional tests under other irradiation conditions, it remains unclear whether the indicated
addition of boron sharply increased the absorbed energy or whether the specific effect of a
particles played a decisive part.
Review of Existing Proposals
Because of the complexity of the situation during intrareactor irradiation, it is hardly
possible to find a single universal criterion connecting radiation effect and irradiation
conditions for any material. We consider the following proposals.
1. Optimal Value of Neutron Threshold Energy. The minimum effect of a neutron spectrum
on irradiation results is achieved by using a fluence of neutrons with a threshold of 300-
400 keV [7]. This conclusion is confirmed by more recent calculations which include the
effect of the disorder regions (clusters) produced by the neutrons and by direct experiments
on irradiation of alloy transistors [9]. For all groups of transistors, the use of 237Np as
a threshold detector is an acceptable compromise. In that case, the effect of neutron spec-
trum on the change in gain is no more than 20%.
2. For materials where atomic displacement is the controlling mechanism, the "number
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of atomic displacements per atom of stopping medium" is used as the MRE [10]. In principle,
such an MRE is universal for neutron, electron, and y irradiation but is limited to certain
materials in which the change in properties is produced by a breakdown in the regularity of
the relative locations of atoms.
3. The various irradiated objects are subdivided into classes, e.g., metals, semicon-
ductors, organic materials, etc., and each class has its own MRE. Such an approach is widely
accepted but it does not eliminate uncertainties in the interpretation of irradiation results.
4. It was proposed (11, 12] and analytically demonstrated [13] that one can use the
neutron component of the absorbed dose as the MRE for irradiation of metals and semiconductor
materials with the dependence of the number of defects on the neutron spectrum being sharply
reduced in this case. An experimental confirmation of this proposal would make it possible
to unify the analysis of test results for practically all classes of materials from the
standpoint of absorbed dose.
There are a number of other very interesting proposals with reference to methods for
describing irradiation conditions but they all reduce to one of those enumerated above.
Design of Radiation Tests for Materials
To discover the cause for the production of radiation breakdowns and to ensure the com-
parability of test results from different reactors, it is necessary to perform irradiation
with several markedly different neutron spectra with and without thermal-neutron filtration
and also with different ratios of neutron and y-ray fluxes.
Thus, considering the present level of development of dosimetric methods and the degree
to which reactors are supplied with the appropriate detectors and instruments, one can re-
commend as a minimum program the following procedures for the irradiation of materials with
provision of dosimetric information.
For Metals and Alloys: reactor irradiation in three different neutron spectra with and
without filtration of thermal neutrons.
For Semiconductors and Semiconductor Instruments: irradiation with isotopic y irradiators
to a dose of several tens of megarad and reactor irradiation in two different neutron spectra
with and without filtration of thermal neutrons.
For Organic Materials: irradiation with isotopic y irradiators and reactor irradiation
in channels with a significant fast-neutron contribution to the absorbed dose (not less than
40-50%) with and without filtration of thermal neutrons.
Dosimetric Information: for irradiation of y irradiators ? absorbed dose in water or
exposure dose; for reactor irradiation ? integral fast-neutron spectrum measured with thres-
hold detectors [14], thermal-neutron fluence, absorbed dose and its y and neutron components
in a hydrogenous material (in polyethylene, for example). Additional information is the y-
ray spectrum or the y-ray absorbed dose in the material. Indirect information about the '-
ray spectrum can be obtained by means of the spectral parameter [15]. It is recommended that
the absorbed dose rate for intrareactor conditions be measured by means of a calorimeter with
polyethylene and graphite absorbers which provides an opportunity not only to determine the
dose components in hydrogen and carbon, but also to calculate their contribution to the dose
in any material [16], the relative number of displaced atoms [13], and the fluence of neutrons
with a threshold of 0.3 MeV.
It is to be hoped that the accumulation of such information and the comparison of the
results of radiation testing under various irradiation conditions will lead, in the final
analysis, to the use of a more universal criterion for any object ? the effective value of
the absorbed dose
NFU = aDen bDth
where the coefficients a and b take into account the relative effectiveness (in comparison
with y rays) of neutrons for objects of a given class.
LITERATURE CITED
1. M. W. Thompson, Defects and Radiation Damage in Metals, Cambridge Univ. Press (1969).
2. R. F. Konopleva, V. L. Litvinov, and N. A. Ukhin, Features of Radiation Damage in Semi-
conductors by High-Energy Particles [in Russian], Atomizdat, Moscow (1971).
527
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3. V. A. Girii et al., At. Energ., 35, No. 1, 48 (1973).
4. D. Hammen et al., in: Effects. of Radiation on Materials and Components, Reinhold Publ.
Corp. New York (1964).
5. V. D. Popov et al., in: Metrology of Neutron Radiation in Reactors and Accelerators
[in Russian]; Vol. 2, Standartov, -Moscow (1972), p. 163.
6. J. Mottef, in: Proceedings of the IAEA Symposium "Neutron Fluence Measurements," Vienna
(1970), Rep. N 107.
7. N. A. Ukhin and A. V. Khrustalev, Preprint IAE 18-80, Moscow (1969).
8. R. W. King et al., in: Effects of Radiation on Materials and Components, Reinhold Publ.
Corp., New York (1964).
9. V. A. Knyazev et al., in: Proceedings of the IAEA Symposium "Neutron Monitoring for
Radiation Protection Purposes," Vol. 2, Vienna (1973), p. 321.
10. J. Buswell, in: Proceedings of the International Conference "Phys.Met.React.Fuel Elem.,"
London (1975), p. 170.
11. B. A. Briskman and V. P. Savina, in: Metrology of Neutron Radiation in Reactors and
Accelerators [in Russian], Vol. 2, Standartov, Moscow (1972), P. 168.
12. E. A. Kramer-Ageev et al., in: All-Union Symposium on Radiation Defects in Semiconductors
[in Russian], BGU, Minsk (1972).
13. E. A. Kramer-Ageev et al., At. Energ., 34, No. 4, 255 (1973).
14. E. A. Kramer-Ageev et al., Izmer. Tekh., No. 1, 61 (1973).
15. Yu. L. Poglin and S. S. Ogorodnik, At. Energ., 38, No. 2, 96 (1975).
16. B. A. Briskman, Components of Absorbed Dose of Reactor Radiation [in Russian], Atomizdat,
Moscow (1976).
17. L. Miric and Z. Ubovic, in: Proceedings of the Third IAEA Nucl. Accident Intercomparison
Experiment, Beograd, May 14-25, 1973 p. 6.
MODEL STUDIES OF THE STRESSED STATE IN PRESSURE VESSELS
N. N. Zorev, Yu. S. Safarov,
V. K. Tutynin, V. N. Sakhelashvili,
and N. L. Narskaya
UDC 620.171.5
The requirements for quality and reliability of reactor vessels and other equipment at
nuclear power stations can be met by using modern methods for evaluation of load capacity
and strength. One of the steps is the determination of the stress-strain state of vessels
under the action of operating toads. This paper describes the results of a study of the
stress state of a reactor vessel by the polarized-optics method of "freezing" strains using
models made of optically sensitive materials. The model was made of hardened epoxy resin
on a 1:20 scale with the preservation of complete geometric similarity to a prototype (Fig.
1) and consisted of a vessel and a dome joined by 60 studs. The vessel wag built up from
three regions: a piping region, a support shell, and the bottom. In the simulation of the
operation of the flange connections (sealing points), a procedure was followed which essential-
ly, involved the selection of a preliminary "tightening" of the studs at room temperature in
order to obtain the required tightening at the "freezing" temperature.
. We consider the practical application of the procedure. According to similarity theory,
the force similarity of model and prototype (subscripts "m" and "p") in the case of uniaxial
stress states is determined by the relations
or crin/Em.--- CY p/Ep , (1)
where a = Zp/Zm is the coefficient of geometric similarity; 0 = pp/pm is the coefficient of
force similarity; X = Ep/Em is the coefficient of material similarity; cm and op are stresses;
Translated from Atomnaya Energiya, Vol. 42, No. 6, pp. 465-472, June, 1977. Original
article submitted April 26, 1976; revision submitted February 7, 1977.
528
This material is protected by copyright registered in the name of Plenum Publishing Corporation, 227 West 17th Street, New York, N. Y.
10011. No part of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic,
mechanical, photocopying, microfilming, recording or otherwise, without written permission of the publisher. A copy of this-article is
available from the publisher for $7.50.
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Fig. 1. External view (a) and diagram (b) of a model of a water-
cooled?water-moderated reactor: 1) bottom; 2) support shell; 3)
support ring; 4) piping region; 5) dome; 6) dome piping; 7) studs;
8) sealing ring; 9) pipe fitting.
Zm and Zp are geometric dimensions; pm and pp are loads; E is the modulus of elasticity. Then
the force stretching a stud during tightening is
Prri==pp(1A04 (2)
An experimental check of the correspondence of the stresses in a stud from tightening
at room temperature and after "freezing" was made on models which simulated the region of
the flange joint of the reactor. Tensile forces of 0.1, 5, 10, and 15 kg were established by
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preliminary tightening. Uniform tightening was accomplished by means of a torque wrench. The
wrench was calibrated in a special device. The gauges on the wrench recorded the forces
produced in the studs during loading. Simultaneous recording of the readings of the gauges
on the wrench and studs by measurement of static strains provided an opportunity to establish
the dependence of the strain in the elastic element of the wrench on the stresses in the studs
during tightening. The deviation from the average stress was ?5%.
After assembly and gluing, the model was loaded by an internal pressure in accordance
with a standard temperature regime which ensured "freezing" of strains in the model. An
analysis of present practice shows that the pressure is usually set arbitrarily based on the
possibility of obtaining a larger number of bands in cut sections of the model in order to
increase the accuracy of the measurements. Strains produced in the model are more than an
order of magnitude greater than the strains in the prototype and thereby geometric similarity
breaks down; this can lead in a number of cases to larger errors than those resulting from
an inequality in Poisson coefficients. At the same time, an attempt to obtain a unique pres-
sure in the model based on equality of the components of the strain tensor is unsuccessful
because of the inequality of the Poisson coefficients of the model and prototype. In loading
a cylinder by an internal pressure, the following relations are valid for the components
e1,2,3 of the principal strains:
(pRIEt)(1? p/2);
e2=(pRIEt)(1/2??);
63 = (?pRIEt) 3/2
(3)
where p is the internal pressure; R is the mean radius of the cylinder; t is the thickness
of the cylinder wall; p is the Poisson coefficient.
If it is assumed that
(i)m= (83.)! (52)m= (82)p: (83)m--= (83)p,
when pm = 0.5 and Pp = 0.3 we obtain, respectively,
p;rf-= 1.133pp (Em/E'p);
0.2
pm= pp (/Epl= oo ;
pm? 0,6pp (Em/Ei3),
i.e., the pressure in the model is indeterminate. The following method is proposed for
determination of stresses in a reactor vessel from measurements in a model in order to elimi-
nate this indeterminacy and to "neutralize" the effect of the Poisson coefficient.
The strain state at a given point is characterized by the strain intensity
8i =
2(1) 17-(81? 82)2 + (82 ? e3)2 (83-802, (4)
+
which does not depend on the Poisson coefficient.* In a uniaxial stress state, the relations
between strains will be 82 = ?1181; 83 = ?Rei; st-82,-_-6,(1+p,) and the strain intensity c=c1. In
the case of a thin-walled cylinder under an internal pressure producing strains whose values
are given by Eqs. (3), the intensity is
8i= (0-12)(pRIEt).
Thus, in order to obtain an identical strain state in model and prototype (eliminating
the effect of the difference in Poisson coefficients in this way), it is necessary that the
*When p = 0.5, Eq. (4) reduces to the more customary expression-in the mechanics of deform-
able media
ei = l/(81 ? c2)2+ (82? 63)2 + (63 61)2.
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strain intensities be equal, i.e.,
Since in the elastic load region
where the stress intensity is
ei= a LIE,
CY = 1/2/2 (a ? a2)2 + (a2? (J 3)2 (cr3 ai)2,
we then have from Eqs. (5) and (6)
(ailE)m= (ai/E)p.
At the same time, the stress intensity at each point of model and prototype (since the
elastic problem is being considered) is directly proportional to the load (pressure), i.e.,
=kp. (9)
Equations (8) and (9) show that the pressure determined by the equality of strain intensities
(5) is
(EE) pp.
(10)
The condition (10) guarantees equal strain intensities and identical order of magnitude for
the strain components in the model and reactor vessel although the disparity of the components
of the strain tensor remains.
With a pressure in the model realized by means of Eq. (10), the components of the stress
tensor, or their differences, can be determined; from them, using Eqs. (7) and (8), the
stress intensity is found to be
{112/2 ? a2)2+ (a? a3)2 + (a3? ai)lp = (E/E,) (ai)m . (11)
Direct transition from the stress tensor in the model to its components in the reactor
vessel, as is done at the present time in similar studies using the relation
(a,)?PpiPm(ai)m (12)
is not rigorous and introduces an indeterminate error in the calculation. In fact, from the
equations of the theory of elasticity
al; =Mijekk-F2Geu, (13)
where
=-- REA! ti,) (I ? 2R); G = E 12 (1 ? ?); (14)
1, (15)
ei-i-e2-1- 83; 6 =
which are brought to the form
ai/E = 8,1(1 + it) ? P? (81 + e2 + e3)
(1+)(t-2)
it follows that
(16)
(aii1E)mr--(crulE)p or (giiip)m= (ciii/p)p
which are sufficient when Eq. (10) is taken into consideration, can be satisfied only when
the right-hand sides of Eq. (16) are equal for model and prototype. However, equality is
not satisfied since
in the model and the right-hand side takes on an indeterminate value, while in the prototype
8i+ 82 ? 83 0 andR=0.3.
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Consequently, we have for the components of the stress tensor the inequalities
(aii/E)m0 (aii/E)p or (cru/p)mo (crii/p)p. (17)
As follows from Eq. (4), the strain intensity at corresponding points of model and re-
actor vessel will be equal in the case where we have the relations
Ret e041 HI-1-06== Rsi --80A4 VAp.
Returning to Eq. (16), we obtain
(18)
(19)
Thus, there- follow from Eqs. (18) and (19) relations-for the differences of the com-
ponents of the principal stresses in the model and reactor vessel,
{(cri ?02)/E]m= [(at ? cr2)/Elp.
(20)
To determine contour stresses, it is sufficient to use the valueof the third principal
stress in the reactor vessel, which is equal to the internal pressure, i.e.,
(21)
on the internal surface and zero on the external surface. The other two stresses are deter-
mined from Eqs. (20) with Eq. (21) taken into account:
(a0p.=(a) p A- (a,--oOngi) tE0 ; (22)
(472)p= (0.3)p? (a2?cr3)m(EPIEm).
In the general case of a stress-strain state based on calculated values for the principal
stress differences in accordance with Eq. (20), it is necessary to consider the equilibrium
equations in addition when making separate determinations of al, a2, and as. Having found
the stress components, one can determine the components of the strain tensor from Eq. (19);
they will be different for model and reactor vessel. Thus, the actual stress-strain state of
the vessel is determined by the approach given.
When the pressure in the model is set higher than follows from conditions (10), this
means that
(em= (ed-p11. (23)
where n is. the number by which the pressure in the model calculated from Eq. (10) is multi-
plied. Such an arbitrary increase in pressure is permissible where the strain has no effect
on the distribution of the effective external forces in the body. If in each specific case
the increase in pressure above the calculated value is sufficiently justified, stress cal-
culations for the reactor vessel are performed by determination of the differences in the
principal stresses from the condition
Rei---02)/Elm=--[(cri-02)/Elpn (24)
-using the equilibrium equations in the general case. The coefficient n for -a given pressure
Pm in the model is given by [see Eqs. (8) and (10)]
n=pmEpippErn. (25)
. After loading and "freezing" of the strains, the model is cut up into thin sections
2-3 mm thick and the optical path difference determined at required points by the method of
band compensation. The measurement error was 0.02 bands.
One of the factors which is extremely difficult to take into consideration when using
the "freezing" method is the fringe effect which is produced by the instability of the proper-
-ties of optically sensitive materials. The creation of a pressure in the model in accordance
with Eq, (10) for simulation using Eq. (5) leads to a situation where the fringe effect be-
comes commensurable with the optical path difference in the "frozen" sections and introduces
a significant error in the measurements.
Studies showed that repeated annealing of the sections has no influence on the fringe
effect (Fig. 2). Thus, in order to determine the actual stresses in three-dimensional models,
it is necessary to measure them in two stages. First the stresses are determined in flat
"frozen" sections. Then the sections are annealed in a constant-temperature bath. Annealing
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Fig. 2. Measurements of stresses
produced by fringe effect in a flat
sample after the first (0), second
(m), and third (L) annealing.
P Il 0,155
18 15
a
180
A
Fig. 3. Experimental verification of
the technique for consideration of
fringe effect in flat samples: A) load-
ing scheme; B) stress curves over height
of sample: 1) total stresses from effec-
tive load and fringe effect; 2) stress
distribution in sample after annealing
(stress curve created by fringe effect);
3) stress distribution over sample height
produced by load.
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1780
3120
Fig. 4. Cross sections of the piping region
in "frozen" sections of the model in the
transverse (a) and longitudinal (b) direc-
tions.
2930
1730
2330
3060
2460
1460
3010
am
M40
?
28W
2930?
1830
1310
tI 2810
2730
1272620
0t
2560
2080
2810
2840
1890
1680
2630
2710
1540
R280
Fig. 5. Curves of contour stresses in a plane normal to a piping-region section.
removes the stress produced by a load, but the "stress" from the fringe effect remains. The
difference of the two measurements (with the sign of the fringe effect taken into account)
is the true stress at a given point of the model. A check of this technique was performed
with flat samples loaded in accordance with a pure bending scheme over the rectangular sec-
tion I, II, III (Fig. 3).
The maximum stress in the central section including the intrinsic mass of the model is
found from the expression
omax = IW =61bH2 [q12 (0.5 ?a/I) + pc],
where q is the distributed load from the intrinsic mass of a beam (the other notation is
shown in Fig. 3). On the other hand, amax is associated with the optical constant of the
material by the relation
(26)
amax.= GO.? (mlb),
(27)
where al'? is the optical constant of the material in the model; m is the order of the band
on the sample contour; b is the thickness of the model.
From these expressions one can obtain the order of the band on the contour:
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1190 1810
1820
Fig. 6. Curves for contour stresses in the plane of a section through
the upper set of piping.
3280
2860
1620 NO
II is
Rimini' ;le*
Ayip
a
Fig. 7. Distribution of Circumferential (a)
and meridional (b) stress in reactor drime
(P = 180 kgf/cm2).
=
6
Igl2 (0.5? all) + pc].
G1,0112
0
(28)
Figure 3 also shows the results of stress measurements in the central section of the
sample. The deviation between calculated and experimental data is 1.2%. These studies con-
firm the correctness of the method developed for including the influence of the fringe effect
in the determination of stresses on the contour of three-dimensional "frozen" models.
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Fig. 8. Distribution of circumferential and
meridional stresses in reactor bottom along in-
ternal (a) and external (b) surfaces.
Fig. 9. Computational scheme for bottom of
reactor vessel: A) bottom; 13) cylindrical
shell.
Using the method described, the circumferential and meridional contour Stresses were
determined. We denote the circumferential, and radial stresses by at, a2, and
a3.: By transillumination of longitudinal and transverse sections of the model, the data
required for stress calculations were obtained in the form
(29)
473-473 = I
where n is the order of an isochromatic band (see the sections in Fig. 4); a, kgf/cm2-band,
is the _optical constant of the material in the section.
The results of stress determinatiOns (in kgf/cm2) for the prototype are shown in Figs.
5-8. The culves-(solid and dashed) correspond to calculated data and the points to experi-
mental data. The stresses for the bottom portion of the reactor, which can be considered
-axisymmetric in view of the separation from piping, were also calculated. from the moment
theory of shells.
The complete system of differential equations describing the stress-strain state of a
shell of rotation with an arbitrary meridional form is written in the matrix form ,
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15
14
12
10
8
6
2
0
,ex
ctin
ex
am
2
4
6
1
1
A
A
Figure 10. Calculated stress curves: a) membrane and bending stresses in el-
liptical bottom A; b, c) total stresses in ellipical bottom A and torospherical
bottom C; B is the cylindrical shell.
where the state vector
(PO
rH )'
r1117?
p is the radial displacement; 15t is the angle of rotation
of the normal to the meridian; H is the thrust; Mm is the meridional bending moment (Fig. 9).
The numerical matrix F =
?
cos 8 (1?R2) cos2 0
0
Ehr
cos 8
0
Eh3r
Eh
0 ?
0
Eh3 cos2 0
12
cos 0
11' r
sin 6
0
cos0
r
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The load vector g--=
1-112 sin 8 cos 9 (s)
Eh
0
sin 0
\ ?cos Of (s),
where f (s) = (q? cos ? q1 sin 0) rds;
,
0
D = qi coS ()
+n sine; qi,q? are intensities of the distributed load tangential and normal to the meridian;
r, h, and 0 are geometrical parameters (see Fig. 9).
Solution of the problem was carried out numerically by orthogonal stepping* for the
boundary conditions Mm .= Mt and Tm = Tt in.a band and damping of the moment solution in a
long cylinder. Force factors were determined from the resultant vector which described the
stress state of the shell:
7',?=H cos0 +--sin 0;
Ti= pl,m+ Eh p;
Eh3 cos 0
= I1Mm + 12 r
Then the membrane stresses
?
m(t)
and the bending stresses
Tm(0
(31)
uB ? BA/mu.
m(t) h2
were found. Here, Tm(t) is the meridional (circumferential) membrane force; Mm(t) is the
meridional (circumferential) bending moment.
To evaluate the effect of bottom shape on its stress state, calculations were made for
two types of model (Fig. 10): an elliptical bottom (a/b = 2) and a torospherical bottom
(R1/a =1.9, R2/a!= 0.4, where R1 and R2 are the radii of the circles forming the meridian
of the torospherical bottom). The bending stresses are of the same order of magnitude as
the membrane stresses (see Fig. 10a) and this points to the necessity of including them, i.e.,
to the correctness of the choice of moment theory. The curves in Figs. 10b and 10c illustrate
the total effect of bending and of membrane tension where the stresses at points of the ex-
ternal surfaces were calculated from the expressions
ex _ m B
m(t)--amMl-am(t);
oin =0M (32)
m(t) n1(0 m(:)?
From a comparison of the curves for the two types of bottom, it is clear that they are
almost indistinguishable with respect to stress state and are equivalent from this standpoint.
The good agreement between calculation and experiment should be noted..
These studies show that for an internal pressure set at 180 kgf/cm2 the stresses in
elements of the reactor vessel do not exceed the yield point, 'which is 55 kgf/cm2. ,The piping
region of the reactor vessel shows the greatest stress. Furthermore, the maximum stress is
concentrated in the transition region between pipe fitting and reactor vessel. The insignif-
icant effect of transition radius on the stress level in the transition region was shown for
. four transition radii.
Stress distributions in plane sections passing through the upper and lower sets of pipe
fitting were practically identical and this is evidence of the significant effect of the
support ring located between the upper and lower sets of fittings.
*S. K. Godunov, Usp. Mat. Nauk, 16, No. 3(99), 171 (1961).
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The agreement of experimental and calculated results indicates the appropriateness of
the use of the "freezing" method for studies of objects of complex geometry since in such a
case the use of computational methods is associated with tremendous difficulty and complexity.
SOME PROBLEMS ON THE MECHANISM OF BREAKUP AND MASS TRANSFER IN
PULSE EXTRACTION COLUMNS
E. I. Akharov and S. M. Karpacheva UDC 66.063.3.023.3
The droplet size de in column extractors enables one to determine the principal techno-
logical indices of the equipment: the specific throughput (W vo '1, do), efficiency (HTU
??2/dD), and rate of phase separation in settling tanks (vi do, he).
Breakup of a droplet into two equal parts requires the energy
Eb==Cnad;. (1)
This energy can be supplied by rotating disks, agitators (columns with a mechanical energy
supply), and reciprocating motion of plates (vibration columns) or of a continuous phase
(pulse columns).
Breakup of the dispersed phase in extraction columns can occur in the center of the
flow of the continuous phase as a result of turbulent pulsations in the flow rate of the
continuous reagent (the change in velocity at opposite ends of the droplet is AU), induced
by reciprocating motion of the continuous phase, by change in the dynamic pressure at the
holes, channels in the packing, the column walls, and the packing surface (rings, plates)
when the flow rate changes and when the droplet strikes the packing surface.
To elucidate the breakup mechanism of the dispersed phase in pulse columns, we carried
out still and motion picture photography of various sections of a model column (cross section
176 x 176 mm; Ho = 1.2 m), using the most common types of packing (Table 1).
We employed a system of 20% tributyl phosphate (TBP) in kerosene (fraction boiling at
170-240?C) with 0.1-0.2 M HNO3 in water (a = 12 dyn/cm, pc =1 g/cm3, Ap = 0.21 g/cm9, Pd =
1.65 cP).
We found that in pulse columns with sieve plates and KRIMZ packing breakup of the dis-
persed phase occurs in the center of the flow of the continuous phase at the entrance and
exit of the holes. When the KRIMZ packing is used, the droplet size becomes essentially
constant at the point along the column at which the dispersed reagent has passed three plates.
The droplet size is determined by the type of packing, the pulse intensity, and the
properties of the liquid system (Fig. 1). Variation of I = fa will provide a specified do
in almost any system [1, 2]. Since theory [3] gives the energy required for breakup of a
droplet as
ED-=- CpcAU2d/s),
(2)
then when the change in velocity AU is due, respectively, to inertial or viscous forces [3,
4]
AU2_--C11 82/3d2g3;
AU2?C2epedp?.
*Conventional symbols are listed at the end of the paper.
fin all cases is the constriction coefficient of the flow [5].
(3)
(4)
Translated from Atomnaya fnergiya, Vol. 42, No. 6, pp. 473-477, June 1977. Original
article submitted May 19, 1976.
This material is protected by copyright registered in the name of Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y.
10011. No part of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic,
mechanical, photocopying, microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is
available from the publisher for $7.50.
539
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dr), mm
50
100 150
IeP cm/min
ZOO
Fig. 1. Dependence of droplet dia-
meter on the pulse intensity; 1)
Raschig rings (10 x 10 x 0.4 and
25 x25 x 0.8 mm) (data from [7]);
2) KRIMZ packing, pulse frequency
100 (0); 176 (A); 310 (v), 360 min-icD.
TABLE 1. Properties of the Stainless Steel Packings
Type of packing
Geometry of the packing
Method of filling
Raschig rings
Sieve plates
KR1MZ?
10x10x0,4mm; ep=0,82
25x25x0,8mm;ep=0,866
do= 3,2mm, F=20-25%
Holes 10x20 mm
cc=10-45?; F=20-43%
Continuous, sectional with gaps
0,05-0,2 mm
h.=50 mm
P
hp= 50 ? 200 mm
*Inventor's Certificate No. 175489; Byull. Izobret., No. 20, 18 (1965); KRIMZ
comes from the initials of the inventors' family names.
These expressions for the breakup of the dispersed phase in pulse columns with sieve
plates and kRIMZ give
3D= eonst (ove)0.68-0,4;
const(olt/pN?33.
For sieve-plate columns c is [5]
or
or
(5)
(6)
e= :t2 (1? F) (fa)312C:*F2hp (7)
c-i?D:= C3 (0./pe)0.6 [F0.8. ,1):)
n 4 1(1? F)? '4i (fa)-112, (8)
C3W67?'6 EF2/(i_F)10.4 (hp/4)0,4;
ci(00.3311.33/p0c.66) x EF0.664.33/(f F)0.331 (fa)-1.0,
dpido---c4weT?.33ReT??33[F2/(1?F)1??33(hp/d0)?33.
*Co = 0.6 is the constriction coefficient of the flow [5].
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(10)
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HTU0
HTUI
1- f a
Fig. 2. Effect of pulse intensity
(I = fa) on the operating efficiency
of the pulse column in various operat-
ing regimes of the equipment: 1, 2,
3) at 12 > 12 > 12.
TABLE 2. Correlation Equations for (ID (experiment)
Property of the system
' Property of the packings
Experimental relation
a=5,8-50; pc=0,998-1,084;
pd=0,67-0,81; itc= 1-2,9;
pd =0,4-4,2
a=12, pc = 1,02; pd=0,796;
pc=1,02; [Ad=1,65
o=1,08-37; pc=0,99-1,0;
pd=0,8-0,958;
Methyl isobutyl ketone?water
a=6-24; pc =1-1,25;
pd=0,6-1,0; pc=1-1,2;
pd=1,2-6
Packed column
Rasching ring (ceramic)
7,76x 25 mm
cp=0,54-0,77
Rasching ring (stainless
steel)
10 xluX0,5 mm
25x25x0,8 mm
Sieve-plate column
d0=5-10 mai
h 0 , 25 .mm
r=0,14-0,61
D-0 81 (4/033)-1,2 [11]
? p
Dpc(f a)/ile= 2,21-10-24 X
x [aperiphipo,528[Apgcvcoo, [Nip 00,17 [8]
2D= C (a/po)0,5(fa)1,4 [9]
D= ?
0 439d [0F0-,5/(5,1fa-FW,)2podo]o,s
Column with KRIMZ packing
hp= 50-250 mm
a=10-450; holes
10 x20-70 x 140 mm
[ 10 ]
0,135F (0/pc)0,6 (fa)-1,0;
F 4a2 sin a (2+ cos a) Nhc1/3-012) [12]
In pulse columns with Raschig rings, the breakup of the dispersed phase depends on the
pulse intensity and the state of the packing surface ? its wettability by the dispersed
phase, which is characterized by the contact angle 6.
For a packing that is not wetted by the dispersed phase at low intensities (100-200 mm/
min), breakup of the dispersed phase occurs when the droplet strikes the surface of the ring.
At high intensities breakup takes place close to the surface of the ring. In this case
theory [6, 7] has
c/D? (av/pcvD?'5,
where vt is the rate of turbulent pulsations.
Assuming vt r.r. fa, we get
or
-d-D= c, co-400.511g., (fa)-i'43
?dp/dp=C5Wei??5Rel-??5.
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(12)
(13)
(14)
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TABLE 3. Effect of Pulse Intensity of
the Parameters of a Pulse Column with
Distributive Packing
Index
1, cm/sec
o 0,33 10,661 1,0 1,73 1 2,0 12,33
Experiment
-dD
0,46
0,32
0,25
0,145
0,10
0,08
ap/bD
0,8
1,3
1,1
1,0'
1,0
1,0
De
1,02*
1,02*1,0
1,0
1,5
2,0
Dei
65*
65
65
60
24
'12
c.%
4,6
5,5
9,0
40,0
23
30
HMI.
92
82
66
48
45
21
Calculation
60
100
180
452
1390
1800
TW
10,5
40,6
11
11;2
13,0
14,3
AlVd*
218
181
111
100
43
33
FUTUrad
12
16,4
22
24,8
..24,1
17,8
HTUm
80
65,6
44
23,2
20,9
3.2
Ke
-1,0
1,22
1,81
3,43
3,8
25,0
Km
1
1,66
3,00
7,53
23,1
30
Kin
1,0
0,7350,6
0,465T
0,165T
0,83
0,06
1,0
2,5
- 6
24
22,5
2400
9,9
21
19,3
3,5
22,9
. 40
0,57
*Assumed for I = 0.66 cm/sec.
tThe substantial change in Km on going
from I = 1 cm/sec to I = 1.73 cm/sec is
plainly due to the error in measuring
HTUi in this pulsation regime.
5
4
b 3
2
3 4
d. mm
D-
5
6
Fit. 3. Dependence of the mass-transfer coef-
ficient Ko, the shape factor of the droplet E,
and the, mass-transfer factOr Km on db:' 1, 2)
change in Ko for A packed pulse column with .
Raschig ring [6] and e single droplet moving
in a space in the column, respectively; 3, 4)
change in E =ratilbro) and Km in a pulse -column
with KRIMZ packing, 'respectively.
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For packings that are wetted by the dispersed phase, the nature of the motion of the
breakup of the dispersed phase are rather different. The dispersed phase moves over the
rings as a film; droplets are formed only in the spaces in the rings or in gaps in the pack-
ing. At low intensities, breakup occurs when the dispersed phase flow off the rings, and
at high intensities in the flow. In the latter case, the droplet size can obviously be
described by:
or
-i-/Dfdp=c6we ?
2,6 [4,1(1-6p)]?.4(deciv. P /dp)?' 41
dpidp =C6Wei?.33Rei?.334/(1 ? 8)1?33 X (deq, pidp)? 33.
Thus, to calculate (ID we have:
for sieve-plate pulse columns
for packed pulse columns
(cripc)0.33-0,6113.0.33 (fa) _0,o-4.2).
JD-- to, ',0,33-0,6..0-,-0,5 uar(1,0-1.43)
k /Pc/ P c
(15)
(16)
(17)
(18)
Table 2 summarizes the experimental relations (from published data)_used to calculate
the droplet diameter in pulse columns. The theoretical expressions for dp [Eq. (17) and
(18)] derived from analysis of the mechanism of breakup plainly give adequately reliable
agreement with the experimental relations.
Reciprocating vibration of the reagent (pulsation) is known [12] to increase the operat-
ing efficiency of a column extractor by a factor of 5-10 (Fig. 2). Information regarding
the mechanism of intensification of the process in pulse extractors is conflicting. This
we attribute to the increase in the mass-transfer coefficient when the surface of the droplet
becomes turbulent [13-16] or to increase in the interfacial area [12, 17]. Analysis of the
mass-transfer process in pulse columns requires information regarding the breakup (shape
and size of droplets), the distribution and structure of the flows in the equipment (uni-
formity of phase distribution, longitudinal mixing), and the efficiency. This requirement
has not in general been met in previous studies.
We have examined this problem in an analysis of the hydrodynamic and technological para-
meters of a model pulse column with KRIMZ packing using the system TBP in kerosene + 30-35
g/liter U + 0.1 M HNO3, characterized by an
We used our experimental results on the longitudinal mixing and the efficiency to cal-
culate the HTU due only to mass transfer (HTUm) for different pulsation regimes:
where
HTUm HTU _ HTUadd
HT0add = X..(De/Wt+Dd/Wa).
For this system with amn ","..11, we assumed X = 1. We calculated W* from
wq==T47,/(1--q; mr=wda
(19)
(20)
(21)
Then data on the size and shape of the droplets and the holdup of the dispersed phase gave
the mass-transfer factor, Km = Ke/Ks [17, 18]:
Km (HTU m/HTUmi) (So/Sr) Ke/Ks. (22)
The results of our analysis appear in Table 3, while the effect of droplet size (pulse
intensity) on Km is shown in Fig. 3.
Table 3 shows that the mass-transfer coefficient diminishes with increased pulse in-
tensity, i.e., intensification of the process in pulse extraction columns is due only to
increase in the interfacial area.
The intensification coefficient is (0.45-0.5)K0, which indicates substantial reduction
in the mass transfer cgefficient during pulsation (Fig. 3, curves 1 and 4). This change in
the mass-transfer coefficient arises from change in the shape of the droplet (curves 3 and 4
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in Fig. 3). Indeed when the pulse intensity is increased the droplet size is reduced; the
droplet becomes spherical. In large nonspherical droplets when (ID >
D. sphr (cr/APAT'5*
dD.sphr
(23)
The extractable substance circulates internally; this is responsible for the high values of
K (curve 2 in Fig. 2). The substance penetrates into the centers of the spherical droplet
only as the result of molecular diffusion. The reciprocating vibration of the continuous
phase promotes diffusion of the substance (turbulent diffusion, but does not affect diffusion
in spherical droplets, in which only circulation of the substance in the boundary layer occurs.
NOTATION
HTUi, height of transfer unit, derived with the laboratory column, cm; HTUm, add, height
of transfer unit due to mass transfer; additional component resulting from longitudinal mix-
ing, cm; a, pulse amplitude, width of hole in KRIMZ packing, cm; aD, bp, major and minor
axes of nonspherical droplet, cm; D, coefficient of longtidinal mixing, cm2/sec; Dp, plate
diameter, cm; dp, droplet diameter, cm; dr), mean hydraulic diameter of droplet, cm; dp,
diameter of packing (Raschig ring), cm; do, diameter of hole in plate, cm; deqv.P, equivalent
diameter of channel in packing, cm; dp.s phr, diameter of the droplet at which it assumes spherical
form, cm; ED, energy required for breakup of droplet, erg; F, relative open area of plate, %;
f, pulse frequency, sec-1; HD, column height, m; hp, plate separation, cm; he, height of
emulsion layer at the interface, cm; I, pulse intensity, cm/sec; ief, effective pulse intensi-
ty, cm/sec; K, mass-transfer coefficient; Nhol, number of holes in the packing; Kes, coef-
ficient characterizing the effect of pulsation on the efficiency and interfacial area, re-
spectively; n, ratio of flow rates; S, interfacial area, m2/m3; vo, flow rate of the droplet,
cm/sec; vp, rate of coalescence of droplets at the interface, cm/sec; W, unit load, m3/m2?h;
a, angle of inclination of vanes; am, distribution coefficient; AU, difference between flow
rates in a section of length (ID, cm/sec; c, energy dispersed by vortexes in unit volume of
continuous phase (dissipation energy), erg/cm3; Ep, free volume of packing, m3/m; p, viscosity
cP; v, kinematic viscosity; p, density, g/cm; Ap, difference in densities of the phases, g/
cm3; a, interfacial surface tension, dyn/cm; 0, holdup of dispersed phase, %; Wei = pcI2do/a,
Weber number; Rei = Idopc/pc, Reynolds number.
INDICES
c, d, continuous and dispersed phases; I, 0, value of the parameter in the presence or
absence of pulsation.
LITERATURE CITED
1. S. M. Karpacheva and E. I. Zakharov, in: Development and Use of Pulse Equipment [in
Russian], Atomizdat, Moscow (1973), p. 131.
2. E. I. Zakharov and S. M.,Karpacheva, Tsvetn. Met. No. 2, 53 (1973).
3. A. N. Kolmogorov, Dokl. Akad. Nauk SSSR, 32, No. 1, 19 (1941).
4. A. Fischer, Chem. Rundschau, 26, No. 19, 1 (1973).
5. J. Thornton, L. Smith, and H. Pratt, Trans. Inst. Chem. Eng., 35, No. 4, 292 (1957).
6. V. G. Levin, Physicochemical Hydrodynamics [in Russian], Fizmatgiz, Moscow (1959), p.
453.
7. S. Z: Kagan and Yu. N. Kovalev, in: Liquid Extraction and Chemisorption Processes [in
Russian], Khimiya, Leningrad (1966), p. 43.
8. W. Widmer, Chem.-Ingr.-Tech., 39, No. 15, 900 (1967).
9. S. M. Karpacheva, E. I. Zakharov, and L. F. Kiseleva, Zh. Prikl. Khim. 37, 2668 (1964).
10. T. Misek, Coll. Czech. Chem. Commun., 29, No. 8, 1755 (1964).
11. T. Maychi and H. Oya, AIChE J., 11, No. 3, 395 (1965).
12. S. M. Karpacheva et al., Pulse Extractors [in Russian], Atomizdat, Moscow (1964), pp.
129, 144.
13. S. M. Golovko, V. N. Zadorskii, and N. V. Vasin, Izv. Vyssh. Uchebn. Zaved., Khim.
Khim. Tekhnol., 15, No. 11, 1737 (1972).
14. A. S. Zheleznyak and B. I. Brounshtein, in: Liquid Extraction and Chemisorption Processes
[in Russian], Khimiya, Leningrad (1966), p. 175.
15. A. P. Perovskii and V. G. Kosykh, in: Processes in Chemical Technology [in Russian],
Nauka, Moscow-Leningrad (1965), p. 213.
16. A. D. Vasenev, A. F. Galeev, and A. I. Turtyanov, in: Liquid Extraction [in Russian],
Khimiya, Leningrad (1969), p. 181.
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17. A. M. Rozen, Nauchn. Dokl. Vyssh. Shkol., Energetika, No. 3, 173 (1958).
18. A. M. Rozen, et al., in: Extraction [in Russian], Vol. 2, Gosatomizdat, Moscow (1962),
p. 320.
INVESTIGATION OF THE ADIABATIC EXPANSION. OF WATER VAPOR
FROM THE SATURATION LINE IN LAVAL NOZZLES
Z. K. Karasev, V. V. Vazinger, UDC 621.694
G. S. Mingaleeva, and E. I. Trubkin
The effectiveness of the operation of a jet thermal pump [1] depends on the perfection
of its basic subassemblies, one of which is the nozzle for the expansion of water vapor from
the saturation line into the two-phase region. As the investigations of [2] show, when
thermal differentials of more than 4 kJ/kg are acting in nozzles, i.e., when a velocity of
more than 70-80 m/sec is reached, the greatest acceleration of a flow is possible in Laval
nozzles. The absence of systematic data on the operation of Laval nozzles for saturated
waver vapor with a pressure of more than 30 bars at the entrance has led to the present in-
vestigations. The results of this research can also be employed in the design of equipment
to limit the flow in the case of a rupture of the pipes, for motors operating with a two-
phase medium, and also in other areas associated with the flow of a two-phase flow at high
velocities and pressure gradients.
1 Preliminary investigations have shown that the effectiveness of nozzles operating with
a two-phase mixture is greater for those which have a more elongated shape than for nozzles
intended for the expansion of steam. Fundamental investigations have been conducted on
nozzles with the dimensions and shape given in Fig. 1. The effect of geometrical parameters
(the length of the nozzle entrance, the degree of expansion, and the expansion angle of the
expanding part of the nozzle) and the initial and exit pressures on the value of the critical
flow rate of boiling water and the pressure distribution along the length of the nozzle (see
Table 1) has been studied.
The characteristic picture of the pressure distribution along nozzle 1 for different
back pressures is presented in Fig. 2. When the back pressure is higher than the calculated
value in the expanding part of the nozzle (as in single-phase Laval nozzles) a discontinuity
is established which travels to meet the flow as the back pressure increases. The flow rate
of the saturated water vapor remains constant and does not depend on the back pressure until
the pressure discontinuity enters the throat of the nozzle.
The pressure variation along the nozzle length (Fig. 3) depends on the initial pressure.
Thus as the initial pressure at the end of the cylindrical section of the nozzle entrance
decreases, a weak-gradient flow appears, which indicates completion of the phase transition
processes, offering the possibility of decreasing the length of the nozzle entrance.
A simultaneous measurement of the pressure and temperature on the nozzle wall (Fig. 4)
ascertained a thermodynamic nonuniformity in the initial section. This nonuniformity reaches
%20?C in nozzle 1 with a comparatively short entrance length.
The initial state of the flow has a significant effect on the velocity coefficient of
the nozzlewn. If = 0.75 for a flow expanding from the saturation line or with under-
heating, then this same nozzle operates with (Pn = 0.9-0.95 after its initial throttling by
1-2 bar, which indicates the role of the vaporization nuclei in the thermodynamic completion
of the process. The actual flow rate exceeds that calculated for an equilibrium expansion.
A throttle grid in front of the nozzle decreases the critical flow rate and approximates it
to the equilibrium value. The critical flow rate for a given nozzle depends only on the
Translated from Atomnaya fnergiya, Vol. 42, No. 6, pp. 478-481, June, 1977. Original
article submitted May 31, 1976.
This material is protected by copyright registered in the name of Plenum Publishing Corporation, 227 West 17th Street, New York, N. Y.
10011. No part of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic,
mechanical, photocopying, microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is
available from the publisher for $7.50.
545
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Fig. 1. Geometrical characteristics of nozzles.
p, bars
80
60
40
20
ZO 40 60 80 1, mm
Fig. 2. Pressure distribution along
nozzle 1 for different back pressures.
TABLE 1. GeometryofInvestigatedNozzles
i ozzle
No.
d'a,dt
e
dex
Li
L2
L3
de-
grees
1
2
3
4
23
80
80
80
3,84
19
19
19
7,5
28
28
30
10
114
114
114
18
65
120
87
69
172
86
..210
3
3
6
. 3
*The linear dimensions are given in mm.
p, bars
80
50
40
20
?
.11111111111111111111:
100
200
300 1, mm
Fig. 3. Pressure distribution along nozzle
4 under calculated flow conditions for dif-
ferent initial pressures (the dots correspond
to experimental measurements).
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T, ?C
280
260
240
220
2000 W 40
50
80 1, mm
Fig. 4. Temperature distribution of
the flow in nozzle 1 under the cal-
culated flow conditions: 0) measured
wall temperature; saturation
temperature.
40 50 60 70 80 90 100
Po. bars
Fig. 5. Flow-rate characteristics of the in-
vestigated nozzles: 0, *) nozzles 1 and 2;
A, and 0) nozzles 2, 3, and 4 (a throttle
grid is mounted in front of them).
initial conditions in front of the nozzle. The flow-rate characteristics of nozzles are
presented in Fig. 5.
The information about the effect of the nozzle entrance diameter on the flow-rate char-
acteristics is often contradictory. For example, it is shown in [3] that the specific flow
rate in the case of the discharge of saturated water vapor through cylindrical channels with
sharp entrance edges is lowered as the diameter increases from 3 to 6 mm, but no diameter
effect is detected in [4] for channels 4-8 mm in diameter with a smooth entrance. In addi-
tion, a significant effect of the relative channel length has been emphasized in [3]. In
our opinion (Fig. 6) the absolute length of the channel, which determines the thermodynamic
nonuniformity of the flow at the end of a cylindrical channel or nozzle, has the main effect
on the specific critical flow rate in channels with a round entrance, since the effect of
lid as the hydraulic friction factor is less significant here. No effect of the degree of
expansion on the flow-rate characteristic of the nozzle has been established in the case of
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bo
20
0
20 40 60 80 100
ientr. mm
120
Fig. 6. Dependence of the critical flow
rate of saturated vapor on the nozzle
entrance length: e) based on the data
of [4];0, 11 4, and()) nozzles 1, 2, 3,.
and 4, respectively.
AT, ?C
10
8
6
4
- %
- 20
- 16
- 12
-8
ra.
- 100
- 80
60
40
2 - 4 - 20
0- 0
Fig. 7.
small-expansion angles (within
0
-280
- 240
ZOO
160
- 120
80
40
0
0.)
N I
N.
O1
0
I
0
1:1
C
L.L.I I
0
0 50 100 150
200 1, mm
Variation of the flow parameters cal-
culated for the limiting models.
the limits of 3 and 60).
The size of the pressure head developed by the thermal pump was an indirect estimate of
the effectiveness of nozzle operation. Other conditions being the same, the pressure head
of the thermal pump was greater for a nozzle with an expansion angle of 3?-than for one with
an expansion angle of 6?.
.The results were elaborated for two simplified models. The first one is based on the
following assumptions: the velocities of the phases are equal, the steam is equilibrium, and
the water is nonequilibrium; the second.one is based on the following assumptions: the temper-
atures of the steam and the water, which correspond to the saturation temperature at a given
pressure, are identical, but the velocities of the two phases may be different. It was as-
sumed in both cases that the flow is one-dimensional and the pressure distribution, tempera-
ture, and velocity in each cross section are two-dimensional, and the two-phase medium was
arbitrarily treated as continuous. With these assumptions, the three conversion equations
permit calculating all the basic parameters in each cross section along the nozzle length:
A02.. [(1?x) vdimd -1H xvwv;
io = (1 ?x) cid wa/2) (i" + 14/2);
/I == Qdwd pA-1- F fr ?F w
where Ffr = (w:12dv) A; 1
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The momentum in the first cross section of the nozzle entrance (II = Qw1 + plAI) was
determined from the assumption that the expansion process up to the first cross section occurs
with c? = 0.96 (found from experiment with a nonboiling flow). Therefore, it follows from
the continuity and energy equations that
Q..TilwisA4(1--x)vdd-zvi;
to =0?x) id +xr +14ws / 2
where for=C(Td1-Fc;(7"1,--Td);V;Ivi==,01,18.
Choosing Td, we find wls and x for the given Jo, Q, and sol.
The results of a calculation of the velocity, vapor content, and superheating along the
nozzle on the basis of an experimental determination of the pressure, which refer to nozzle
2 with an initial saturated vapor pressure of 80 bar, are presented in Fig. 7. The first
pressure sampling is taken as the origin of coordinates. The velocity increases to 280 m/sec
and the vapor content to 0.187 in the case of the expansion of boiling water in a nozzle
from 80 to 18.5 bar. The maximum possible superheating of the liquid phase reaches 9.4?C,
and toe maximum possible ratio of the velocities of the two-phases (m = wv/wd) is nearly
constant along the length and is equal to 1.12. Calculations carried out in a similar fashion
in the initial pressure range from 30 to 100 bars show that the maximum possible value of
this velocity ratio varies from 1.08 to 1.15 and the maximum possible superheating of the
liquid phase at the end of the nozzle increases from 2 to 12?C as the initial pressure in-
creases, while the coefficient Son is within the limits 0.92-96. Comparison of the velocity
at the end of the nozzle entrance with the calculated speed of sound based on the equations
of [5-7] has shown good agreement within 10% limits with the recommendations of'[5], which
have been obtained for an equilibrium homogeneous two-phase medium. The flow velocity in
the expanding part of the nozzle is greater than the speed of sound, and it is 1.3 times
greater at the nozzle cutoff than the local speed of sound calculated from the formula of [5].
It is possible on the basis of the investigations performed to offer the following re-
commendations on the design of effective nozzles for the expansion of water vapor from the
saturation line: nozzles should have an elongated entrance but no longer than 100 mm; it is
desirable to produce vapor formation nuclei in front of the nozzle in the form of fine vapor
bubbles; the expansion angle of the expanding part should be no greater than 3?.
CONVENTIONAL NOTATION
Q, mass flow rate, kg/sec; q, specific flow rate, kg/(m2.sec); A, area of the straight-
through cross section, m2; p, pressure, N/m2; w, velocity, m/sec; x, mass vapor content; i,
specific enthalpy, J/kg; s, specific entropy, J/(kg.deg K); v, specific volume, ms/kg; cp,
specific heat of water vapor on the saturation line at constant pressure, J/(kg.deg K); F,
force, N; Fw, reaction of the wall, N; I, momentum, N.
SUBSCRIPTS
0, initial state; 01, at the nozzle entrance; 1, at the first pressure sampling; d, v,
drip and vapor media; thr, throttling; fr, friction; s, isoentropic flow (process).
LITERATURE CITED
1. E. K. Karasev, in: Problems of Atomic Science and Engineering. Reactor Construction
Series. No. 2, Izd. TsNIIatominform, Moscow (1973), p. 3.
2. Boiling Adiabatic Flows [in Russian], Atomizdat, Moscow (1976).
3. V. D. Keller, V. K. Malltsev, and D. A. Khlestkin, in: Digest of Lectures at the Fifth
All-Union Conference on Heat Exchange and Hydraulic Drag upon the Motion of a Two-Phase
Flow in the Units of Energy Machines and Apparatus [in Russian], Leningrad (1974), p. 235.
4. K. S. Polyakov, in: Transactions of the LPI. Turbomachine Series, No. 247, Mashino-
stroenie, Moscow-Leningrad (1965), p. 16.
5. V. V. Sychev, Inz.-Fiz. Zh., No. 6, 64 (1961).
6. V. V. Dvornichenko, Teploenergetika, No. 10, 72 (1966).
7. M. E. Deich and G. A. Filippov, Gasdynamics of Two-Phase Media [in Russian], Energiya,
Moscow (1968).
549
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THE ORIGIN OF THE TRACKS OF FISSION FRAGMENTS IN
WHITLOCKITE FROM THE BJURBoLE METEORITE
V. P. Perelygin, S. G. Stetsenko,
and N. Bhandari* UDC 523.165
Searches for the tracks of far transuranic elements in minerals from meteorites and
from the surface of the moon can be carried out in two different ways. First of all, there
are the searches and the identification of tracks from the stopping of heavy nuclei of the
primary cosmic radiation in the region Z 90 up to Z = 110-114 [1-3]. Such nuclei produce,
during their retardation prior to stopping, extended zones of defects in crystals whose length
is proportional to their atomic number [2, 3]. However, due to the fact that heavy and
superheavy nuclei have relatively short ranges in the material of meteorites [3, 4] and a
relatively high probability of nuclear interaction with decelerating medium [5], it is
necessary for the conduct of such searches to expose sections situated near the preatmospher-
ic surface of the meteorites at a depth of greater than 5-3 cm [3]. The trend in searches
is the detection and identification of the tracks of fragments from the spontaneous fission
of heavy transuranic elements synthesized about 4.6 billion years ago during the formation
of the solar system [6, 7].
The conditions of outer space have facilitated the preservation of the tracks of spon-
taneous fission of nuclei in minerals from the time of the cooling of the parent bodies of
the meteorites, i.e., for 4-4.5 billion years [6, 7]. Thus, the minerals from meteorites
are more suitable objects for the search for the effects of spontaneous fission of relatively
short-lived transuranic nuclides than are crystals and glasses of terrestrial origin, whose
track age does not exceed 2 billion years [8, 9].
This article was undertaken in connection with investigations [10] on a search for far
transuranic elements in samples of terrestrial minerals, ores, rocks, and also in meteorite
samples.
The BjurbOle meteorite belongs to the stony meteorite group ? the L-chondrites of petro-
logical type IV with spherical olivine-hypersthene chondrules. Its fall occurred in the
spring of 1899 in Finland. The mass of the Bjurbnle meteorite was about 330 kg [11], and
during its fall it broke up into a large number of fragments. The age of this meteorite
(8-11 million years) was determined from cosmogenic isotopes in 1960 by Eberhardt and Hess
[12]
As follows from the analysis of the preatmospheric masses of the St. Severin, Keyes,
Allegan, and Pi'ibram stony meteorites [13, 14, 15] carried out by the method of dielectric
track detectors, the losses of material upon passage through the terrestrial atmosphere amount
to from 20 to 70%. The data of Lavrukhina et al. [16] for the Pfibram chondrite show that
the degree of ablation can exceed 99%.
Taking account of the effects of ablation and crushing during passage through the dense
layers of the atmosphere, it is possible to conclude that it is extremely difficult to de-
tect in stony meteorites surface sections which have a track density for nuclei of the iron
group of more than 106 per 1 cm2. It has been shown in [17] that the most suitable indicator
of the depth of location of samples away from the preatmospheric surface is olivine. Actual-
ly, since the uranium content in olivine does not usually exceed 10-1?-10-11 g/g, it contains
practically no background of tracks from spontaneous fission. This permits one to determine
*Laboratory of Physical Research, Ahmedabad, India.
Translated form Atomnaya Energiya, Vol. 42, No. 6, pp. 482-485, June, 1977. Original
article submitted August 2, 1976.
550
This material is protected by copyright registered in the name of Plenum Publishingrorporation, 227 West 17th Street, New York, N.Y.
10011. No part of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any meant electronic,
mechanical, photocopying, microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is
available from the publisher for $7.50.
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Declassified and Approved For Release 2013/04/01: CIA-RDP10-02196R000700090006-3
with a high degree of reliability the contribution produced by fast cosmic nuclei to the
track density in the crystals of feldspars, pyroxenes, phosphates and so on which are ad-
jacent to the olivine.
_Samples of the BjurbOle meteorite with a total mass of 15 g selected from one spot were
ground into a powder with particle size less than 1 mm. Transparent homogeneous crystals
were sorted out from the powder under a stereomicroscope at a magnification of 20-25x. In
all, about 200 transparent crystals from 100 to 600 microns in size were selected, which were
mounted in an epoxy resin and polished.
Since olivine is one of the principal components making up the composition of the chond-
rules of this meteorite, the etching of the mounted samples was carried out in a special solu-
tion suggested in 1971 by Krishnaswami et al. [18] for revealing tracks in olivine. The
crystals along with the etchant were enclosed in a hermetic Teflon container which excluded
the possibility of variation of the composition of the solution during the chemical process-
ing. The temperature of the etching solution was 110?C, and the etching time was 6-8 h. The
developed samples were examined under the microscope at a magnification of 1400x. It did
not prove possible to detect in the olivine crystals tracks from heavy cosmic nuclei or
fission fragments. Only the defects of the structure of the olivine were revealed ? capil-
lary inclusions, dislocations, and microcracks. The upper limit to the track density of
heavy charged particles in the olivine was< 5 .102/cm2. It is possible to conclude on the
basis of the known cosmic age of the BjurbUle meteorite and the limit obtained on the track
density that the depth of location of the investigated sample was 2)
where the indices 1, 2, and 3 denote quantities pertaining to the angles T1,
T2, Ts. ?
(4)
The variations of the quantities on the right-hand side of Eq. (4) can serve as a con-
trol On the error of the method.
To determine the temperature one can use, e.g., two opposing light beams of frequency wo = 2w0
with recording of the lines 5/2wo and 3wo . The hydrodynamic corrections can be eliminated if one
measures the splitting of the lines in the direction of one of the probe beams (1 cos .61.1 = 1,
!cos q) 1 = 0):
(0=5/2(00; A)j),, = 471 ? 10-2T;
3w0; Aka =1.8.10-2T.
(5)
(6)
*The question of the absolute value of the shift of the CS lines goes beyond the framework of
the model adopted.
tHere and later the estimates are made for a laser frequency wo = 2.1015 sec-1 (neodymium
laser).
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When T 1 keV the splitting comprises '?,50-150 I and is easily found.
Probing of the plasma with a light beam having a variable frequency can also give in-
formation on the spectrum of the Langmuir oscillations.
In experiments on the heating of a plasma by a laser pulse the choice of diagnostic
methods is very limited (direct methods of measurement of the velocity of dispersion of the
plasma, e.g., are entirely absent). Therefore, the study of the fine structure of CS lines
can be an important source for obtaining data on the parameters of the plasma and the mech-
anism of its interaction with the laser radiation. At the same time, there is no doubt that
the arguments presented are in serious need of experimental verification.
i thank N. G. Koval'skii and M. I. Pergament for a discussion.
LITERATURE CITED
1. J. Bobin et al., Phys. Rev. Lett., 30, 594 (1973).
2. H. Pant et al., Opt. Communs, 16, No. 3, 396 (1976).
3. V. P. Silin, The Parametric Interaction of Powerful Laser Radiation with a Plasma lin
Russian], Nauka, Moscow (1974).
4. V. D. Shafranov, in: Problems of Plasma Theory [in Russian], Part 3, Gosatomizdat,
Moscow (1963), p. 3.
USE OF THE CALORIMETRIC METHOD TO MEASURE THE ENERGY
RELEASE IN A COMPENSATING ROD
A. S. Zhilkin, V. P. Koroleva,
N. P. Kurakov, L. A. Chernov,
and E. V. Shestopalov
UDC 621.039.517
The development and refinement of experimental methods of determining the energy release
in the compensating rods of a reactor are urgent for the practical work of reactor construc-
tion and for the development of calculating methods. To obtain greater reliability it is
desirable to measure the energy release by several independent methods. Along with the re-
lative physical method described in [1], in the present report a highly sensitive calori-
meter is proposed and used for the measurement of the energy release in a boron-containing
rod of a critical installation. Despite the wide reputation of the calorimetric method, it
is used mainly for dosimetric purposes [2] and for the determination of heat release in spec-
mens under the conditions of power reactors [3] because of its relatively low sensitivity.
Sensitive calorimeters (10-4 W/g) have recently been created which have made it possible
to measure the energy release in fissionable materials in a critical installation [4]. And
only the appearance of such sensitive temperature detectors as microthermistors has made it
possible to increase still more the sensitivity of the calorimetric method (up to 10-6 W/g).
The aim of the present report is the creation of a highly sensitive calorimeter, the
measurement of the energy release in an absorbing rod of a critical installation, and the
comparison of the measurement results of the calorimetric and physical methods.
Calorimetric Method. The proposed highly sensitive calorimeter belongs to the quasi-
adiabatic type. The principle of its operation is based on the measurement of the rate of
change of the temperature of a specimen isolated from the external medium. The temperature
of the specimen varies by an linear law during a certain time, which depends on the thermal
insulation of the specimen, its geometrical dimensions, and other factors. The cylindrical
specimen of boron carbide 19.2 mm in diameter and 30 mm high was insulated from the other
parts of the rod by foam plastic and by an air gap 1 mm thick, and from the lateral walls
Translated from Atomnaya Energiya, Vol. 42, No. 6, pp. 502-503, June, 1977. Original
article submitted September 20, 1976.
576
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10011. No part of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic,
mechanical, photocopying, microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is
available from the publisher for $7.50.
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Fig. 1. Diagram of calorimeter
for measuring radiation energy
release in a boron carbide rod:
1) holder (aluminum); 2) outer
shield (aluminium); 3) insulator
(foam plastic); 4) cylindrical
specimen of boron carbide; 5)
MT-64 thermistor; 6) remaining
part of absorbing rod.
of the reactor channel by an air gap of 5-7 mm. During heating of the specimen in the re-
actor through radiation energy release the length of the linear temperature rise was 150-
200?C. This is also confirmed by calculations. With more prolonged heating the heat ex-
change between the specimen and the surrounding medium shows up, which disturbs the linear
rise of the specimen temperature. For a linear temperature rise with time the specific
energy release in the specimen per unit time owing to the (n, a) reaction in boron and the
absorption of the y and R radiation of the reactor is determined from the equation
Q=CpdTIT,Vdt W/4rWL (1)
where C is the specific heat capacity of the specimen; dT/dt is the rate of change of the
specimen temperature; W is the reactor power. Thus, to determine Q one has to measure dT/dt,
C, and W. ?
P
A diagram of a colorimeter for measuring the radiation energy release in a boron carbide
rod placed at the center of a PF-4F8 critical installation [5] is presented in Fig. 1. The
temperature detector, and MT-64 thermistor (R = 64.5 k at 20?C), was fastened to the lateral
surface of the thermally insulated specimen. The sensitivity threshold of the calorimeter
was (1-2)*10-6 W/g, and a power of 10 W liberated in the active zone of the installation was
enough to perform the measurements.
In order to allow for the effect of heat exchange between the surrounding medium and the
specimen on the rate of temperature change, the quantity dT/dt was measured during three
time periods with the same duration of 150-200 sec, corresponding to different reactor states:
dT/dt = (dT Idt)m- 1/2 ((dTldt)i?(dTldt)fl, (2)
where (dT/dt) i is the rate of change of the temperature of a specimen placed in the reactor
in the initial period; (dT/dt)m is the rate of temperature change in the main period, when
dynamic equilibrium is established in the heat exchange between the specimen and the surround-
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ing medium; (dT/dt)f is the rate of temperature change in the final period after the power
drop.
The specific heat capacity of the boron carbide was determined with the help of an
electric heater of cons tantan wire q.0.19 mm in diameter wound on the surface of the boron
carbide specimen. The mass of the heater and the thermistor comprised 't,2% of the mass of
the specimen. The correctness of such a method of determination of Cp was verified experi-
mentally on aluminum.
It was shown through calculation that when the boron carbide specimen was heated by a
surface heater the temperature at the center of the specimen becomes equal to tne temperature
at the surface 5-10 sec after the start of heating. During the heating time of 150-200 sec
the heat losses from the specimen to the external medium do not exceed 5%. The power W of
the installation was determined by measuring the specific energy release in the reactor fuel
with the calorimeter and subsequent integration over the volume of the active zone.
Knowing all the quantities in Eq. (1), one can calculate the radiation energy release
in the boron carbide: Q = (1.05?0.1)-10-6 W/(g-W). The standard error for several series of
measurements is presented here. To compare the results of the calorimetric and physical
methods, in which we measured the energy release in natural boron and only that due to the
(n, a) reaction, we made a recalculation in which we allowed for the boron content in B4C
the (n, a) reaction to the total
(0.737) and through calculation obtained the contribution of
energy release of (1.29 ? 0.13).10-6 W/(g.W).
Physical Method. In this case the ratio of the cross section for 16B capture and 235U
fission, averaged over the reactor spectrum at the monitoring point, was determined with a
semiconducting gold-silicon detector. The coefficient of nonuniformity of the energy distri-
bution was measured at the same point using a KNT-5 ionization chamber. The ratio of the
numbers of neutron captures at the monitoring point and in the boron-containing rod (averaged
over the cross section of the rod) was obtained by the method of track detectors (nitro-
cellulose with a layer of boron). All the experiments were conducted at the middle of the
height of the active zone. In contrast with [1], we additionally studied and introduced a
correction for the effect of the empty channel on the cited ratio of captures in boron. The
value of the specific energy release has the form
Q = (1.34 ? 0.03) 10-6 W/ (g-W).
Thus, the results obtained by the two independent methods agree well within the limits
of the experimental errors, which confirms their correctness and the possibility of using
such calorimeters in physical installations.
The authors express deep gratitude to A. A. Kutuzov, Yu. G. Pashkin, and Yu. E. Shvetsov
for helpful advice and some calculations.
LITERATURE CITED
1. V. A. Kuznetsov et al., At. Energ., 33, No. 5, 926 (1972).
2. V. M. Kolyada et al., ibid., 21, No. 6, 520 (1966).
3. B. Lewis, Nucl. Sci. Eng., 18, 76 (1964).
4. 0. A. Gerashchenko, V. B. Klimentov, and A. V. Nikonov, At. Energ., 33, No. 3, 232
(1972).
5. A. I. Mogil'ner et al., ibid., 24, No. 1, 42 (1968).
6. S. Genna et al., At. Energ. Rev., 1, No. 4, 239 (1963).
578 ?
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BEHAVIOR OF CHROMEL-ALUMEL THERMOCOUPLES DURING AN
EMERGENCY SHUTDOWN OF THE BR-10 REACTOR
A. S. Kruglov, P. V. Vyrodov, UDC 621.039.531
and M. I. Redchenko
In measuring the temperature, with thermocouples, in the zone of action of the intense
radiation of a reactor, the allowance for the influence of radiation effects, which can dis-
tort the thermocouple readings, acquires importance. Despite the considerable number of
published reports on this question, there is no single opinion concerning the existence of
an instantaneous effect of the action of the radiation field on the thermo-emf. For example,
some authors have not observed instantaneous deflections in the reading of Chromel-Alumel
(CA) thermocouples during a sharp drop in reactor power, with the initial flux density of
fast neutrons reaching '?.1013 neutrons/(cm2?sec). In other cases instantaneous deflections
in the readings of CA thermocouples reaching -6.6 and -10?C were observed during the shutdown
of a reactor with a flux density of fast neutrons of '1,1013 and q6.501012 neutrons/(cm26sec),
respectively [1]. For the SM-2 reactor with a flux density of fast neutrons of '1,4?1014
neutrons/(cm2.sec) the instantaneous deflection for CA thermocouples reached +26?C [2]. Such
disagreement of the experimental results casts doubt on the reliability of thermometry using
thermocouples under the conditions of the active zones of reactors.
In order to clarify the behavior of CA thermocouples during sharp variations in the
power of a fast-neutron reactor we conducted tests on thermometry during a shutdown of the
BR-10 reactor. The thermodynamic assembly, for which a massive specimen of stainless steel
with two thermocouples of KTMS (CA) cable with a cross section of 2 x 0.06 mm mounted inside
served as the basis, was placed in a vertical channel of the BR-10 reactor located in a nickel
shield at a distance of q,40 cm from the axis of the central zone. In addition, five more
analogous thermocouples were built into different elements of the assembly. The hot junc-
tions of the thermocouples had thermal and electrical contact with the thermocouple cover
and with elements of the assembly. The cold junctions were located outside the reactor and
the connecting lines were made of CA conductors. During the operation of the reactor the
specimen was heated to a temperature of 215?C, which exceeded by 50-70?C the temperature of
the channel wall, separated from the assembly by an air gap. The thermo-emf was measured with
an FZO digital ampere voltmeter with a sensitivity of 10 pV. The measuring instrument was
connected with the thermocouples through an automatic commutator device allowing measure-
ments to be made with an interval of 1 sec. The FZO readings were recorded on an MP-16
digital printer. Up to the time of the test all the thermocouples were irradiated by a
neutron flux of from '?,1.24,1020 to '1,1.24.1021 neutrons/cm2 distributed along a length of '?,0.5m
(in the zone of the hot junction). A temperature drop reaching 'k/00?C was observed over the
same length, with the maximum of the heating falling in the region of the hot junction.
The thermo-emf of the CA thermocouples was measured during a planned shutdown of the
reactor through the actuation of the quick-acting accident protection (QAP). Before the
actuation of the QAP the reactor power was 3 MW, the neutron flux density in the zone where
the specimen was located was '\,4?1013 (for E 0.1 MeV) and '\,701013 neutrons/(cm2?sec) (for
all energies) at this power, and the dose rate of y radiation was 103 R/sec. The sharp drop
in reactor power is not equivalent to the complete "turning off" of the radiation field,
since y radiation from the radioactive fission products of the fuel and activation products
in the construction materials of the active zone, the shield, and the coolant remains after
the reaction ceases. According to the calculated and experimental data the dose rate of y
radiation is reduced by more than an order of magnitude as a consequence of the cessation
of the fission reaction.
Translated from Atomnaya Energiya, Vol. 42, No. 6, pp. 504-505, June, 1977. Original
article submitted October 4, 1976.
This material is protected by copyright registered in the name of Plenum Publishing Corporation, 227 West 17th Street, New York, N. Y.
10011. No part of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic,
mechanical, photocopying, microfilming, recording or otherwise, without written permission of the publisher. A copy of this articla is
available from the publisher for $7.50.
579
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TABLE 1. Variation of Thermo-emf during
and after Activation of QAP (T = 0)
T,sec
E,mV
x,sec
E, mV
;sec
EanV
Tsec
E,nfV
--120
8,06
8
8,05
18
8,05
28
8,04
--60
8,05
9
8,07
19
8,06
29
8,04
0
8,07
10
8,06
20
8,06
30
8,04
1
8,05
11
8,05
21
8,05
40
8,03
2
8,06
12
8,06
22
8,04
50
8,03
3
8,06
13
8,05
23
8,06
60
8,00
4
8,05
14
8,05
24
8,05
120
7,98
5
8,07
15
8,06
25
8,04
180
7,87
6
8,06
16
8,06
26
8,04
240
7,74
7
8,06
17
8,05
27
8,05
300
7,58
The data obtained showed that after the activation of the QAP none of the seven thermo-
couples changed their readings within the limits of the experimental accuracy. The character
of the variation of the thermo-emf of one of the thermocouples mounted in the specimen is
presented in Table 1. The instability of the thermo-emf does not exceed 20 UV, and the
relatively smooth decrease in the thermo-emf only begins 60 sec after the activation of the
QAP and is obviously connected with the natural cooling of the specimen.
On the basis of these data, one can conclude that a fast shutdown of the BR-10 reactor
does not cause instantaneous deflections in the readings of cable CA thermocouples exceeding
30 pV (less than a degree). Possible distortions in the readings connected with the uneven
buildup of radiation defects in the thermocouple materials and with the presence of a temper-
ature gradient along the irradiated section can obviously affect the values of the measured
thermo-emf only upon a change in the temperature gradient and cannot have a significant
effect on the character of its variation at the moment of a discontinuity in the intensity
of the radiation field.
Thus, the results of the experiment show that instantaneous effects of variation in
the thermo-emf of CA thermocouples, the theoretical possibility of which may be connected with
a change in the Fermi level of electrons in metals under the effect of a radiation field [3],
are not detected under the conditions of irradiation in the BR-10 reactor.
LITERATURE CITED
Temperature Measurements in Nuclear Reactors [in Russian], Atomizdat,
Samsonov, and V. A. Tsykanov, Fiz. Met. Metallov., 22, No. 4, 747
and S. Keeton, Nucl. Appl., 5, Nb. 5, 322 (1968).
1. B. V. Lysikov et al.,
Moscow (1975), p. 81.
2. N. V. Markina, B. V.
(1971).
3. G. Dau, R. Bourassa,
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TOTAL NEUTRON CROSS SECTIONS OF 236Th IN THE THERMAL ENERGY REGION
AND 236Ra UP TO 1 keV
R. N. Ivanov, S. M. Kalebin,
V. S. Artamonov, G. V. Rukolaine,
N. G. Kocherygin, S. I. Babich,
S. N. Nikol'skii, T. S. Belanova,
and A. G. Kolesov
UDC 621:039.556
The results of measurements of the total neutron cross sections at for 235Th and 226Ra
by the transit-time method are presented in the report. The measurements were carried out on
the reactors of the Institute of Theoretical and Experimental Physics and the Scientific-
Research Institute of Atomic Reactors using a selector with synchronously rotating rotors
suspended in a magnetic field [1, 21. Counters filled with 3He served as the neutron de-
tectors. The best resolution of the spectrometer was 35 nsec/m and the background did not
exceed 4%.
Thorium 230. The specimen was prepared from Th02 powder in which the ratio of the
amount of 232Th to 236Th was equal to 1.463 ? 0.03. The impurities in % were: Nd < 1.10-5,
La < 1.10-5, Ce < 14,10-5, Pr < 1.10-5, Sm < 1410-5, Eu < 1.10-5, Gd < 2.10-6, Zr < 5.10-3.
The measurements were made in the thermal region of neutron energies on a specimen with
a 236Th content of (4.23 ? 0.05) .1021 atoms/cm2. The transmission was measured relative to
a specimen in which the amount of 232Th02 was equal to its content in the thorium specimen
being studied. The data were corrected for the cross section of the oxygen [3] which was
directly bound with the 236Th.
The results of the measurement of the total cross section for 235Th are presented in
Fig. 1. The values of at calculated from the parameters of the positive resonance taken
from [4] are also presented here. At the thermal point the experimental value is at = 54 ?
3 b while the calculated value is 31.8 b. A value of at = 71.8 b was obtained at the thermal
point in [4], where the scattering of slow neutrons through small angles was not taken into
account. In a calculation of the 1.431 eV resonance by the shape methodfrom the Breit-Wigner
one-level equation for 236Th we obtained a value of 16 ? 4 b for the potential scattering
cross section. If one assumes that neutron capture is the main reaction for 236Th in the
thermal energy region then from the data obtained one can determine the value of this cross
section at an energy of 0.025 eV, aa = (38 ? 5) bi which is considerably greater than the
value of aa = 23.2 ? 0.6 b recommended in the atlas [3].
The data published on the cross section aa at the thermal point have a scatter of from
21.5 to 61 b [5-9].
Radium 226. We studied RaSO4 powder which contained the following impurities: Na2SO4 1%
and BaSO4 4.75%. The measurements were made in the energy region of 30-1400 eV on specimens
with a 226Ra content of (5.56 ? 0.53).1021 and (8.12 ? 0.78)
41026 atoms/cm2.
We found 38 resonances. For 31 of the resonances we determined the values of the neutron
widths Fri by the area method and for five of them we determined the values of the radiation
widths Fy. The average value of Fy = 25 ? 2 meV was used in analyzing the other resonances.
The resonance parameters of the levels and the measured resonance [10, 11] with an energy of
0.539 eV, the parameters of which are taken froth [10], are presented in Table 1.
A diagram of the distribution of the number of neutron levels as a function of the
energy is presented inFig. 2. It follows from the diagram that the omission of levels begins ,
with 950 eV. Up to this energy we calculated the following': D = 32 ? 3 eV; r = 3.2 ? 0.8
Translated from Atomnaya Energiya, Vol. 42, NO. 6, pp. 505-506i June, 1977. Original
article submitted October 4, 1976.
This material is protected by copyright registered in the name of Plenum Publishing Corporation, 227 West 17th Street; New York, N.Y.
10011. No part of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic,
mechanical, photocopying, microfilming, recording or otherwise, without written permission of the publisher. A cop); of this article is
available from the publisher for $7.50:
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Fig. 1
Fig. 1. Total neutron
cross section
Number of resonances
; :
00 470 ;00 ;170 1;00 /00 E, eV
Fig. 2
for 230Th: __ -) calculated dependence of
at with allowance for positive resonances; 40) experiment.
Fig. 2. Diagram of distribution of number of neutron resonances as a function
of energy for 226Ra.
TABLE 1. Parameters of Neutron Resonances of 226Ra
E0, eV
rr meV
r,, meV
E0, eV
r?, meV
Eo, eV
rn, meV
0,539-F0,003
26,5+3,0
0,015+0,001
377,3-E1,6
88+24
809+6
13-E8
35,844-0,15
0,054-0,01
398,6=F1,8
187=F41
8284-6
57+31
39,95:F0,15
26,4?2,6
0,36+0,04
459+2
244-11
887-E7
68-E37
55,80+0,18
22,4=F5,9
6,0=F0,9
471:F3
234-11
907+7
280-F100
88,5+0,3
19,24-7,0
21,24-2,4
516+3
26-E11
9454-7
47+22
155,3+0,7
2,04-0,4
526-F3
40-E17
1007+8
217,8=1:0,9
1,4-E0,4
548+3
54-3
10294-8
237,7-E1,0
31,2-E7,6
1384-17
5854-4
114-4
10784-8
259,8+1,1
2,4-E0,8
630+4
422-E33
1166-E10
263,0-E1,1
15,2-E5,4
6694-4
33-F17
1222+10
292,3+1,4
53-E14
681-F5
21-E11
1336-E13
329,4-F1,4
1534-45*
745+5
98T60
13851-13
348,2+1,5
243-E61
7854-6
67?29
meV; So = (1.0 ? 0.3)610-4, and I = 236 ? 16b.
A statistical analysis of the data obtained for the neutron resonances below 950 eV
shows that the distribution of distances between levels coincides with a Wigner distribution
while the reduced neutron widths follow a Porter-Thomas distribution for one degree of free-
dom. The experimental value of Ag = 0.48 (Dyson-Mehta statistics [12]) agrees with its
theoretical value of 0.35 ? 0.11.
LITERATURE CITED
1. S. M. Kalebin et al., Prib. Tekh. Eksp., 3, 79 (1970).
2. S. M. Kalebin et al., in: Proceedings of Conference on Neutron Physics [in Russian],
Vol. 2, Naukova Dumka, Kiev (1972), p. 267.
3. Brookhaven Nat. Lab.-325, 3rd ed., New York (1973).
4. S. M. Kalebin et al., At. Energ., 26, No. 6, 507 (1969).
5. A. Jaffey and E. Hyde, Argonne Nat. iab.-4249 (1949).
6. E. Hyde, Argonne Nat. Lab.-4183 (1948).
7. H. Pomerance, Oak Ridge Nat. Lab.-1620 (1953).
8. M. Cabell, Canad. Phys., 36, 989 (1958),
9. R. Attre, ibid., 40, 194 (1962)..
10. S. M. Kalebin et al., Yad. Fiz., 14, 22 (1971).
11. M. I. Pevzner et al., At. Energ., 1, No. 4, 67 (1956).
12. E. Dyson and M. Mehta, Mat. Phys., -4, 701 (1963).
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TOTAL NEUTRON CROSS SECTIONS AND RESONANCE PARAMETERS
OF 1752176Lu UP TO 200 eV
S. M. Kalebin, V: S. ArtamonoV,
R. N. Ivanov, G. V: Rukolaine,
T. S. Belanova, A. G. Kolesov, and
V. A. Poruchikov
UDC 621.039.556
The results of measurements of the total neutron cross section at for 175'1761.11 by the
transit-time method are given in the report. The Measurements were carried out on the re-
actors of the Institute of Theoretical and Experimental Physics and the Scientific-Research
Institute of Atomic Reactors using neutron choppers with magnetic suspension of the rotors
[1, 2]. Counters filled with 3He served as the neutron detector. The best resolution was
35 nsec/m, the statistical accuracy was kept within the limits of 0.5-1.5%, and the neutron
background did not exceed 4%.
Lutetium 175. We studied Lu203 powder which contained the following impurities by %:
176Lu 0.23; Yb < 0.1; Fe, Cu, Pb < 0.01; Si < 0.02; Al, Ca < 0.05. The measurements were
conducted in the region of energies up to 200 eV for two thicknesses of specimens containing
0.14.1022 and 0.76.1022 atoms/cm2 of 175Lu.* The total neutron cross section in the thermal
energy region was corrected for the contribution made by the 0.14-eV resonance of the impurity
isotope 176Lu. The calculated curve of the total neutron cross section determined by the
positive resonances of 175Lu ispresentedinFig. 1. The total neutron cross section of 175Lu
at the energy of 0.025 eV is equal to 30 ? 1 b. The parameters of the neutron resonances
up to an energy of 57 eV are found by the shape method; the parameters above this energy are
found by the area method.
The positions of the resonances (E0) and the values of theneutron width (2grn) and
tne total width (r) are presented in Table 1. Analogous data are contained in [3, 4].
Simultaneously with the calculation of the parameters of the resonances by the shape
method we found a = 7.5 ? 1.0 b for 175Lu. If one assumes that neutron capture is the main
reaction in the narrow energy region then from the values of at and al) obtained the cross
section for this reaction at the energy of 0.025 eV is ay = 22.5 ? 1.5 b. It agrees well
with ay = 23.4 ? 2 b recommended in the atlas [3]. .
From the data presented in Fig. 1 we determined the parameters of the negative resonance
(see Table 1). A detailed analysis of the transmission curves in the region of 14 eV is
interesting. A resonance is cited in the published reports for Eo = 13.83 eV with the
parameters 2g1'n = 20 ? 2 meV [3, 4] and a spin I = 3 [5]. In the determination of the para-
meters of this resonance for a specimen with a thickness of 0.14.1022 atoms/cm2 the trend of
the theoretical curve (Fig. 2a, dashed curve) clearly indicated that the resonance with an
energy of 14 eV should be considered as double, consisting of two levels which have merged
because of thegreatthickne?softhespecimen. With suchariassumption the subsequent calcula-
tions revealed two resOnances at E0 = 13.93 and 14.17 eV with the parameters 2grn = 761
and 6.8 meV, respectively, which described the experimental data well (see Fig. 2a). In
order to conclusively ascertain, the existence of two levels in the region of 14 eV, we mea-
sured the transmission with a thinner specimen with a thickness of 0.608.10 21 atoms/cm2. Both
resonances are well seen on the transmission curve of the thin specimen (dots on Fig. 2b). The
*The measurements were conducted under conditions for which the effect of neutron scattering
through small angles was excluded.
Translated from Atomnaya Energiya, Vol. 42; No. 6, pp. 506-509, June, 1977. Original
article submitted October 4, 1976.
This material is protected by copyright registered in the name of Plenum Publishing Corporation, 227 West I 7ih Street, New York, N.Y.
10011. No part of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic,
mechanical, photocopying, microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is
available from the publisher for $7.50.
583
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TABLE 1. Parameters of Neutron Resonances of 1 7 6Lu
E0, eV
r, meV
2grn, meV
Eo, eV
2grn, meV
?1,65
201=-0,14
96,7-P0,3
54+5
2,621+0,007
63+2
0,229+0,006
99,8+0,4
18,0+0,3
4,78+0,01
65+3
0,318-F0,086
101,2+0,4
5,5+1,4
5,22+0,01
80+9
1,14-F0,18
103,1+0,4
9,3+0,8
11,27+0,06
88+8
3,13+0,50
107,1+0,4
41,6+3,2
13,93+0,07
74+4
7,83=F0,36
*109,5+0,4
0,5+0,3
14,16+0,07
82+6
8,12-F0,37
113,0+0,5
4,1+0,4
15,38+0,07
73+4
1,75=F0,06
115,3A-0,5
35,1+2,5
20,54+0,08
87+6
2,22-F0,26
118,2-F0,5
8+5
23,52+0,09
96+6
6,1=F0,7
127,5-F0,5
38,3+5,0
28,02+0,12
89+5
1,80-F0,12
129,8-F0,5
53,3+5,2
30,20+0,13
103+9
7,19471,20
138,0+0,6
41,4+3,1
31,10+0,13
94+9
3,05-F0,28
143,0+0,7
2,2+0,7
36,58+0,15
126+8
6,22T1,20
146,3-F0,7
3,4+0,8
40,6540,17
150+5
13,2-F3,1
148,7-F0,7
1,2+0,3
41,16+0,17
148+12
3,50:1=0,21
151,1+0,7
3,2+0,7
49,48+0,18
156+30
11,6=111,6
155,5+0,7
4,2+1,4
50,35+0,18
136+16
7,95-F1,15
158,4-F0,7
10,1+3,5
53,58+0,18
0,47-F0,07
163,9+0,7
9,5+2,6
56,88+0,18
4,22-F0,63
169,2-P0,8
8,4+2,9
61,15+0,19
0,43-F0,07
171,2-F0,8
3,3+1,0
69,44+0,19
0,24+0,12
69,97+0,19
3,2=F0,9
175,3-F0,9
24+9
74,0+0,2
0,25-F0,07
180,8A-0,9
12,4+4,8
81,1+0,2
0,15=11=0,05
185.1+0,9
42+15
85,2+0,3
5,85+1,20
192,8+0,9
54+16
88,2+0,3
4,3-F0,5
195,8-F0,9
5,4+1,5
Doubtful resonances.
TABLE 2. Parameters of Neutron Resonances of 1 75Lu
Eo, eV
r, meV
2grn, meV
Eo, eV
2grn, meV
? 0,141+0,003
73+4
0,0864+0,0047
55,98+0,18
28,9+3,4
1,568+0,005
61+2
0,480+0,005
4,370+0,010
68+4
0,419+0,022
58,54+0,13
0,49+0,11
6,130+0,014
58+7
1,438+0,160
59,63+0,18
0,27+0,08
8,14+0,02
177+17
0,0287+0,0016
60,68+0,18
6,7+0,8
9,73+0,03
68+2
1,365+0,012
64,02+0,19
3,4+0,5
10,79-T-0,04
73+2
1,697+0,017
64,96+0,19
4,6+0,6
11,44-F0,04
69+8
1,072+0,073
65,84+0,19
0,13-F0,06
11,88+0,05
69+8
0,442+0,015
67,81+0,19
12,0+1,6
16,90+0,07
277+17
0,052+0,007
70,6+0,2
19,1+2,1
19,06+0,07
129+14
0,139+0,006
71,6+0,2
8,3+1,3
19,94-F0,07
86+7
0,409+0,008
75,6+0,2
13,7+1,6
21,76+0,08
75+5
1,364+0,042
78,3-T0,2
8,7+1,5
24,49+0,10
85+7
3,514+0,069
79,7+0,2
1,9+0,5
27,15T-0,11
98+15
11,04+0,16
82,7-T0,2
16,4+1,9
*28,80+0,12
84,2+0,2
0,53+0,25
29,34+0,12
59+32
2,25+1,38
85,4+0,2
39,5+4,0
29,53+0,12
188+47
3,28+1,15
89,4+0,3
13,4+2,5
31,83+0,14
136+15
2,55+0,10
92,4T-0,3
19,2+2,1
32,66T-0,14
46+22
0,35+0,04
95,6-T0,3
5,0+1,3
33,25+0,14
86+10
2,70+0 17
98,9+0,3
10+3
36,97+0,15
5,31+1,13
104,7+0,4 ?11+4
38,30+0,15
1,76-F0,60
105,1T0,4
42,06+0,16
62+12
5,66+0,54
108,6-T0,4
21+7
42,55+0,16
42+22
1,46+0,12
112,6+0,5
7+3
45,13+0,17
117+27
1,49-0,05
122,2+0,5
2,3+1,0
46,22+0,17
90+22
3,66+0,26
124,1+0,5
9,4+4,3
47,80+0,17
0,08+0,04
126,4-T0,5
10,5+4,5
49,02+0,18
11,2+5,0
133,5+0,6
135-F15
50,13+0,18
3,0+1,6
135,4+0,6
59+10
*51,35+0,18
0,15+0,06
52,13+0,18
150+31
4,57+0,25
*Doubtful resonances.
584 -
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Fig. 1. Total neutron cross section for 175Lu:
calculated dependence of at with allowance
for positive resonances; ...) experiment.
14,6
14,2
148
60
50
6'
0
4
att
13,4 5,313
30
8
20
110
Fig. 2.
1022 (a)
14,2 13,8
13,4 E, eV
10
60
50
30
Z 20
10
50 100 150 E, eV 0
Fig. 2 Fig. 3
Transmission curves for 175Lu specimens of different
and 0.608.1021 atoms/cm2 (b) near 14 eV.
50
100
thicknesses
150 E, eV
of 0.14.
Fig. 3. Diagram of energy distribution of number of neutron resonances of 175Lu
(a) and 176 Lu(b).
parameters of these resonances are presented in Table 1 and the theoretical curve correspond-
ing to them is given in Fig. 2b. An identification of the spins of the 175Lu resonances in
the region of 14 eV requires additional study. Analogous measurements at E % 41 eV revealed
two resonances at 40.65 and 41.16 eV, which agrees well with the data of [4].
Resonances with energies of 119.45 eV (2grn = 1.240 meV) and 124.45 eV (2grn = 0.246
meV) are cited in [4]. From the transmission curve obtained for 175Lu in this region in our
work it is hard to draw a conclusion concerning the existence of these levels.
In Fig. 3a we present a diagram of the energy distribution of the number of neutron
levels, from which it follows that an appreciable omission of levels is not observed below
195.8 eV. For this energy region we calculated the following: D = 3.78 ? 0.26 eV; 2griq =
1.18 ? 0.23 meV; So = (1.56 ? 0.34).10-4.
Lutetium 176. The specimen was prepared from Lu203 powder which had the following iso-
topic composition: 1761u 64.3% 175Lu 35.7%. The measurements were made in the region of
energies up to 135 eV for two thicknesses which corresponded to a 176Lu content of 1.09.1021
and 9.25.1021 atoms/cm2. We found 61 resonances below the energy of 135 eV, of which 41 were
detected for the first time. The parameters of the neutron resonances up to an energy of
585
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52 eV were calculated by the shape method and those above 52 eV by the area method (Table 2).
Analogous data are contained in [3, 6].
A diagram of the energy distribution of the number of neutron resonances for 176Lu is
presented in Fig. 3b. It follows from the diagram that the omission of levels begins above
70 eV. Up to an energy of 72 eV we calculated the following: D = 1.74 ? 0.17 eV; 2g4. =
0.58 ? 0.14 meV; So = (1.70 ? 0.43)910-4.
LITERATURE CITED
1. S. M. Kalebin et al., Prib, Tekh. Eksp., 3, 79 (1970).
2. S. M. Kalebin et al., in: Proceedings of Conference on Neutron Physics [in Russian],
Vol. 2, Naukova Dumka, Kiev (1972), p. 267.
3. Brookhaven Nat. Lab.-325, 3rd ed., New York (1973).
4. H. Lion et al., Phys. Rev., 11, 1231 (1975).
5. A. Namenson et al., Nucl. Phys., A237, 45 (1975).
6. Block, Oak Ridge Nat. Lab.-2718 (1959), p. 28.
586
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Declassified and Approved For Release 2013/04/01: CIA-RDP10-02196R000700090006-3
CONFERENCES
ALL-UNION CONFERENCE ON THE PREPARATION OF OPERATING PERSONNEL
FOR ATOMIC POWER PLANTS
L. M. Voronin, G. N. Ushakov,
and V. M. Gordina
High-quality trained personnel who have undergone special theoretical and practical prep-
aration are needed to assure the start-up and the reliable and safe exploitation of newly
introduced atomic power plants (APP) in the USSR and member countries of the Council of Mutual
Economic Aid (CMEA). This was the topic at the All-Union Scientific-Technical Conference
organized by the Atomic Energy Bureau on February 9-11, 1977, at the Novovoronezh APP. Taking
part in it were more than 150 specialists from 38 organizations, including the Novovoronezh,
Beloyarsk, Kolsk, Leningrad, Armyansk, and Kursk APP, the Bilibinsk atomic thermoelectric
powerplant, representatives of the West-Ukraine, Kalinin, and Smolensk APP under construction,
the Kurchatov Institute of Atomic Energy, the Moscow Power Institute, and others. A total of
39 reports were heard and discussed. They included review reports of the Atomic Energy
Bureau and the Kiev Institute of Automation devoted to the preparation of operating personnel
for APP in the USSR and abroad and to the principles of construction of educational?training
centers; reports on the preparation of exploitational and operating personnel in the educa-
tional-training center of the Novovoronezh APP and at the Kolsk APP, and on methodological
problems.
In the reports of the Kiev Institute of Automation the principles required by the developers
were formulated, based on the results of studies of the educational-training center for
operators of power units with a power of 300 MW, including a study of operating activity, the
programming approach to training, the selection of simulators, the algorithmic approach to
training, and the universality of preparation programs.
The creation of training centers equipped with simulators ? imitators of operating APP ?
was recognized as effective. At the Novovoronezh APP there is already an educational center
for the preparation of qualified specialists. During the years of its existence of prepara-
tion of over 1000 foreign specialists, mainly from member countries of the CMEA, has taken
place at it. At the end of 1976 a new building was placed in use, equipped with specialized
auditoriums in which provision is made for the use of the latest technical means of training.
The training branch intended for the training of operators of water-cooled-water moderated
VVER-440 power units closes in 1977.
Great interest was aroused by reports on the principles and results of the development
of psychophysiological criteria for the selection for training by the leading specialists
of nuclear power engineering and on the principal psychophysiological aspects of the profes-
sional preparation of operators, presented by the Institute of Biophysics of the Academy of
Medical Sciences (AMS) of the USSR and by the Institute of Labor Hygiene and Professional
Illness of the AMS of the USSR.
However, as shown by an analysis of the principal problems in the organization of the
preparation of operating personnel for APP, the simulators presently being used in the USSR
and abroad are distinguished by great variety both in the circuit design and mathematical
provision and in their functions and capacities and in the completeness and accuracy of the
duplication of the corresponding stage. This occurs because they are still in the develop-
ment stage and unified principles for their construction and criteria and standards for
equipping them and training on them have not been defined. Insufficient work has been done
on the generalization of experience on operation in emergency situations and during equip-
ment failures with the analysis of the actions of the operating personnel, and no criteria
have been developed for the selection of perSonnel from the point of view of psychological
Translated from Atomnaya tnergiya, Vol. 42, No. 6, p. 510, June, 1977.
This material is protected by copyright registered in the name of Plenum Publishing Corporation, 227 West 17th Street, New York, N. Y.
10011. No part of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic,
mechanical, photocopying, microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is
available from the publisher for $7.50.
587
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Declassified and Approved For Release 2013/04/01: CIA-RDP10-02196R000700090006-3
compatibility, etc.
The conference yielded many interesting results, which permits a considerable degree
of expansion of the research being conducted in this direction. It developed the appropriate
recommendations for organizations occupied with the preparation of personnel for APP.
The reports presented at the conference will be published.
FIFTH INTERNATIONAL CONFERENCE ON THE PEACEFUL USE OF UNDERGROUND NUCLEAR EXPLOSIONS
I. D. Morokhov, M. P. Grechushkina,
and V. N. Rodionov
There were 56 people from 26 countries present at the conference of experts under the
aegis of the Technical Committee on the Technology of Peaceful Nuclear Explosions, which
took place at Vienna (Austria) on November 22-24, 1976. The nine reports presented for
discussion can be divided into the following three groups.
Scientific Reports. F. Prieto (Mexico) noted in his report that the equation of state
for solids at high pressures, when the dependence between the velocities of shock waves and
particles is linear, can be represented in a dimensionless form such that it will also be
valid for certain liquids and, under certain conditions, even for gases.
The report of S. Reagan (USA) was devoted to the results of research on the equation
of state of molybdenum in which a wave of record intensity (20 Mbar) was recorded by direct
measurements. The shock wave, produced by the rapid heating of a plate of 235U owing to
fission during irradiation by neutrons of a nuclear explosion, acted on a specimen of molyb-
denum located next to it. Its velocity of displacement was found during the passage of a
plane wave between two points in the specimen. The velocity of particle motion was determined
from the doppler shift of six neutron resonances in the neutron energy range 200-800 eV.
Experimental-Industrial Peaceful Nuclear Explosions and Projects. Calculations of the
thermal cycle in the burial of highly active wastes in the cavity of an underground nuclear
explosion showed the possibility of accomplishing their storage so that the melting and sub-
sequent solidification of the rock takes place owing to the liberation of heat (L. Schwarz
et al., USA). As a result of the low rate of heat removal the hot center will be retained
for hundreds of years and contact with water will thereby be eliminated. As an example, the
author calculated that the wastes from the operation for 10 years of 109 GW power stations
can be placed in the cavity created by a nuclear explosion with a power of 5 kton which occurs
at a depth of 2000 m in dry silicate rock with a low porosity.
The results of recent studies in the region of the Rio Blanco triple explosion for the
purpose of gas stimulation were presented by L. Ballou (USA). In the first report on this
subject made at the Fourth Conference, it was noted that a zone uniformly permeable to gas
could not be obtained as a result of the three explosions of 30 kton each at depths of 1780,
1900, and 2040 m. To clarify the reasons for the isolation of the fragmentation zones,
additional studies were made both of the initial characteristics of the collector and of the
state of the rock near the deepest explosion. Very interesting data, although incomplete,
were obtained. Nevertheless, the absence of the expected considerable increase in gas output
and the weak permeability between the upper and lower plates, explained by the incorrect
original estimate of the initial state of the collector, made the financing of the experiment
difficult and unfortunately led to the cessation of the studies of this unique explosion.
The report of V. N. Rodionov et al. (USSR) was devoted to a study of the filtration
properties of a rock massif. The permeability of the massif was found through a determina-
tion of the flow rate of air forced under different pressures into specially drilled holes.
588
Translated from Atomnaya Energiya, Vol. 42, No. 6, pp. 510-511, June, 1977.
This material is protected by copyright registered in the name of Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y.
10011. No part of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic,
mechanical, photocopying, microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is
available from the publisher for $7.50.
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Declassified and Approved For Release 2013/04/01: CIA-RDP10-02196R000700090006-3
The results of measurement at different distances from the center of an explosion with a
power of 1 kton snowed that the permeability increases toward the center and hardly changes
over a long time (about a year). In the authors' opinion, this method can be applied to
the study of the degree of destruction of rock near an explosion.
A report on a project for the construction of a canal from the Mediterranean Sea to the
Qattara Depression in the Sahara (E. el-Shasli, Egypt) elicited great interest. The average
depth lies at 70 m below sea level and it is proposed to use the flow of water into it from
the Mediterranean Sea for the operation of electric power stations. After the depression
is filled to a depth of 60 m, which will occur in 10 years according to the calculations,
tile evaporation of water from the water table of the lake which is formed can compensate for
the inflow of water, equal to 650 m9/sec, and thus provide for the prolonged operation of the
hydroelectric stations (HES). The length of the channel over the various planned routes is
70-80 km and the total power of the explosions needed for the construction is estimated at
155 Mton.
It is proposed to reduce the radiation danger to the minimum through the use of charges
of small fission power, by choosing days with a suitable wind direction for the explosions,
and by evacuation of the local population, which comprises less than 25,000 people within
the limits of 80 km from the canal route. After the depression is filled the HES with a
power of 1200 MW will operate for the coverage of peak loads.
Safety. Data on ground motion and damage to buildings during the Rulison explosion
were taken into account in predicting the consequences of the Rio Blanco experiment on gas
stimulation which was conducted in the same sedimentary rock (report of F. Holzer, USA). The
results of these and other explosions made it possible to widen the range of application of
the prediction method which takes into account both the power of the charge and its depth
of placement. It was also noted that the peak accelerations grow with an increase in power
and depth and the characteristics of the spectra for deeper explosions are shifted into fre-
quency intervals which are dangerous to buildings.
In a report on assuring the minimum radioactive contamination of the natural environment
during the conducting of underground nuclear explosions for peaceful purposes (Yu. A. Izrael'
and M. P. Grechushkina, USSR) estimates are given of the composition and distribution of
radioactive products among the different contamination zones during underground and excavat-
ing nuclear explosions. Experimental data are cited which show that a large portion of the
radioactive products formed remains in the central and epicentral zones even in the case of
an excavating nuclear explosion.
The main products which spread to great distances are 8929?Sr, 197Cs, 140Ba, and some
isotopes ? products of the interaction of neutrons with the construction materials of the
cnarge. The authors suggest approaches to the development of criteria for the possibility
of conducting peaceful nuclear explosions under conditions of radiation safety.
In the report of K. Edvarson (Sweden) it is proposed to apply to peaceful nuclear ex-
plosions the general conditions and values of the limiting allowable dose (LAD) from the
Recommendations of the International Commission on Radiation Protection in the same way as
to other radiation sources created by man, such as atomic power plants. In the author's
opinion, along-with the LAD it is necessary to introduce a general means of estimating radia-
tion loads in order to provide an answer to whether nuclear explosions are justified and
advisable from the point of view of future radiation consequences and to compare them with
other forms of human activity leading to exposure of the population. For this it is proposed
to use the concept of the collective dose commitment, which is the sum of the individual doses
produced by events which can be analyzed in practice.
589
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NEW APPARATUS AND INSTRUMENTS
A RADIATION-CHEMICAL INSTALLATION WITH AN EU-0.4 PLECTRON
ACCELERATOR FOR OBTAINING ORGANOSILICON MONOMERS
B. I. Vainshtein, D. M. Margolin,
I. A. Ryakhovskaya, and N. G. Ufimtsev
For the development of the radiation-chemical technology of some organosilicon monomers
which are important for industry and for the conducting of scientific-research work in this
field a factory pilot installation has been created using an EU-0.4 electron accelerator de-
veloped at the L. Ya. Karpov Scientific-Research Institute of Physics and Chemistry on the
basis of an RUP-400-5 x-ray flaw-detecting apparatus.
The installation 1 (Fig. 1) is placed in a specially equipped room consisting of a
chamber 2 and a control room 6. The biological protection 10 is made of brick and the pro-
tective metal rolling door 9 is shielded with lead. The protective observational viewing
window 3 with a size of 500 x 500 mm is covered with leaded glass. The control panel for
the accelerator 5 and for the technological process 4, the auxiliary technological equipment
8, and the physicochemical monitoring instruments 7 were located in the control room. The
installation is equipped with the appropriate kinds of interlocks which are connected sequen-
tially and provide for the operation of the accelerator in the case when the protective door
is closed, for supply water for cooling and air for the membrane of the beam outlet, etc.
The technological diagram of the installation (Fig. 2) consists,of the electron accelera-
tor 4, the reagent supply system 1, an evaporator 2, the radiation-chemical apparatus 3, and
a system 5 for condensation of the reaction products. The E1J-0.4 electron accelerator in-
cludes the high-voltage transformer unit of the RUP-400-5 apparatus, a modified accelerator
tube of type 1.5BPV2-400, a vacuum system with a chamber for the emission of the electron
beam, a system for forming the electron beam, and a mounting system. The latter provides
for the mounting and operation of the installation. The reagent supply system consists of
two vessels and a piston batcher with an adjustable supply rate of 5-30 g/min of the reaction
mixture. Heating of the reagent mixture to 350-400?C is accomplished with the evaporator.
The radiation-chemical apparatus is built in the form of a parallelepiped with a capacity
of 70 liters and with a titanium membrane 20 p thick for the admission of the electron beam.
The condensation system consists of air and water heat exchangers connected in series with
traps for the absorption of the emerging gases.
Fig. 1. Arrangement of radiation-chemical instal-
lation.
590
Translated from Atomnaya tnergiya, Vol. 42, No. .6, pp:512-513, June, 1977.
This material is protected by copyright registered in the name of Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y.
10011. No part of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic,
mechanical, photocopying, microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is
available from the publisher for $7.50.
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Declassified and Approved For Release 2013/04/01: CIA-RDP10-02196R000700090006-3
Fig.
2. Technological diagram of
installation.
Principal characteristics of the installation:
radiation-chemical
Height, m 3
Required power, kW 10-15
Accelerator power supply from alternating current
grid with a frequency of
50 Hz and voltage of 220V
Electron energy, MeV 0.3-0.35
Current in beam, mA up to 2
Dimensions of beam cross section at exit
from accelerator, mm 40 x 500
Total flow rate, liter/min:
of cooling water up to 5
of air 25
Mass, t 1
Principal parameters of the chemical process: temperature of reaction volume up to 350-
400?, atmospheric pressure.
The installation operates as follows. The accelerator is turned on and the assigned
conditions relative to the energy and current of the electron beam are established, the beam
enters the reaction volume through the exit window and the entrance membrane of the radiation-
chemical apparatus located below the accelerator, inducing a chemical chain reaction of con-
densation of the reagents, which are continuously supplied in the vaporized state with the
help of the system 1 and the evaporator 2 (see Fig. 2). Form the system 5 with the accelera-
tor turned off samples of the condensate are selected which are analyzed on the LKhM-72
chromatograph and a laboratory fractionating column.
591
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THE RID-41 UNIVERSAL HOSE-TYPE y-RAY FLAW DETECTOR
V. N. Glebov and V. G. Firstov
The RID-41 y-ray flaw detector was developed and built at the All-Union Scientific-Re-
search Institute of Radiation Technology for monitoring steel products 60-250 mm thick and
products made of light metal and alloys based on them 200-600 mm thick. The instrument
consists of a radiation head, an ampule duct inside which the radiation source which is at-
tached to a toothed flexible cable moves, a receiving hose to hold the unused part of the
toothed cable, a remote control panel, and an exposure meter. The collimating heads and
adapters which fit on the RID-41 form frontal and annular radiation beams, which permits the
effective use of the instrument in different systems of radiographic monitoring and in radio-
metric systems.
The radiation head (Fig. 1) is designed for storing the radiation source in the interval
between exposures and for moving it with a velocity of 0.1 and 1 m/sec along the flexible,
rigid, or combination ampule duct to the collimating head or adapter at a distance of up to
12 m. The head consists of a cylindrical casing within which are located the biological
protection unit, the movement drive, the radiometer pickup for signaling about the radiation
environment, and a mechanism for the emergency return of the source. The control of the
working cycle and of the process of charging (discharging) of the y-ray flaw detector takes
place remotely from a distance of up to 100 m.
The RID-41 y-ray flaw detector is the first domestic instrument of such a class which
permits the complete automation of the radioscopy process using an exposure meter whose
detector is placed at the point where the film is located. The radiation source can be re-
turned to the storage position by hand with the emergency return mechanism in the case of
failure of the electromechanical drive, with overexposure of the personnel being eliminated.
The control panel (Fig. 2) consists of the power supply, control, and detector units
placed in a common housing.
The electrical signaling system indicates the position of the radiation source by means
of signal bulbs: a green color corresponds to the storage position in the radiation head and
a red color to the radioscopy position in the collimating head.
Fig. 1. Radiation head of RID-41 y-ray
flaw detector.
592
Translated from Atomnaya tnergiya, Vol. 42, No. 6, pp. 513-514, June, 1977.
This material is protected by copyright registered in the name of Plenum Publishing Corporation, 227 West 17th Street, New York, N. Y.
10011. No part of this publication may be reproduced; stored in a retrieval system, or transmitted, in any form or by any means, electronic,
mechanical, photocopying, microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is
available from the publisher for $7.50.
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Declassified and Approved For Release 2013/04/01: CIA-RDP10;02196R000700090006-3
Fig. 2. Control panel.
The electromechanical tracking system indicates the location of the radiation source
relative to the storage position at any time with an error of no more than 0.05 in.
The power supply of the apparatus is taken from a three-phase alternating-current line
with a voltage of 380 V and a frequency of 50 Hz.
The dimensions of the radiation head on the carriage are 1400 x 800 x 800 mm, those of
the control pannel are 490 x 400 x 220 mm, and the masses are 1200 and 45 kg, respectively.
A component part of the instrument is the UKT-D25 packing transportation complex for
transporting the recharging container with the radiation source. In assembled form the com-
plex withstands major accidents (fire, a temperature of 800?C, a fall from a height of 9 in
onto a concrete base and from 1 in onto a steel spike). Two "Co sources with a total expos-
ing dose rate of up to 3.6.10-6 A/kg (1.4.10-1 rd/sec) are stored in the recharging container.
The RID-41 y-ray flaw detector is designed for work with one of the following sources
(MRTU 10-62-68): type II: exposure dose rate 3.1.10-6 A/kg (1.2.10-2 rd/sec); type IV: 9.10-6
(3.5.10-2); type V: 3.6.10-6 (1.4.10').
The charging of the source is accomplished with a manipulator in a protective chamber
using a special device which makes it possible to reliably fasten the radiation source in the
holder; the reloading of the holder from the recharging container into the radiation head is
done with the ampule duct.
An experimental model of the y-ray flaw detector is presently being employed successfully.
593
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BOu.N ttE V 1E Wb
A. A. Van'kov, A. I. Voropaev, and L. N. Yurova
THE ANALYSIS OF A REACTOR-PHYSICS EXPERIMENT*
Reviewed by A. D. Zhirnov
The problems of increasing the accuracy of the prediction of the physical characteristics
of reactors occupy an important place in the practical work of a reactor physicist. Not only
the attainment of the design characteristics but also the determination of the role and the
place of one or another reactor in a nuclear power system essentially depend on this.
Two courses have been outlined here: the computational analysis of experimental results
in a relatively simple geometry and model experiments. In the first case the constants are
refined while in the second case the calculating model is tested. However, the tests answer
only a small number of questions, and a considerable portion of the parameters must be deter-
mined (predicted) without relying on direct measurements. How can the nuclear constants be
refined when one has an integral reactor experiment? The authors of the book under review
provide the answer to this question. The statistical method of refining the group constants
used in the calculations of fast reactors are presented in considerable detail in the book
on the basis of their own work and the work of other authors.
Very roughly the essence of the method consists in the following. Suppose that one
knows With a certain accuracy a set of constants (or the region of their determination) and
one knows the reaction of a reactor functional to single perturbations of the group constants,
i.e., the sensitivity coefficients. Having available a set of critical experiments where this
functional is determined with sufficient accuracy one can, having approximated the calculated
value of the functional to the experimental value, by the least-squares method, e.g., correct
the constants or obtain their shift. In this way one is able to make the calculated and ex-
perimental values of Keff converge to 0.3-0.4%. In analyzing in detail the mathematical
apparatus used and citing examples the authors dwell on both the advantages and defects of
the method. For example, in choosing the matrix of errors of the group constants and justify-
ing their approach they note the subjective character of this choice and do not rule other
approaches which might lead to a different set of shifts of the constants. The stability of
the shift in the determination of different functionals can serve as a control here. The
method is well illustrated with graphs and tables.
The prediction of a characteristic of a fast reactor which is important for nuclear
safety, the Doppler and sodium void coefficients, is discussed separately. By refinement
of the constants one is able to reduce the error of the calculated prediction of the void
col.Ifficients for a number of critical experiments and for the startup of the BN-350. Never-
theless, the authors correctly note that the nature of the temperature and power effects is
very complicated and does not yet provide a basis for confident conclusions. The work which
has been done pertains mainly to the "startup states" of reactors. A detailed analysis for
deep fuel depletions, when higher isotopes of plutonium appear with still greater uncertain-
ties in the constants, still awaits its investigators. The method developed is a good help
in this work.
The book contains a large biblography, is written concisely and interestingly, and will
undoubtedly find grateful readers.
*Atomizdat, Moscow (1977), 88 pp., 61 kopecks.
594
Translated from Atomnaya tnergiya, Vol. 42, No. 6, p. 515, June, 1977.
This material is protected by copyright registered in the name of Plenum Publishing Corporation, 227 West 17th Street, New York, N. Y.
10011. No part of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic,
mechanical, photocopying, microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is
available from the publisher for $7.50.
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Declassified and Approved For Release 2013/04/01: CIA-RDP10-02196R000700090006-3
E. A. Stumbur
THE APPLICATION OF PERTURBATION THEORY TO THE PHYSICS OF NUCLEAR REACTORS*
Reviewed by V. N. Artamkin
The book under review is the latest in the series "Physics of Nuclear Reactors" published
by Atomizdat and is the first one in the world literature devoted to perturbation theory for
nuclear reactors. Although various aspects of perturbation theory are described in numerous
periodical publications and textbooks and not one monograph on the general problems of the
theory of nuclear reactors has gotten by without the corresponding (usually summary) section,
until now nobody has taken the trouble to gather together all the available material. At the
same time, the degree of mastery of perturbation theory is now the lodestone by which the
qualification of a reactor physicist is determined. Only after many years of work are the
capability and skill developed for the extraction of the maximum of information from the
minimum of data using sufficiently precise approximate methods, based on perturbation theory
in particular. There is no doubt that E. A. Stumbur's book will promote the accelerated prep-
aration of qualified specialists and the general elevation of reactor culture. It is only
depressing that there are no more than 1600 of such "lucky ones," which, not suprisingly, is
more than the meager press run of the book.
Two of the six sections are essentially devoted to perturbation theory as such: the
general propositions are presented in Sec. 4 while several examples of the use of perturba-
tion theory in practical reactor work are presented in Sec. 6. Whereas the first of these
sections has a general character and touches upon almost all the main aspects of the theory
in one form or another, the choice of illustrations in the latter clearly reflects the author's
taste: they all elucidate problems of the interpretation and analysis of experimental data
obtained on a reactor. The material presented in Sec. 5 ("Void coefficients and the analysis
of their properties") is connected with perturbation theory mostly in a historical way.
A considerable part of the book (40% by size) is by way of an introductory general-
educational section. Evidently, the material presented here will not prove to be new to the
majority of readers. It is hard to imagine, of course, that such a book would interest a
person who did not know how the fields in reactors are described (Sec. 1) or what the im-
portance of neutrons is (Sec. 2), although there might be some difficulties connected with
the use of the formalism of functional analysis (Sec. 3). And yet the material of the book
would not be complete without these three introductory sections. At the same time, it must
be noted that the author was able to demonstrate no little pedagogical mastery, without
which it would be impossible to fit such extensive material (even with a considerable lack
or rigor) into such a limited space.
Twice in the course of the book, in the foreward and in the last phrase, the author
enumerates the problems and tasks which did not find reflection in the book. Unfortunately,
there are quite a few. However, this does not hinder the preparation of the book as a
finished unified whole, keeping in mind the purpose formulated by the author and its size.
All the same, a certain feeling of dissatisfaction remains. First of all, its title gives
reason to assume that its contents are broader, and second, several of those aspects of
perturbation theory and its applications which make it attractive for a reactor physicist
and induce him to adopt the methods of the theory were left outside the limits of the book.
The book under review is distinguished by accuracy and clarity of presentation and by
good language, and the individuality of the author is seen in everything. It is pleasant
*Atomizdat, Moscow (1977), 128 pp, 87 kopecks.
Translated from Atomnaya Lergiya, Vol. 42, No. 6, pp. 515-516, June, 1977.
This material is protected by copyright registered in the name of Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y.
10011. No part of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic,
mechanical, photocopying, microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is
available from the publisher for $7.50.
50.5
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Declassified and Approved For Release 2013/04/01: CIA-RDP10-02196R000700090006-3
to note that from the very beginning the author has rejected the term "neutron flux density,"
which is foreign to reactor physics and which has unfortunately so contaminated the language
of current publications.
L. S. Tong
BOILING CRISIS AND CRITICAL HEAT FLUX*
Reviewed by N. S. Khlopkin
The book, written by a well-known specialist in the field of heat transfer during boil-
ing in nuclear reactors, a scientist of the Westinghouse Co. (USA), contains data on the
effect of various mode parameters and the channel geometry on the boiling crisis, presents
calculating equations for critical heat loads, and gives recommendations on their use.
In the introduction, the flow structure and the mechanism of development of the boiling
crisis at a heated surface in the four principal cases of heat removal are discussed: during
boiling in a large volume; during bubble boiling; in a dispersed-annular mode of motion; at
a transversely bathed cylinder.
In the second section, on the basis of the local theory of the crisis according to which
its development is caused by a certain set of local conditions, calculating equations for
these cases of heat exchange are presented which take into account the effect on the critical
heat loads of the various parameters, the structure and prehistory of the flow, the size and
location of bubbles, the thickness of the liquid film, the instability of the flow, the
channel geometry of the bundle of rods, the spacing grids, and the nonuniformity of the flow
distribution along the length and the circumference. Experimental data on the heat-exchange
crisis of different liquids are generalized on the basis of similarity theory and similarity
criteria are proposed for the crisis of boiling in a large volume and in a stream during
dispersed-annular motion. At the end ?of the section empirical equations for the calculation
of critical heat loads are presented with an indication of the source and the regions of
applicability. Here it is also noted which of the functions are obsolete.
In a separate table data are given on studies which have been made on bundles of rods
with an indication of the nature of the bundle and the range of mode parameters and the
number of points obtained, but without giving the equations for the calculation. The sources
are indicated so that the information can be looked up when necessary.
It must be noted that the calculating equations based on the allowance for local con-
ditions do not always yield satisfactory results. The method which determines the critical
power of the installation and which is based only on the independently controllable para-
meters is sometimes more successful.
A brief theoretical analysis of the boiling crisis is made in the third section.
The heat-exchange crisis during boiling in a large volume is analyzed using the Helm-
holtz instability. The equation obtained turned out to be very close to the semiempirical
dependence of S. S. Kutateladze which was derived using similarity theory. The heat-ex-
change crisis in the bubble mode was calculated with allowance for the balance of heat and
of the mass of coolant and for the hydrodynamic behavior of the main stream under the as-
sumption that the crisis sets in when the limiting superheating of the liquid sublayer is
reached. A calculating equation for estimating the effect of nonuniformity of the specific
*Translated from English [original report available from National Technical Information
Service, U.S. Department of Commerce, Springfield, Va. (1972)1, Atomizdat, Moscow (1976),
100 pp., 54 kopecks.
596
Translated from Atomnaya Energiya, Vol. 42, No. 6, pp. 516-517, June, 1977.
This material is protected by copyright registered in the name of Plenum Publishing Corporation, 227 West 17th Street, New York, N. Y.
10011. No part of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic,
mechanical, photocopying, microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is
available from the publisher for $7.50.
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Declassified and Approved For Release 2013/04/01: CIA-RDP10-02196R000700090006-3
heat flux (the stream form factor) is derived on the basis of these concepts. Adopt-
ing the concept of vapor bubbles forcing the liquid boundary layer away from the wall,
the author has obtained a semiempirical equation for the critical heat flux which does not
differ greatly from the experimental data. The mass exchange of liquid between the liquid
film and the vapor core of the stream is taken into account for the dispersed?annular mode
of stream motion. However, the analytical estimates of the rate of mass removal and precipita-
tion of the liquid have considerable errors. Empirical values are usually used for them.
The value of the theoretical section consists in the attempts to develop a model for
the complex thermophysical and hydrodynamic phenomenon, the crisis, which would make it pos-
sible to generalize the cumbersome experimental material with allowance for the physical
meaning of the coefficients which enter into the equations. As for the accuracy of the
theoretical equations, it is lower than the accuracy of empirical equations.
Recommendations on the selection of the functions presented in the book for calculations
of critical heat loads are given in the final section. There are errors in the writing of the
equations, however. For example, in Eq. (3.25) and (3.17) it should be "local" instead of
,"irregular" in accordance with p. 54; in (3.25) it should be q(h) instead of qh and in the
equation for F on p. 54 it should be q(z) in the numerator instead of qcr(z); in (3.26) "c"
can take on negative values if S is defined as indicated. And this contradicts the physical
meaning.
The book introduced is valuable as a reference text for specialists making calculations
of critical heat loads ? to acquire confidence in the correctness of the equations used; for
experimenters making studies of the heat-exchange crisis ? for the comparison and analysis
of the experimental data obtained; for theoretical heat engineers ? for the development of
models of the crisis. However, a careful analysis of the applicability of the equations
recommended in the book is required for each concrete case of calculation. Where possible
it is preferable to use data of direct measurements on full-scale assemblies of heat-releas-
ing elements. Responsible calculations can only be based on data obtained in such a way.
597
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