SOVIET ATOMIC ENERGY - VOL. 36, NO. 1

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• Document Page Count: 117
• Document Creation Date: December 27, 2016
• Document Release Date: January 25, 2013
• Sequence Number: 1
• Publication Date: July 1, 1974
• Content Type: REPORT

Declassified and Approved For Release 2013/01/25 :CIA-RDP10-021968000400030001-7
Russian Original VoY. 36, No. 1, January, 1974
July, 1974
SATEAZ 36 (1) 1-112 (1974)
SOVIET
ATOMIC
ENERGY
ATOMHAA 3HEP1'VIA
(ATOMNAYA ENERGIYA)
TRANSLATED FROM RUSSIAN
CONSULTANTS BUREAU, NEW YORK
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SOVIET
ATOMIC
ENERGY
~, I_ .
Editor: M. D. Miliionshchikov?
-~ .
~'Depury Dlre~ctor
r
Soviet Atomic Energy is acover-to-cover, translation of Atomnaya
Energiya, a publication of the Academy of`Sciences of the USSR.
.
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original glossy photographs and. artwork. This.serves to decrease
the necessary-time lag, between publication of'ihe 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 ~nergiya:
I. V,;Kurchatov Institute of Atomic Energy
Academy of Sciences of the USSR
` Moscow, USSR ,
? ~
Associate Editors `N. A. Vlasov
A. A. Bochvar
N. A. Dollezhal'
V. S. Fursov
I. N. Golovin
V. F. Kalinin
A. K. Krasin,
. A. I. Leipunskii`
V. V,. Matveev ?
M. G. Meshcheryakov
P: N. Paler
V. B..Shevchenko
V. I. Smirnov
A. P. Vinogradov
A. P. Zefirov
Copyright, ?1974 Plenum Publishing Corporation, 227 West 17th Street, New,York,
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SOVIET ATOMIC ENERGY
A translation of Atomnaya Energiya
July, 1974
Volume 36, Number 1 January, 1974
CONTENTS
Engl./Buss.
ARTICLES
The BFS-1 Complex: A Microtron for Studying Fast-Reactor Neutron Spectra
- A, I. Leipunskii, V, V. Orlov, Yu. A. Kazanskii, V. P. Zinov'ev,
F; I. Ukraintsev, N. A, Klintsov, A. V. Shapar', G. N. Ankin, V. T, Vasin,
V. F. Efimenko, P. S, Klemyshev, V:? L. Parashchuk, V. A, Romanovskii,
V. E, Rydkii, G. V. Sazonenkov, V. I, Sokolov, V. F. Filyaev,
1 3,
and Yu. M. Choropov ............... ......................................
Luminescence of Secondary Uranium Minerals at Low Temperatures
- B. S. Goborets and G. A. Sidorenko. ................................. 5 6
Ion Field-Emission Microscopy of Uranium. Preliminary Results - A. L. Suvorov,
G. M, Kukavadze, T. L. Razinkova, B. V. Sharov, V. A. Fedorchenko,
A. F. Bobkov, and B. Ya. Kuznetsov ................................. ..... 13 14
The Flexural Strengths of Disperse Materials Based on Uranium and Molybdenum
Dioxides between 293 and 1870?K - L. E, Kakushadze and R. B. Kotel'nikov .... 19 19
Radiation-Induced Swelling of OKh16N15M3B Steel - V. N. Bykov, A. G. Vakhtin,
V. D. Dmitriev, L. G. Kostromin, A, Ya. Ladygin, and V. I. Shcherbak........? 24 24
Utilization of Pulsed Sorption Columns for the Decontamination of Liquid Radioactive
Wastes - F. V. Rauzen, E, I, Zakharov, B. E, Ryabchikov, V.D. Konorchenko,
and E. G. Odintsova ........................................................ 28`~ 27
Calculation and Prediction of Radioactive Contamination of the Lower Atmosphere by
Atomic Power Station Stack Discharges - N, E. Artemova ..................... 33 32
REVIEWS -
Mixed-Radiation Dosimetry - B. A. Briskman and R. B. Novgorodtsev ............... 39 39
ABSTRACTS
Application of the Bubnov-Galerkin Method to a Multigroup Calculation of a
Two-Dimensional Reactor - I. P. Kukharenok ............................... 51 51
Measurement of the Absolute Intensity of the,278 keV Line of Npzss
- L. N. Yurova, A. V. Bushuev, V. I. Petrov, A. G. Inikhov,
52 51
V: N. Ozerkov, and V. V. Chachin ..........................................
Parameters of the Radiation.Field near an Apparatus Used for Agricultural
Irradiation - V. P. Bulatov-and E, I. Tsygankov ............................. 53 51
LETTERS TO THE EDITOR
Calculation of Integrated Cross Sections of Compton Interaction, Scattering and
Absorption of y Quanta for Statistical Modeling of Transport Processes
- O. S. Marenkov and V. N. Mitov .......................................... 54 53
Calorimetric Dosimetry and a Procedure for Irradiating Samples, in Electron
Accelerator Investigations of the Radiation Stability of Petroleum Oils
A. D. Stukin and G. I, Shor ..............................................: 56 54
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CONTENTS.
Engl./Ross.
Stages of Formation of Uranium and Rare Metal Mineralization in Sedimentary Rocks
- V. I. Danchev ........ ................................. ....... ....
.
58
55
,Geochemical Isotopic Anomalies and the Hypothesis of Natural Nuclear Reactors
~
- R. S. Prasolov ...............................:...... ....................
61
57
T.he Thermal Conductivity of Uranium Dioxide - V, I, Kolyadin, E, P, Il'in,
A, G, Kharlamov, and V, V.?Yakovlev .........................................
64
59
High-Altitude Measurements of the Natural Radioactivity of the Air - A, E, Shem'i-zade
and R. U. Mambetov ............................ ..........................
.
67
61
Condensation Type Cryostat Channel for Low-Temperature
Exposures
- V, D, Parkhomenko, B, N: Goshchitskii, S, F, Dubinin, P, M, Korotovskikh,
S, K. Sidorov, V, G: Chudinov, and Yu, G, Chukalkin ..................... ....
69
62
Dispersity of Aerosols Formed during Combustion of the Coolant and Materials of the
Core of a Fast Reactor - B, N, Rakhmanov, S, V, Malyutin, O, M,? Zaraev,
and I, E, Konstantinov .:......................................................
73
64
Determination of the Yields of Certain Fragments in the Fission of 238U by Reactor
Spectrum Neutrons - L, N, Yurova, A, V, Bushuev, and A, F, Kozhin...........
75
66
Uranium and Plutonium Losses with Steel in Thermal Decladding of Fuel Elements
- G, P, Novoselov, Yu, D, Dogaev, and S, A, Perevozchikov ...................
79~
69
Thermal Decladding of Oxide Fuel Elements with Steel Separated from Nuclear Fuel by
Filtration - G. P, Novoselov, A, T, Ageenkov, V, F, Savel'ev, and S, E, Bibikov
81
70
Temperature Measurement in High-Flux Reactors Using .Thermocouples
- N, V, Markina and V, A, T sykanov .........................................
84
72
Tlie Calculation of the Space-Energy Distribution of Secondary Annihilation Radiation .
- Sh, S. Nikolaishviliand G, N, Dzhashiashvili,,,,,,,,,,,, ,,,,,,,,,,,,,,,,,,
87
74
Test Facility for Studying Kinetics of Release of Gaseous Radioactive Fission Products
from Irradiated Materials - D, M, Skorov, A, I, Dashkovskii,
A, G. Zaluzhnyi, and O, M, Storozhuk .........................................
89
76
Interaction of Thermal Neutrons with iszn?Eu Nuclei - I, A, Kondurov, A, M, Berestovoi,
A, I, Egorov, E, M, Korotkikh, and Yu, V, Petrov .............................
92
77
Limitations of Effective Accelerating Fields in Ring Accelerators - V, G, Makhan'kov
? and M, G. Meshcheryakov ....................................................
94
78
COMECON NEWS
Vth International Conference on Mossbauer Spectroscopy (Bratislava, September, 1973)
- A, G, Beda and E, P, Stepanov ............................. ..............
97
80
Collaboration Daybook .............................................................
100
81
INF OR MAT ION
The International Symposium on Mathematical Models of Power-Industry Economics
- Yu. I, Koryakin ................................. ........................
101
82
The Tenth International Mineral Processing Congress - M. L, Skrinichenko............
105
84
All-Union Symposium on Radiobiology and Radioecology, Syktyvkar, Stepember, 1973
- R. M, Aleksakhin .........................................................
108
86
Meeting of the International Commission on Radiological Protection - Yu. I, Moskalev , ,
110
87
The Russian press date (podpisano k pechati) of this issue was 12/25/1973.
Publication therefore did not occur prior to this date, but' must be assumed
to have taken place reasonably soon thereafter.
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THE BFS-1 COMPLEX: A MICROTRON FOR STUDYING
FAST-REACTOR NEUTRON SPECTRA
A. I. Leipunskii,* V. V. Orlov,
Yu. A. Kazanskii, V. P. Zinov'ev,
F, i. Ukraintsev, N. A. Klintsov,
A.
V.
P.
V.
G.
V.
Shapar', G. N. Ankin,
Vasin, V. F. Efimenko,
Klemyshev, V. L. Parashchuk,
Romanovskii, V. E. Rydkii,
Sazonenkov, V. I. Sokolov,
F ilyaev, and Yu. M. Choropov
V.
T,
S.
A.
V.
F.
Highly accurate knowledge of the basic characteristics of reactors is required in the design of indus-
trial fast power reactors. That is the reason why comprehensive investigations are carried out on fast
zero-power critical assemblies, with the purpose of improving the accuracy of power reactor character-
istics calculations. An important place is reserved for the study of neutron spectra in this research. Here
the reason is that the number of reactions taking place in a reactor (fission, absorption, scattering) is de-
termined by the interaction between neutrons and nuclei over a broad range of energies (from 100 eV up to
10 MeV), where the interaction cross sections undergo significant changes. Consequently, exact deter-
minations of the number of processes taking place in a reactor are required in order to obtain precise in-
formation on neutron spectra. An experimental study of neutron spectra is also needed in order to verify
and correct many-group systems of constants, and in order to effect improvements in computational pro-
cedures. Various topics associated with the study of neutron spectra in fast reactors were taken up in dis-
cussions at the International Conference on Fast Reactor Spectrometry held at the USA Argonne National
Laboratory [1]. The time-of-flight method offers the greatest range, of all the methods available for mea-
suring neutron spectra in fast critical assemblies. The lower energy limit in the time-of-flight method is de-
termined by the ratio of the number of neutrons in the spectrum at that energy to the background of delayed
neutrons, and falls within a range from 10 to 100. eV, depending on the hardness of the spectrum. The up-
per limit is set by the path length, 'and accordingly by the power of the pulsed neutron source.
The time-of-flight method makes it possible to measure the spectra of leakage neutrons from a sub-
critical reactor (usually at keff - 0.9 to 0.97). The neutron spectra within a critical reactor can be ob-
tained from the measured spectra, if corrections owing to the absence of spectra of scalar and vector neu-
tron flux, perturbation of the spectrum by the exit channel, and the difference in the neutron spectra in a
critical reactor and in a subcritical reactor, are introduced. These corrections have been studied by sev-.
eral authors [2-4], and they have been established as small ones. For example, according to data reported
in [l, 3, 4J, the difference in the spectra at the center of the core in a critical reactor and in a subcritical
reactor (keff - 0.85 to 0.87) does not exceed 10~ over the energy range from 100 eV to 10 MeV. Perturba-
tion of the spectrum by a 90 mm diameter exit channel amounts to no more than 3`~6, according to data re-
ported in [2]. At energies below 1 MeV, the difference in the spectra of the scalar neutron flux and vector
neutron flux at the center of the reactor, where the neutron flux gradient is small, amounts to several
percent [2J. Consequently, computational corrections, amounting to about 15`,x, have to be introduced into
the measured spectra of neutron leakage from the center of the reactor core.
Translated from Atomnaya Energiya, Vol. 36, No. 1, pp. 3-5, January, 1974. Original article sub-
mitted February 1, 1973.
? 1974 Coasult?nts Bureau, a division o f Plenum Publisleiag Corporation, 227 U'est 17th Street, New York, !1~. Y. 10011.
No part o/ this ~pu.blicatioa m?y be reproduced, stored in ? retrieval system, or tr?nsmitted, in nay form or by any means,
electronic, mechaai--cal; photocopying, microfilming, recording or'otherwise, without written permission of the publisher. A
copy o f this article is ?vailable /rom tlce publisher for $15.00.
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Fig. 1. Layout of the BFS-1 microtron facility. 1) detector in
shielding;- 2) collimators; 3) gates; 4) BF5-1; 5) core; 6) target; 7)
electron guide; 8) microtron.
At the Power Physics Institute (FEI) measurements of neutron spectra in fast assemblies by the
time-of-flight method are carried out on the BFS-1 microtron facility,
The BFS=1 testing facility makes iE possible to simulate fast assemblies with core dimensions to
1.6 m, The design features a 2 m diameter steel tank standing 2 m high, and surrounded by shielding.
Matrix tubes 50 mm in diameter, 1 mm wall thickness, into which reactor materials are loaded in the form
of discs 46.7 mm in diameter and either 10 mm or 5 mm thick, are set upright in the steel tank, on a sup-
port plate with a 51 mm pitch hexagonal grid. Details on the design of the test facility are given elsewhere
[5].
The pulsed mode of reactor operation is brought about by inserting a subcritical assembly, (keff - 0.9
to 0.97) of a pulsed neutron source into the core, A uranium target or lead target of .the type used in the
electron accelerator. of a microtron facility is used as such a pulsed neutron source, and is placed at the
boundary separating the core and the reflector. The target is a cylinder 40 mm in diameter and 60 mm
high; it is positioned in one of the reactor channels and is cooled by compressed air,
The microtron employed is described elsewhere [6]. The microtron is operated in the following
mode when the reactor spectra are measured by the time-of-flight method: the energy of the accelerated
electrons is 29 MeV; pulse width is 2 ?sec; current pulse is 10 mA; frequency 50 Hz; mean neutron yield
from target 1011 neutrons/sec,
The neutron flux is extracted from the center of the assembly through an exit channel of cross sec-
tion 100 x 100 mm, The extracted neutrons are directed to an evacuated neutron guide set in the earth; this
guide is a 500 mm diameter steel tube, The guide diameter is increased to 800 mm at a distance of 170 m,
and to 1000 mm at a distance of 550 m, The total length of the neutron guide extends 750 m; measuring
chambers are placed along the length of the guide at distances of 53 m, 230 m, and 760 m from the center
of the reactor.
-The neutron beam extracted from the reactor is .shaped by two collimators. The first collimator is
located within the neutron guide at a distance of 10 m from the reactor center, the second in the first mea-
suring chamber at a distance of 53 m from the reactor,
The dimensions of the collimating holes are selected such that neutrons cannot gain access to the de-
tector from the walls of the exit channel, and the same applies to neutrons scattered singly on? the walls of
the neutron guide, When measurements are taken on a 230 m path length, the diameters of the holes are
selected at 64 mm and 86 mm, respectively for the first and second collimators. There are also several
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x~
~ a f0-z
.~
t ~ ~~= ~.'
?ti~~
'~
~~:
ii~
103
Energy, eV
Fig. 2. Spectrum of leakage neutrons from
center of BFS-27 critical assembly.
baffles placed in the neutron guide to cut down on the num-
ber of neutrons getting through to the detector. after being
scattered. Figure 1 shows the basic components of the
BFS-1 microtron test facility and the time-of-flight spec-
trometer,
A high-efficiency neutron detector consisting of a
polyethylene block accommodating 120 type SNM-43 helium
gas counters was developed for the express purpose of aid-
ing measurements of neutron spectra by the time-of-flight
method; each counter is enveloped by a cadmium jacket,
The parameters of the detector are: neutron recording
efficiency 0.1 for E = 30 eV and 0.063 for E = 200 keV (it
is a smooth function of the energy) *; the time indeter-
minacy is 2-3 ?sec; the area of exposed surface is 2500 cm2; background proper is 1 cps. The detector
is virtually insensitive to v-radiation. Time analysis is carried out in the course of the measurements us-
ing the equipment of the FEI reactor measurements data center, and the experimental data are processed on
the Nairi-2 computer.
At the present time, measurements of neutron spectra are being taken for neutrons from BFS assem-
blies over the energy range from 30 eV to 200 keV, on a path length extending 230 m. The soft part of the
spectrum of leakage neutrons from the BFS-27 assembly is shown in Fig. 2 as an illustrative example.
The measurements were taken at reactor breeding ratio k = 0.9. The energy resolution BE/E amounted
to 60~ for neutrons of 100 keV energy and 6~ for 1 keV neutrons. The count rate for the 5 ? sec channel,
at microtron current 7 mA, was 6.103 cph at the peak of the spectrum and 1.5 ? lOZ cph at energy 1 keV.
The constant background due to delayed neutrons and to the intrinsic background of the detector was 4 cph
for the 5 ?sec channel, Background was measured immediately prior to the next burst from the reactor
on a time window 1280 ?sec in width. The variable (correlated) component of background was measured
by the resonance-filter method. The resonances used were those of cobalt (132 eV and 5.015 keV), r_ian-
ganese (337 eV and 2.375 keV), and sodium (2.85 keV). The relative background levels are listed below:
Neutron energy, eV
132
337
2850
5015
Total background, `,~
35
13
20
7
Variable component of
background, ,~
1
1.5
15
5
Corrections were. introduced when the results of the measurements were processed, the most essen-
tial corrections being for resolution .and for attenuation of the beam of neutrons in dead pockets of the neu-
tron guide and in air. For the example cited (see Fig. 2), the first correction was 30~ at neutron energy
170 keV and less than 5~ at neutron energies below 1 keV, This correction was greatly affected by the du-
ration of the- reactor pulse, and decreased as the pulse width was narrowed. The size of `the second cor-
rection is due to the presence of a relatively large amount of air (15 m) and aluminum (8 mm) on the path
traversed by the neutron beam, The error in the measured spectrum of leakage from the subcritical reac-
tor is?15~over the energy range from 500 eV to 10 keV, 20~ below 500 eV, and 30~ at energies above
50 keV.
Note that the time-of-flight method does not enable us to obtain all of the necessary information on
the neutron spectrum of a fast reactor. A hydrogen proportional counter is being used at the BFS-1 test
stand in studies of the neutron spectrum over the energy range from 10 keV to 2 MeV, and a scintillation
spectrometer with a stilbene crystal is being used for studies in the range of energies upwards of 0.8 MeV.
In conclusion, the authors avail themselves of this opportunity to express their heartfelt thanks to
I. G. Morozov for his kind and persistent attention to the progress of the work, to Yu. Ya, Stavisskii and
A. I. Abramov for their helpful discussions when the facility was in its design stage, and also to all the
staff members of FEI who took part in building the facility.
*In the energy range from 30 eV to 50 keV, the relative variation in effectiveness, measured experimen-
tally, is known to within 10~.
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1.. Fast Reactor Spectrum Measurements and Their Interpretation, IAEA, Vienna (1971), p. 138.
2. J. Sanders, in: Fast Reactor Spectrum Measurements and Their Interpretation, IAEA, Vienna (1971),
p. 36. _
3. T. Oei, RCN-122 (1970).
4. M: Coates, .et al. , J. Nucl. Energy, 22, 547 (1968). _
5. V. V. Bofldarenko, ~et at. ? At. Energ. 24, 82 (1968).
6. A. I. Abramov, et al, , The 30 MeV Microtron at FEI [in Russian], Preprint FEI-211 (1970).
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B. S. Gorobets and G. A. Sidorenko UDC 535.372:549.755.35
Numerous secondary uranium minerals, in which uranium is present in the form of uranyl UOz+,
must be isolated and recognized in the search for uranium ore. All uranyl minerals are very important
indicators of uranium mineralization; some of the uranyl minerals, such as vanadates, uranite, hydrox-
ides, and schroekingerite, are of interest as independent raw material sources, when these minerals are
present in considerable quantities.
Mineralogists have for a long time. used uranyl minerals ability to emit ultraviolet luminescence ra-
diation [1-3]. But contrary to the synthetic uranyl compounds [4], the photoluminescence spectra of the
uranyl minerals were determined only for a few objects with a bright radiation at room temperature [2, 5].
The majority of secondary uranium minerals exhibit a bright luminescence radiation only at sufficiently
low temperatures [6]. The goal of the present work was the determination of the luminescence spectra of
the rather complete group of uranyl minerals at 77?K and 298?K and the interpretation of the spectra in
terms of physics; detailed investigations of the compositions of the minerals on the basis of the resulting
data; and the development of the luminescence technique as a reliable and rapid method for recognizing
uranyl minerals present in traces.
The luminescence spectra were recorded with an ISP-51 spectrograph which was equipped with a pho-
toelectric FEP-1 attachment, an FEU-38 photomultiplier, and an EPP-09mZ recording potentiometer. A
test tube, which 'was made of quartz glass and contained a mineral sample of 5 mg or more, was placed
into a transparent Dewar vessel containing liquid nitrogen. Luminescence of the mineral was excited with
the light of an SVD-120A mercury-filled quartz lamp. A UFS-2 filter and a layer of 10`~ CuSOq solution
were used to single out the excitation interval ranging from 40,000 cxri i to 25,000 cm-1. The accuracy of
'the frequency measurements and the luminescence spectrum amounted to 10-20 cm-1 in the interval 20,000-
16,000 cm-1. This accuracy suffices in work on minerals, because various samples of a particular mineral
are characterized by a shift of identical lines in the spectrum, with the shift frequently exceeding the above-
indicated accuracy limits. The shift results from impurities, various degrees of order of the structure,
and deviations from stoichiometry in the samples considered. Table 1 lists the frequencies of the spectral
lines; the frequencies were averaged over all samples of a particular mineral. The objects of our inves-
tigations were initially identified on the basis of the Debye diagrams.
The minerals examined were separated into two isostructural groups with basically different lumines-
cence spectra. The first group comprises the uranites: autunite (15),* torbernite, and mixed copper-cal-
cium uranites (4), natroautunite (1), uranospinite (2), novaLekite (2), uranocircite (4), and the following
minerals which had been synthetized by I. G. Zhil'tsova from aqueous solutions: Ca-autunite; H-autunite;
(Ca,?Ba, H)-autunites. The second group comprises phosphuranylite (6), Sr-phosphuranylite (1), and re-
nardite (2).
*The number of samples is indicated in parentheses. All minerals of the first group contained 6-8 mole-
cules HZO (meta-form), but, for the sake of simplicity, we use their names without the prefix "meta."
Translated from Atomnaya Energiya, Vol. 36, No. 1, pp. 6-13, January, 1974. Original article '
submitted April 19, 1973.
?1974 Consultants Bureau, a division o f Plenum Publishi.ag Corporatt:on, 227 IT'est 17th Street, New York, N. Y. 10011.
No part o/ this publication may be reproduced, stored in u retrieval system, or transmitted, i.n any form or by an}' means,
electronic, mechanical, photocopying, microfilming, recording or otlaerwi~se, without written permission of the publisher. ,~
copy o/ this article is. available from the ptt.blislter for .$15:00.
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TABLE 1. Spectroscopical Characteristics of the Luminescence of Secondary Uranium
Minerals at 77 ?K
Note, 1. The frequencies of the I series of uranocircite and uranospinite 1') are by about 100 cm"i smaller than
the indicated frequencies, 2, The frequencies of the antisymmetric oscillations are indicated in brackets,
3, Smeared "lines",which were approximately determined, are indicated in parentheses, ~ The- frequencies of
the fully symmetric oscillations were. singled out in groups consisting of several lines,
The minerals of the first group are characterized by a bright yellow-green luminescence at 298 ?K;
the spectrum consists of a series (I) of equidistant lines {Fig, la), The spectra of all the minerals are
almost identical, but in the case of natroautunite, the lines are slightly shifted towards higher frequencies
(ka ~ 19,980 cm-i), whereas the shift goes toward lower frequencies in the case of uranospinite and urano- ,
circite (ko ~ 19,830 cm-i) when compared with autunite (ka ~ .19,920 em-f). The shifts are an important
indicator for the recognition of these minerals: which are hard to analyze, Three types of spectra are ob-
served at 77?I{, In the spectrum of -type I, which is characteristic of ali synthetic samples (Fig, la), only
the series I of equidistant lines is observed, which is also found at 298?x. in the spectrum of type II (ma-
jority of natural samples), two series (II and II) of equidistant lines are present (Fig. lb). In the spectrum
of type III (Fig, lc), a single, new series (III) of equidistant bands is observed; series I usualiy fails to
appear at low temperatures,
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18000 19000
Fig, 1
H
i
a.
K2
Ko
i ~.
\
Ki0
/'~ ~
b
~._.~
~~~ ~ ~
\~~?-
/f
~~
C
~ ~ (H, Na, Ca)~-~
'~
~
.~
.
~.~-- ~ ~
~. ~-
d
~~ ~~,
i
~ i
~
~..
- ~
~ i-~ i l ~~
~-
e
18000 X9000 20000 K, cm' 1
Fig, 2
Fig, 1. Luminescence spectra of uranyl phosphates and uranyl arsenates at 77?K (solid lines)
and `L98?K (dash=dot lines): a-c) meta-autunite group, spectra of types I, II, and III; 1) initial
spectrum; 2) after 30 min heating at 150?C; 3) after storing the heated mineral for several days
in air; ratio of the intensit scales: Imax Imax ~ 3; I I I 100:1 :20; . d hos huran lite
Y 77 ~zsa i~z~a~ )P p Y
-renardite group, Imax : Izm$ x ~ 100.
Fig, 2. ,Luminescence spectra of uranyl silicates at 77?K (solid line) and 298?K (dash-dot line):
a) uranophane; b) ~3-uranotil; c) K and (K, Na, Ca)-boltwoodites; d) Na-boltwoodites (the dashed
curve refers to a mixture with an unknown phase); e) soddyite (Republic of Zaire); Imax ; Izss x
10-100 for the various samples,
The interpretation of the luminescence spectra of the minerals is based upon both experimental and
theoretical data which were previously obtained for synthetic uranyl compounds. (4]. The line with the high-
est frequency ko.corresponds to a pure radiative electron transition from the lowest excited uranyl state
?~u?~zg'~u~g~u into its ground state o2u6gauag. In accordance with thefranck-Condon principle,-the relaxation
of the nuclei begins after the transition, The oxygen nuclei tend to assume a new equilibrium position in
the linear O-U-O molecule, Three types of oscillations of the oxygen nuclei relative to the uranium nu-
cleus can be excited: fully symmetric oscillations, antisymmetric oscillations, and deformation oscillations
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 Declassified and Approved For Release 2013/01/25 :CIA-RDP10-021968000400030001-7
TABLE 2. Crystallochemical Formulas of [4]. Fully symmetric oscillations are most closely as-
the Meta-Autunites Examined (in the order sociated with electron transitions. The corresponding
of increasing structural imperfection); Re- quantum is characteristic of the spectrum of type I
suits of Chemical Analyses (Fig. la) and is Mks = 825 cm-I in autunite; up to 6 of
Type
of the
Formula
tadia-
[temark
tion
center
a*. 77
Cap,oo[(UOz)z,oo(P0a)z,ool?G,1Hz0
I
Synthetic,
spectrum
on Fig.
la
Caos7[(UOz)z,os(P04),scl?G,1Hz0
I, II
Natural,
spectrum
on Fig.
lb
these quanta, which correspond to equidistant lines kn
= ko-nOks (where n = 1, 2, ...) can be emitted (see
Table 1). In the antisymmetric oscillations, the poten-
tial energy excess which appears after an electron
transition does not decrease in a first approximation
and, hence, these oscillations should not be excited at
all. Nevertheless, the emission of one quantum of an-
tisymmetric oscillations is observed in structures in
which O-U-O is structurally unbalanced (4]. For ex-
ample, in autunite in which the uranyl imbalance amounts
Cnoso[(UOz)z,iz(I'Oa)z,ool?7,4H2O I, II The same to 0.2 A, one observes at 77?K, in addition to the kn
Caoso[(UO~)Z,~a(POy)~,ooJ?~,sH2o I, II lines of series I, narrow lines of an antisymmetric ura-
Cao,73Sro,~s[(UOz)z,,s(l Or,)z,ool?8,5Hz0 lII Natural,
spectrum nyl oscillation beginning with n = 1: kna) ~k
_ ~-I- a'
on Fig, where aka ~ 900 cm-I (see Table 1 and Figs, la and b).
lc
In order to study the contribution of water to the
formation of I, II, and III centers which correspond to
the radiation series I, II, and III, the minerals were dehydrated by heating them for 30 min in air at some
temperature between 50 and 250?C. Heating at 50-100?C irreversibly destroyed almost all II centers (see
Fig, lb), The stability of the I and III centers is higher, and their radiation disappears completely only
after heating to 250?C. The process is partially reversible: after keeping the samples for several days in
air, the lines of the I and III series were restored to about" 20`,~ of the initial intensity (see Figs, la-c,
dashed lines).
In order to determine the nature of the I, II, and III centers, we considered the structure of meta-
autunite. It is generally accepted [7, 8] that meta-autunite consists of {[UOz]z+[PO4]3-}n- layers the charge
of which is compensated for by interlayer Cat+, Na+, and other cations, Caz+ ions occupy half of the cat-
ion positions, which results in various degrees of order in the redistribution of Caz+ and the coordinated
water molecules; Na+ ions occupy all cation positions and, hence, the structure of natroautunite has in-
creased order and stability, Four water molecules enter into the first coordination sphere of the cation
(water I) and the other water molecules, usually two (water H), appear in the second coordination sphere,
The excess over thes.e..six molecules can be present in the form of a weakly bound interlayer water or in
the form of oxonium (H3O+) which compensates for deficiencies in the Caz+ and Na+ cations.
Substantial deviations from stoichiometry (Table 2), with which various degrees of order of the in-
terlayer cations and; hence, of the water are correlated, are observed in natural autunites: I, II, and III
centers are formed as a consequence of the nonunifor~in binding of the uranyl to the various forms of water.
The form of the binding of the water in the autunites as studied on the same samples with the technique of
weight-loss (DTG) curves, Three steps of the weight loss [2 + 4 + (0-2)]H2O are characteristic of syn-
thetic autunite, which is distinguished by complete st ichiometry and a completely ordered structure from
natural autunite (see Table 2), Two molecules of II IAter, which is weakly bound to the cation, are lost at
temperatures of about 50?C, The removal of this water does not influence the luminescence spectrum. The
second step corresponds to the loss of I water (4HZ0) din the range 50-250?C, whereupon autunite transforms
into the form Ca(UOz)z(PO4)z(0-2)HzO and no longer emits luminescence radiation (the subsequent step of
losing (0-2)H2O at 400-500?C is therefore not associated with a transformation of radiation centers). Thus,
4HzO, along with UOz, are constituents of I centers, i, e. , these centers are complexes with water, These
centers are the main luminescence centers in phosphates and arsenates of the uranite-structure group.
The lowest thermal stability (up to 50-100?C), irreversible destruction of the centers, and lumines-
cence at low temperatures only are ,characteristic of the II centers. II centers do not appear in synthetic
autunites in which stoichiometry is rigorously .observed. All these properties imply that, obviously, the
H3O+ oxonium of low 'thermal stability (2HZ0 = H3O+ + OH-) and uranyl are constituents of the II centers,
The attenuation of II centers at 298?K seems to be related to the. simultaneous presence of oxonium and OH-,
which dissipates the energy of the optical excitation of uranyl [9].
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16000 19000 20000 21000 K,cm-1
p 3t--__--
~"~
4 -
2 ~
0-i-- 1-
H3
~~_
k, a
-^_ KO
- XZ I
~ ' 1
L- ~ ` 1 \ Kfl~ ,1
/1 i
~.
h1 b
.
HZ
ND
l ~
KII
~ ~
I
K3 ~ '~
/ \ / ~ ~ . `
%~ ~~ ~ ~ 1 KIO .
i.-. ~
~.
H'
C
XO
~7,
n2
I
K10
/~ '~
/ \
,/~
HJ
/~~ ~.
/~ ~\
~. ~ \
i '~.
d
Hu
H~
Hp
~.~
i'~ ~ ~
/ /. '~.\.
H3 ' ~
~~ ~\_
17000 16000 19000 20000 K,cm-1
Fig. 3. Luminescence spectra of carbonates, sulfate-car-
bonates, and sulfates at 77?K (solid line) .and 298?K (dash
-dot line): a) schroekingerite (the electron-oscillation terms
are indicated in the insert); b) uranotallite, Imax : I sz eax ~ 2;
c) meta-uranopilite, Imax : I ss x a 3; d) zippeite, Imax ; I2sa x
30.
In autunites, which are characterized by a III spectrum, the water is also lost in the temperature
range 50-250?C, but, in distinction to the autunites with I and II spectra, the DTG curves are smooth, be-
cause the gradual loss of '~unordered~T water is superimposed upon the steps corresponding to the loss of
the fixed I and II water (2 + 4H2O). The III centers are, in essence, groups of centers which can be formed
by uranyl when order-destroying water molecules accumulate in the neighborhood of an uranyl molecule.
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17000 18000 19000 K , cm-1
Fig. 4. Luminescence spectra at 77?K
of: a, b) vanadates: carnotite and cal-
ciocarnotite; c) uranyl molybdates; d)
uranium hydroxides.
It seems that small regions of I and II autunites are present
in the III autunites, because the luminescence of the "ordered"
I centers is observed at 298?K; the I centers so far appeared
attenuated as "unordered" III centers. The III centers begin
to emit luminescence radiation at low temperatures, when the
narrow tines of the I centers usually cannot be recognized on
the background of the smeared lines of the fully symmetric
oscillations of the III centers (line width 500-600 cm-i).
The influence of CaZ+, Bat+, and MgZ+ cations upon the
uranyl radiation is necessarily smaller than the influence of
the various types of bound water. The luminescence spec-
tra of autunite, uranocircite, and the other minerals are
therefore hardly distinguishable. However, the influence of
the CuZ+ cation upon the uranyl is clearly visible in torber-
nite: torbernite does not emit luminescence radiation. It is
well known that CuZ+ is an active attenuator of the lumines-
cence and, at the same time, a highly polarizing agent. The
consequence is a substantial overlap of the wave functions of
CuZ+ and uranyl and the subsequent quenching of the (exter-
nal) luminescence. But even when only a few percent of Ca
(in mixed copper-calcium uranites) are present, lumines-
cence is excited and spectra of types I, II, or III are observed
in the various samples. This luminescence must be most
likely related to layers of Ca-autunite in the bulk of torber-
nite, because an isomorphous CaZ? CuZ+ replacement is un-
likely.
A creak, green radiation at 298?K and a bright yellow-
green radiation at 77?K (Fig, ld) are characteristic of the
minerals of the isostructural group of phosphuranylite-re-
nardite. The cations (Ca, Sr, Pb) have no influence upon
Both the low intensity and the diffuse band structure at 298?K are related to the
reduced number of water molecules in the coordination sphere of uranyl (e, g. , reduced relative to the I
centers in autunite), and to the presence of OH` ions. Similar reasons for the attenuation of the uranyl
luminescence in aqueous solutions at 298?K were described in [9], where electrolytes stimulating the dis-
sociation of HZO molecules into H+ and OH-were introduced into the solutions. However, uranyl lumines-
cence is observed at lower temperatures, because, due to the enhanced association of the water in the co-
ordination sphere of the uranyl, the interaction between uranyl and the quenching OH- `ions is reduced. The
same arguments can be used for an (at least qualitative) explanation of the partial or full attenuation of the
luminescence of silicates, some sulfates, uranium molybdates, and uranium hydroxides at 298 ?K.
We studied uranophane (9); (3-uranotil (8), boltwoodite (5), soddyite (2), and some of their synthetic
analogs. All these minerals (except for soddyite) are characterized by a weak green radiation at 298?K
and a bright yellow radiation at 77?K. The luminescence spectra of these minerals are a reliable charac-
teristic for their recognition at 77?K (Fig. 2). Uranophane and ~3-uranotil, which have identical composi-
tion but different structures, are particularly distinguished (see Figs. 2a, b). An important detail is ob-
served in the spectra of several Na boltwoodites (Fig. 2d, dashed line) and of synthetic NH4 boltwoodites
(see Table 1) in which two series (I and II) of fully symmetric oscillations have been detected. Series I is
homologous and the relative shift of the lines of series I is related to different influences of Na+ and NH4
cations upon uranyl. The II series coincide in Na and NH4 boltwoodites. Series II can be explained by the
presence of a finely dispersed admixture of a certain "cation-less" uranium phase, the low concentration
of which in these minerals has so far prevented the identification of these minerals from x-ray diffraction
patterns. The decrease in Mks to 740-770 cm-1 indicates that the uranyl ion is extended by about 5?,b in the
structure of the silicates.
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 Declassified and Approved For Release 2013/01/25 :CIA-RDP10-021968000400030001-7
Carbonates, Sulfate -Carbonates, and Sulfates
The most frequently found minerals of these classes were studied; uranotallite (2), bayleyite (1),
schroekingerite (7), meta-uranopilite (2), and zippeite (4). No differences were observed in the lumines-
cence spectra of isostructural carbonates, i? e? , in the luminescence spectra of uranotallite and bayleyite
(Fig. 3b). The spectrum of schroekingerite is distinguished from the two spectra at 77?K (Fig. 3a). The
high brightness of the bluish-green radiation within a broad temperature interval can be explained by the
high concentration of water of crystallization and by the absence of OH- in these minerals, The value ~.ks
830-840 cm-1, which is the usual value for the known uranyl salts [4], is characteristic of UOZ+. The
emission lines are narrow, which means that the structure has a high order and the minerals, are homo-
geneous, Numerous lines of O-U-O deformation oscillations are observed, and other lines, which are
apparently related to oscillations of lattice ions associated with uranyl, appear in the spectra. Antisym-
metric uranyl oscillations were not established in carbonates and sulfates, because the O-U-O configura-
tion is balanced. Anti-Stokes transitions (klo and kli on Fig? 3) are observed at 298?x. These transitions
are photothermal transitions: the energy of the optical excitation of an electron in UOz+ appears added to
the energy of the thermal lattice vibrations so that a stationary population of the sublevel n* = 1 of the ex-
cited UOZ state is formed during a lifetime of about 10-q sec. In this case, a fluorescence transition can
takeplacefrom the two sublevels n* = 0 or n* = 1 of the excited state to the sublevels n = 0 or n = 1 of
the ground state. In this case, the following emission frequencies are observed; klo = ko + pks and kll
= kl + pks (the meaning of the notation can be inferred from the insert of Fig, 3a). The anti-Stokes lines
make it possible to determine the wavelengths of the fully symmetric O-U-O oscillation in the excited
state; OkS ~ 650 f 30 cm-1 in the case of uranotallite and 700 f 30 cm-1 in the case of schroekingerite;
it is well known [4] that Aks is always smaller than L~ks, because the force constant of the bond in the
O-U-O molecule is in the excited state smaller than in the ground state.
The sulfates uranopilite (5H2O-meta-form) and zippeite are characterized by a moderate green lu-
minescence, The corresponding spectra are shown in Figs. 3c and d. The radiation at 77?K is bright and
yellow-green; broad bands of fully symmetric oscillations appear in the spectra, k10 and kil anti-Stokes
transitions are observed in uranopilite at 298?K as well as at 77?K? The Mks value is about 600 cm-1, where-
as Mks ~ 800 cm-1. The partial attenuation of the luminescence at 298?K proves that OH- ions are present
in zippeite and uranopilite.
Vanadates, Molybdates, and Hydroxides
Common features of the minerals of these classes are the absence of radiation at 2~8?K, a bright
yellow luminescence at 77?K, smeared TMlines" in the spectra, and low Oks values. The vanadates were
represented by carnotite (3), calciocarnotite (3), and francevillite (1). The luminescence spectra of carno-
tite and calciocarnotite are complicated, There appear at least two uranyl centers of different nature
(Figs. 4a, b). So far it has been difficult to interpret these centers. The radiation spectrum of francevil-
lite consists of a diffuse band with Amax ~ 18,300 cm i.
Almost identical are the luminescence spectra of the molybdates which were represented by umohoite
(5) UO2(MoO4(H2O2)) ~ 2HZ0; (C a, "1Va)-uranomolybdates (8) (Ca, Na) ? (UOZ)3(MoO4)3(OH)3. 8H2O; and iriginite
(1) UOZ(Mo2O7(H2O)2) ? H2O (Fig? 4c). Obviously, a uranyl center of a certain type causes the luminescence
in these minerals. The structural similarity between (Ca, Na) uranomolybdates and iriginite is beyond
any doubt, These molybdates are typical hydrated Us+ molybdates with a rather incomplete structure.
The additional Ca2+ and Na+ cations are rather unimportant, Umohoite has a particular structure, but is
related to both (Ca, Na) molybdates and iriginite by a layer structure, weak binding between the layers,
and interlayer water, Umohoite is a mineral with variable U4+/Us+ ratios; the color of the mineral be-
comes brighter when Us+ predominates, Luminescence can be observed in the bright varieties of umohoite,
whereas no radiation is emitted from the_dark varieties,
Uranium hydroxides form a group of minerals having the formula Mex (UOZ)yO2x(OH)2((yy_x)] ? [z + x
+ y] ? H2O, wherein Me stands for Ca, Ba, or Pb. The luminescence spectra of the hydroxides consist of
three highly diffuse bands (Fig? 4d). The luminescence spectra of the hydroxides are distinguished from
those of the molybdates by the absence of the band at 19,000 cm-1. An influence of the Me2+ cation upon
the radiation centers could not be observed in hydroxides, X-ray diffraction patterns have revealed that
hydrogen-oxygen molecules of various forms are the principal connecting link in the cementing of the
layer of U hydroxides, It seems that a polymerization of the uranyl-hydroxyl radicals [UOZ(OH)]z+ takes
place. It is generally accepted [10] that the polymerization leads to an attenuation of the luminescence at
298?K and causes ayellow-orange luminescence at 77?K.
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 Declassified and Approved For Release 2013/01/25 :CIA-RDP10-021968000400030001-7
Recognition of Uranyl Minerals
Considerable difficulties are encountered in the recognition of secondary uranium minerals, because
these minerals are similar in their color, optical constants, type of aggregation, and paragenesis, Fur-
thermore, the difficulties originate from the problem to isolate the minerals in pure form in amounts which
suffice for chemical analysis, The possibilities of Debye diffraction patterns are limited because extensive
work must be done to obtain these patterns, though their reproducibility is low. It is therefore most con-
venient to use the luminescence technique in the beginning for the recognition of the minerals, particularly
since the minerals are not destroyed and need be present in amounts of only a few milligrams, In order to
supplement the previously known [1-3] luminescence determinations at room temperature, we propose to
make luminescence measurements at low temperatures, because a greater number of luminescent minerals
can be studied at low temperatures and the accuracy of the measurements increases considerably. It is
then usually possible to unambiguously recognize a particular mineral or to determine at least the crystal-
lochemical group of related minerals with the same structure, The standard luminescence spectra of Figs,
1-4 or the characteristics listed in Table 1 can be used for the determination of the minerals.
The proper choice between two or three minerals of a single class is the greatest problem in routine
mineralogical determinations,. Typical errors are made in the pairs uranophane-~3-uranotil, uranopilite
-zippeite, carnotite-calciocarnotite, Low-temperature luminescence spectra are particularly useful in
similar cases,
1. V. G. Melkov and L. Ch. Puklial~skii, Search for Uranium Deposits [in RussianJ, Gosgeoltekhizdat,
Moscow (1957).
2. M, I; Chaplygin, in: Geochemical Exploration of Ore Deposits in the USSR [in Russian], V. I, Kras-
nikov (editor), Gosgeoltekhizdat, Moscow (1957),, p. 377.
3. E, Z. Bur~yanova, Identification Manual of Uranium and Thorium Minerals [in Russian], Nedra,
Moscow (1972) .
4. E, Rabinovich and R. Belford, Spectroscopy and Photochemistry of Uranyl Compounds [Russian
translation], Atomizdat, Moscow (1968).
5. W, Tufar, Neues Jahrb. Mineral, Abhandl. , 106, No, 2, 191 (1967).
6. B. S, Gorobets, et al, , in: Radioactive Elements in Rocks. Abstracts of the Reports of the 1st All-
Union Radiochemical Conference [in Russian], Press of the Inst, of Geology and Geophysics of the
Siberian Division of the Academy of Sciences of the USSR, Novosibirsk (1972), p, 150.
7. E, S, Makarov and V. - I, Ivanov, Dokl. Akad, Nauk SSSR, 132, No, 1, 673 (1960).
8. M, Ross, et al. , Amer, Mineralogist, 49, 1603 (1964).
9. T, S, Dobrolyubskay;a, Zh, Prikl, Spektr, , 15, No, 4, 642 (1971).
10. T. S, Dobrolyubskaya, Luminescence Techniques of Uranium Determinations [in Russian], Nauka,
Moscow (1968).
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 Declassified and Approved For Release 2013/01/25 :CIA-RDP10-021968000400030001-7
ION FIELD-EMISSION MICROSCOPY OF URANIUM,
PRELIMINARY RESULTS
A,
L,
Suvorov, G. M, Kukavadze,
T.
L.
Razinkova, B. V. Sharov,
V.
A.
Fedorchenko, A, F, Bobkov,
and
B,
Ya, Kuznetsov
The unique possibilities provided by ion field-emission microscopy (first of all, the resolution on the
atomic scale on surfaces of objects examined) are nowadays by no means fully used. The reason is that
it is difficult to obtain ion field-emission images of nonstandard materials (which are of greatest importance
for practical applications). Furthermore, it is difficult to interpret the resulting images and to identify on
these images atoms of a particular species, molecular complexes, defects of various types, etc.'
Analyses by ion field-emission microscopy were not made on new materials (particularly fission ma-
terials), mainly because a complete theory of image formation in field-emission microscopes is missing
and because the principles on which the evaporation by a field is based are not yet fully understood. The
present state of the theory and the most important papers which resulted in progress in this field have been
discussed in a review [1]. '
The goal of the first stage of our work was to determine the possibilities provided by field-emission
microscopical investigations of uranium samples, to find the most efficient conditions for the analysis of
these samples, and to develop several related methodological details. No doubt, research on the crystal
structure, defects, and the dynamics of several processes occurring on both the surface and in the bulk of
uranium (and other fission materials) can render new information, when field-emission microscopes are
employed,
Except for work on the field-emission microscopy of uranium dioxide [2], there are practically no
publications on the ion field-emission or electron field-emission microscopy of fission materials and, par-
ticularly, uranium. The work must concern primarily the ion field-emission microscopical investigation
of alloys, since the results obtained in. this area are of greatest importance.,
Fig. 1. Optical image of the profile of a typical symmetric
uranium point (x600).
Translated from Atomnaya Energiya, Vol. 36, No. 1, pp. 14-18, January, 1974. Original article
submitted March 6, 1973.
m 1974 Consultaats /3ureau, a division o f Plenum Publi.shiag CorporaGi.on., 227 Quest 17th Street, New fork, iV. Y. 10011.
No part of ~tltis publiratfon may be reproduced, stored in a retrieval system, or transmitted, in any form or by any_ineans,
electronic, mecltanrcal, pleotocopying, micro/ilmi.ag, recording or otherwise; wi,tlaouG written permission of tlt.e pu.blis{aer. a
copy o/ this article is available from the publisher for .$15.00.
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 Declassified and Approved For Release 2013/01/25 :CIA-RDP10-021968000400030001-7
Fig, 2 Fig, 3
Fig, 2. Ion field-emission image of the surface of a sample of c~-uranium; the image was ob-
tained in a gas mixture of He + 0.5`J~ HZ at T = 78?K. The (010) pole protrudes into the cen-
ter of the image.
Fig, 3. Electron field-emission image of a sample of ~-uranium which was smoothed by heat-
ing and additionally purified by field-induced evaporation in vacuum,
1. Samples, We used in our work as initial materials wire with a diameter of 0.3 mm (technical,
natural uranium) for the samples which had the form of sharp points. We employed the well-known tech-
nique of automatic electrochemical etching [3]. The electrolyte consisted of 12~ by weight chromium an-
hydride, 8~ by weight orthophosphoric acid, and a small amount of water, The etching was effected in two
stages at a constant voltage, In the first stage, the diameter of the wire was reduced to 0.1-0.05 mm (U
= 12 V, Ietch ~ 100 mA); in the second stage, the voltage was reduced to 4 V in proportion to the decrease
in the current (I ~ 30-5 ?A), Sharp points with a typical radius of curvature of less than 1000 A at the tip
were formed (Fig, 1). When the etching was effected with a variable voltage, the surfaces of the resulting
points were rough and "pitted," After the etching, the finished points were washed in alcohol.
2. Experimental Setup. All experiments were made. in the semimetallic sectional field-emission
microscope of [4] at a sample temperature of 78?K, The vacuum system of the setup was modernized and
made it possible to work with spectrally pure H2, He, Ar, and' Ne, and mixtures thereof as image-forming
gases. The partial pressures of the gases were monitored with the aid of the omegatron of (5]. First, an
oil-diffusion pump was used for the pumping and, afterwards, an electric discharge sorption pump, The
entire system was heated, The initial minimum pressure amounted to about 5 ? 10-$ mm Hg.
Athree-chamber electro-optical UM-92 converter with magnetic focusing [6] was used to enhance the
brightness of the field-emission image, The images were photographed with a "Zenith B" camera having
a "Helios-40" lens, The photographs were made on fluorographic RF-3 film, The exposure time depended
upon the gas used and amounted to 1-30 sec,
The electron field-emission images of uranium samples were obtained in a sealed electron field-em-
ission microscope of glass, The highest vacuum was in this case ~l ? 10-to mm Hg,
3. Experimental Results and Their Discussion. Theoretical estimates of the evaporating field, which
were given in [7] for uranium in vacuum at T = 0?K, rendered the value Fevap = 424 MV/c m, This means
that the use of pure helium as the image-producing gas does not provide a stable ion field-emission image,
because the field which is required for autoionization of helium is 450 MV/cm, i, e, , the uranium sample
evaporates while the reflection image of the sample is made. Neon (Fimage = 370 MV/c m) or argon
(Fimage = 230 MV/c m) are more advantageous from the theoretical viewpoint. However, it is generally
accepted that a satisfactory, stable ion field-emission image of a sample depends not only upon the rela-
tion between the image-producing and the evaporating fields, Of additional importance is the dynamics of
the field-induced evaporation process which is introduced for the final polishing of the sample surface to be
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201 ? : Z11; .. ~ ~? .
.. ~ 221
? . ,? ??
? ~
~.
10l ,~ ?212 ????,,111 .. ~~ . . ?
? ????
? 12l?? ?!
? M? ? ? ? r ? ? ? f?? ?
? ? ~ ? - ? f,i1 ? r
??
i02 ?~~ ? ?21~? ? 112 y ~ '?1Z2 ??
? ??
... '? ?? ?
? ?.
..ti?'I13ti ??:? ? ? ?: ?? ? ? ??
? ti ? ? ? ?? ? ? ? ?
? ?~ ?
e- ? ? ? ?: ?. .. ?
001 ~ ?~ ?~ ?-?g j?? ? `?' i? ~a? ~f ??a~ ??; ?? e ? ? t ~ ? i? _
01Z Ofl 02f C3.- 0I0
Fig, 4. Computer-calculated model of a sector of the ion
field-emission image of an uranium sample oriented paral-
1e1 to the [001] axis.
projected. The conditions of field-induced evaporation determine the final form of the sample surface.
,Since no complete theory is available at the present time, the corresponding conditions can be correctly
chosen only on an empirical basis derived from numerous experimental data and general concepts. The
highly efficient technique of interaction ("f acilitatingTM or "promoting") between field-induced evaporation
and autoionization is in a similar development stage [8, 9].
We used Hz, He, Ne, Ar, and gas mixtures of He + 0.5~ H2, He-Ne, and Ne-Ar as image-forming
gases in research on uranium samples.
Helium. The first good result was that high mechanical strength of the needle-like uranium samples
could be obtained. The samples sustained the tensile stresses generated by fields of the order of 500-550
MV/cm. A total of about 50 samples were analyzed four of which, i, e. , 896, were destroyed by the inter-
action with the field. This proves that the theoretical strength of materials [10] can be obtained in samples
used for autoionization microscopical analysis. Thus, there exist unique possibilities of verifying the the-
ory. The destruction of the samples can be explained by their symmetry [ll] which results in a shearing
stress component.
Moreover, our results make it appear doubtful that the above estimates of the evaporating field are
reliable. The figures are apparently higher in the case of real samples. This conclusion was drawn in ob-
servations of stable helium autoionization images of uranium, which were obtained after field-induced evap-
oration at increased temperatures. However, information on the crystal structure of the samples cannot
be deduced from the resultCng images, though a certain symmetry can be recognized on some of the images
and ?though some sets of bright spots can be tentatively identified with rings of plane nets (edges of crystal-
lographic planes).
Argon. Argon was more efficient as the image-forming gas. The images obtained with argon exhibit
to an increased extent the "tendency to the formation of aring-shaped structure." However, we must re-
call that the argon autoionization images were unstable in all cases. This can be explained by the inadequate
vacuum conditions which in the case of field-emission microscopy with argon played a deciding role. (The
examples of autoionization images of aluminum samples are very significant in this context [12, 13];
changes in the vacuum resulted~in a sharp improvement of both quality and resolution of the autoionization .
images obtained with argon.)
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Neon, The autoionization images which were obtained of uranium samples in pure neon had a quality
similar to those obtained in argon, but were characterized by increased stability, The brightness of the
images obtained in neon is by about an order of magnitude lower than the brightness of the images obtained
with argon or helium,
The results of the experiments which were,made.with pure gases in the above-indicated vacua lead to
the .general conclusion that the field-induced evaporation in these gases does not result in a final form of the
tip of the point required for a detailed reflection .pattern of the point's surface,
The best results were obtained when helium with added hydrogen was used as the working gas, The
preliminary evaporation of the sample with the aid of a field was effected in this gas mixture and also the
~~71
1 I 1
I
2 52l
~lD 421 J7I i~,
32 41/~
-f
u
7J
7:
C
75
poi (_,.--~-.~_ _ za;
Jy
21 21 T21T
~-z31
)47~
4/~ X4/3
301401,
4~ i/ 3li
142
PJi
%32 ~~
150
74D
130
051 Zf0
091 171
cil al zil
'CSZ ?52 ri1 _..
Ofl
J12 ~14i
~I04 IOi
f
O~PS 12541IJ 11c21~4 42 Y11
124 /73 T34
4
IJ4
074 ? 77 233 NlJi Z1
113 241 2J2
J3l
ai X743 132 121 142 ~J41
~c7z zJl
241
X052 251
Oil
r041
X051
:i
X117
~J14
203 J04
r~ ~13J17
J22 421
32 /
??, oGi:
J i1~
217
177
:1 ~JI/471
1. I
7
3ab~
P30
4JB
J20
210
520
110
410
510
Fig, 5, Standard stereographic projections of cx-urani-
um upon a) the (010) plane; b) the (001) plane.
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 Declassified and Approved For Release 2013/01/25 :CIA-RDP10-021968000400030001-7
ion-emission image was obtained. Atypical ion field-emission image is shown in Fig: 2. This image
made it possible to determine the crystallographic orientation of the sample and to make the crystal faces
appear. The images obtained were rather stable. It seems that a further improvement of the image qual-
ity will depend upon a careful selection of the amount of hydrogen added and, no doubt; upon improvement
of the vacuum.
Attempts to obtain ion field-emission images in pure hydrogen and in He-Ne and Ne-Ar gas mix-
tures (concentration 50 :50) were unsuccessful.
The possible plastic deformation and partial destruction of the uranium samples in the field-emis-
sion microscope imply that distorted rings of the edges of crystallographic planes, which are unresolved
in their atomic details, appear on the image. Moreover, typical streaks of contrast, which were several
times discussed [14], appear on all these images. Since asymmetry can be the reason for the deformation
.and the destruction of the tip in the ion field-emission microscope -with certain contrasts developing
- checking the samples (preliminary checks under an optical microscope and subsequent checks after the
analysis in the ion field-emission microscope by means of an electron microscope) is not only desirable
but also necessary for the correct interpretation of the ion field-emission images.
Figure 3 is the electron field-emission image of an uranium tip which had been smoothed by heating
and additionally cleaned by field-induced evaporation in vacuum. Though this image cannot be considered
satisfactory, the image still provides some information. With the aid of the methods of electron field-
emission microscopy, one can obtain even in the present stage some information on the emission charac-
teristics of uranium.
4. The Interpretation of the Images. The interpretation of the images, particularly the determina-
tion of the crystal orientation of samples and the. indexing of the crystal faces, were effected in two ways:
modeling of ion field-emission images, and calculation of standard projections.
Figure 4shows acomputer-calculated model of an ion field-emission image of an a-uranium sample
which was oriented parallel to the [001] axis. The description of the program used for the calculations is
included in [15]. A qualitative comparison of the calculated models and the real ion field-emission images
makes it possible to determine the crystallographic orientation of the sample. Moreover, the modeling of
ion field-emission images of uranium samples must evidently aid in the subsequent understanding of the
field-induced evaporation process (obtention of the required final form of the tip). The interpretation of
the ion field-emission images by means of the defect structure of uranium cannot be completed without
modeling. ,
Three stereographic projections of a-uranium (Fig. 5) on the (001), (010), and (110) planes were cal-
culated for indexing the faces on the images of the uranium samples. Our interpretation of the ion
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field-emission images has shown that a (010) pole protrudes into the center of the image, This effect is
related to the technology of preparing the samples for the investigations. A wire with a diameter of 0.3
mm is usually obtained [16] by wire drawing at low temperatures (150-200?C), which renders a texture of
the {0.10} type,
The calculations were made for the faces with the low indices (it is well known that the morphological
importance of a crystal face is inversely proportional to the quantity S = hz + kz + ZZ, where h, k, and Z
denote the Miller indices), The calculations were extended to the {521} faces. The calculated stereograph-
ic projections made it ~possible~to determine the crystallographic orientation of the uranium samples ex-
amined and to provide the main faces with indices (see Fig, 2),
The authors thank Yu. G. Abov for his inte"rest in the present work.
? LITERATURE CITED
1. A. L. Suvorov and V. V. Trebukhovskii, Usp, Fiz. Nauk, 107, 657 (1972).
2. R. Morgan, J. Mater. Sci., 5, 445_(1970).
3. M. I. Elinson; V. A. Gortkov, -and G. F, Vasil~ev, Radiotekhnika i Elektronika, 11, 204 (1957).
4. V. A. Kuznetsov, G. M. Kukavadze, and A. L. Suvorov, Pribory i Tekh, Eksperim. , No. 2, 152
(1969).
5. A. P, Averina, Pribory i Tekh. Eksperim. , No, 3., 123 (1962).
6. M, M. Butslov et al. , Pribory i Tekh. Eksperim. , No. 6, .137 (1971).
7, E, V. Muller, Usp, Fiz, Nauk, 92, 293 (1967).
8. E, Muller et al. , J. Appl. Phys, , 36, 2496 (1965).
9. E. V. Muller, in: The Field--Emission Microscope [Russian translation], Mir, Moscow (1971), p. 94.
10. R. I, Garber, Zh, I, Dranova, and I, M. Mikhailovskii, Fizika Metallov i Metallovedenie, 30, 445
(1970). "
11. E, V. Muller, Usp, Fiz. Nauk, 77, 481 (1962).
12. R. Morgan, J. Mater, Sci? 7, 361 (1972). ,
13. P. Turner, B. Regan, and M Southon, Electron Microscopy and Analysis, The Institute of Physics,
London-Bristol (1971), p, -252.
14. S. Ranganatan, in: The Field-Emission Microscope [Russian translation]; Mir, Moscow (1971),
p. 127.
15. T. L. Razinkova, A. G. Sokolov, and A. L. Suvorov, Automation of Scientific Investigations [in Rus-
sian], Zinatne, Riga (1972), p. 208.
16. Yu. N. Sokurskii, Ya, M, Sterlin, and V. A, Fedorchenko, Uranium and its Alloys (in Russian],
Atomizdat, Moscow (1971)...
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 Declassified and Approved For Release 2013/01/25 :CIA-RDP10-021968000400030001-7
THE,FLEXURAL STRENGTHS OF DISPERSE MATERIALS
BASED ON URANIUM AND MOLYBDENUM DIOXIDES
BETWEEN 293 AND 1870?K
L. E. Kakushadze and R. B. Kotel'nikov UDC 621.039.542.33
The serviceability of disperse fuel elements depends largely on their mechanical strength. The
strengths of ceramic materials are usually measured by means of flexural or compression tests.
We investigated cermets of uranium dioxide particles, -315 + 200 ? diameter, covered with molyb-
denum, and mixtures of uranium dioxide granules with molybdenum powder. The preparation technology
and some properties of these materials were described in a previous article [1]. By hot pressing we first
prepared billets in the form of cylinders 10 mm in diameter and 20 mm long. With a diamond disc, without
cooling liquid, from each billet we cut four rectangular specimens 3 x 3 mm in cross section and 15 mm
long. In the cross section of each such specimen there were about 150 uranium dioxide particles. In the
course of his argument, Tret'yakov [2] remarks that in flexure tests on cermets of the VK (tungsten-cobalt)
hard-alloy type, the admissible ratio between the intersupport spacing and the specimen thickness is 3-4.
In our experiments this quantity was 3.43. The marks left by the diamond cutter were longitudinal in direc-
tion. Specimens with surface defects were smoothed -along the axis with dry M-20 emery paper. In all
cases at least one face was left untouched after diamond cutting. In the tests, the specimen was placed on
the support so that the diamond-cut face was subjected to tension. Thus we can assume that all the speci-
mens had identical mechanical treatment of the surface -diamond
~ cutting with the marks along the principal axis. The specimens were
Fig. 1. Scheme of apparatus
for flexural tests. 1) Tanta-
lum sheet heater; 2) tungsten
loading prism; 3) specimen;
4) tungsten support prisms.;
5) molybdenum block.
not heat-treated after cutting, The density of the specimens was 96
f 1.290 of the theoretical value, The scheme of loading and heating is
shownin Fig. 1. The distance between the tungsten support prisms
was 10.3 mm. The specimen was loaded at 4.08 ? 105 N/mz ? sec (or
2.5 kg/mm2 ? min). The specimen was heated by radiation from a strip
heater through which power-frequency current was passed. To reduce
heat losses, the high-temperature zone was protected by a system of
screens. The specimen surface temperature was measured by means
of an OPPIR-09 optical pyrometer through holes in the screens and
heater, The pyrometer was calibrated against aplatinum-platinum
-rhodium thermocouple attached to the specimen. The working me-
dium for the tests was high-purity helium containing about 0.012 vol, qb
of impurities (according to its rating). Before admission to the ap-
paratus, the helium was repurified by passing it through silica gel,
copper shavings at 1000?K, and silica gel again. The fact that the lat-
tice parameters of the phases of the cermets and the specimen surface
color remained unaltered after the high-temperature experiments
showed that the helium was pure enough for the work. During the ex-
periments, the excess helium pressure in the furnace chamber was
5 ? lOz N/m2 (or 3.7 mm), and the volumetric flow rate through the
chamber (which was about 5 ? 10-3 m3 in volume) was 4.2 ? 10-s m3/sec.
At each temperature we tested at least three or four specimens from
Translated from Atomnaya Energiya, Vol. 36, No. 1, pp. 19-23, January, 1974. Original article
submitted April 12, 1973.
? 1974 Consultants Bureau, a di.visioa of Plenum Publishing Corporation, 227 [('est 17th Street, New )'or/r., !V. }'. 10011.
No part of t{ti.s publication may be reprodu-ced, stored i~n a retri.eual system, or transmitted, in nay form or by aa~}' means,
electronic, mechanical, photocopying, micro/ilmi.ag; recordirt~ or otherwise, witltou.t written permission o/ the publisher. ;l
copy o/ this article i.s auai able /rom the pu.hlisleer /or ~IS.00.
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Fig, 2. Static flexural strength of cermets of uranium di-
oxide and molybdenum vs composition at various tempera-
tures in ?K: a) 293; b) 1240; c) 1570; d) 1870. 1) cer-
met of molybdenum-coated particles; 2) mixed cermet,
a given batch, The flexural strength was calculated by means of the usual formula, On inspecting the bro-
ken specimens we found no appreciable bending even after tests at 1870?x. Fracture occurred in a plane
perpendicular to the principal axis of the specimen, Regardless of the type of cermet, the composition,
and the test temperature, fracture occurred across the particles of uranium dioxide, There was no prefer-'
ential fracture in the molybdenum interlayers or "husking" of the uranium dioxide particles from the molyb-
denum matrix. The deviations from the mean value, plotted in Figs, 2 and 3, correspond to the root-mean-
square deviations of the experimental points for the given type of cermet, From these results we can draw
the following conclusions,
Over the whole range of compositions and temperatures examined, the coated-particle cermet is
stronger than the mixed cermet,
At temperatures between 293 and 1500?K we observe that the strengths of both types of cermet tend'
to rise with increasing molybdenum concentration. The cause of this is apparently that molybdenum is
stronger than uranium dioxide, as we see from Fig, 4. Here we must bear in mind that the experimentally
measured flexural strength may be several times greater than the tensile strength, despite the fact that
fracture is due to tensile stresses in both cases. For example, Ruderilio [3]; who investigated a material
based on silicon carbide at 290-1670?x, showed that the flexural strength is 2,5 times higher than the ten-
sile strength, Consequently the ten"sile strength of uranium dioxide must be below the values given by
curves 1-3 in Fig, 4.
When the temperature increases from 293 to 1500?K we observe a tendency for the strengths of -both'
types of cermet to increase, On the basis of investigations of a number of substances, Savitskii [4] showed
that the strengths of brittle materials increase on heating owing to the appearance of a certain amount of
plasticity which reduces the stress concentrations and makes it possible for the material to exert its total
strength, The temperature range of maximum strength of brittle materials is 0.5-0.8 times the absolute
melting point for tension, and 0.5-0.7 times the absolute melting point for compression, So we can assert
that the strengths of the cermets increase with temperature, mainly owing to increase in the strength of
the uranium dioxide (see Fig, 4) and decrease of the microstresses in the cermet, The strength of molyb-
denum decreases with rise of temperature, But the strength of the cermets can nevertheless rise over a
certain temperature 'range owing to the increasing strength of the uranium dioxide, because the molybdenum
concentration is not very great, The thermal expansion coefficient of uranium dioxide is about twice that
of molybdenum, Therefore stresses arise in microvolumes of the cermet during cooling of the specimens
in the course of preparation, A similar problem of the stresses in atwo=phase system was solved theoret-
ically by Zaitsev ~[5], and the results were confirmed in investigations of the microstresses in the carbide
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500 1000 1500 ~ ?K 500 1000 1500 7~ ?K
Fig, 3. Static flexural strength of cermets of uranium diox-
ide and molybdenum, of various compositions, vs tempera-
ture, 1) Cermet of molybdenum-coated particles; 2) mixed
cermet, Molybdenum concentrations in cermets; a) 5.2 and
5; b) 10.3 and 10; c) 16.0 and 15; d) 20.3 and 20 vol, ~,
phase of a VK alloy by the x-ray method [6]. The theoretical results for cermets of molybdenum-coated
uranium dioxide particles are plotted in Fig, 5. In the calculation it was assumed that EMo = 2.94 ? 1011
N/m2 [7], EUOZ = 1.71 ? 1011 N/m2 [8], and ??Mo =?U02 = 0.3. Information on the thermal expansion of the
phases was taken from [9] and [10]. During cooling, ~n uranium dioxide undergoing "hydrostatic" pressure
from the molybdenum, all three principal stresses are equal, but in the molybdenum the determining stres-
ses are the shear stresses T, which reach a maximum at the interior surface of the coating lying against
the uranium dioxide, At high temperatures, stress relaxation occurs. When the cermet is cooled below
the recrystallization temperature of molybdenum (about 1270?K); the stress removal process ceases. The
stresses should not exceed the yield point for molybdenum or the breaking strength for uranium dioxide,
At 293 ?K the yield point of molybdenum is 70 ? 10~ N/m2 [7], which is below the calculated-value Tmax ~ 100
? 10~ N/m2 (see Fig. 5). Consequently, on cooling, either the molybdenum should undergo plastic deforma-
tion, or the dioxide.should break, or fracture should occur at the boundary between the uranium dioxide
and the molybdenum, .From Figs, 4 and 5 we see that the. stresses in the uranium dioxide particles, at any
rate for molybdenum concentrations less than 10 vol. qn, exceed even the flexural strength, which, as "we
have stated, is higher than the tensile strength. Despite the decrease of these stresses due to plastic de-
formation of the molybdenum, they remain high, It was on account of precisely these stresses that, in all
the prepared cermets we observed particles of uranium dioxide broken by cracks, as we see, for example,
on a photograph of the microstructure (see figure in [1]). During heating, the tensile stresses in the ura-
nium dioxide particles decrease, and at a certain temperature compressive stresses, which of course are
much better withstood by brittle materials, can arise in them, In this state the cermet can be likened to
prestressed concrete. Under the compressive stresses, the cracks, which arose in the uranium dioxide
particles during cooling, begin to grow. Therefore the strength of the cermets increases as the tempera-
ture rises to a certain limit,
As the temperature increases from 1500 to 1870?x, for cermets with molybdenum concentrations of
5 and 10 vol. ~ we observe a tendency to increase of strength, whereas the strength of cermets with molyb-
denum concentrations of 15 and 20 vol, ;~ falls. This phenomenon is apparently due to a decrease in the
strength of pure molybdenum to a value below that of uranium dioxide. It must be borne in mind that the
area of contact between the uranium dioxide particles and the molybdenum is four times as great as the
diametral cross section of the former. Therefore fracture does not at all have to occur in the weakest
phase, i, e? , the molybdenum. In tests throughout the investigated temperature range, fracture occurs
across the uranium dioxide particles, as we have already stated. A decrease in the strength of the molyb-
denum has little effect on the strength of specimens in which its concentration is low, but specimens in
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 Declassified and Approved For Release 2013/01/25 :CIA-RDP10-021968000400030001-7
(
Curve
Mat-
erial
Method of pre-
paration
(Characteristics I
'
;, ~
~ [
of material
a ar_
1
uoz
~ ~
Y=84.3%theor.
Particle dimen-
sions of original
powder, 10 -l5?
z
uoz
Cold-pressed and
Y = 909? theor.
sintered in argon at
Particle dimen-
2270?IC fot 30 min,
sions of original
~~ it
Specimen dimen-
0 - 5
d
:sions: 38.1xb.4
?
er,
ow
P
3
u02
x3.2 mm
y= 82. 8'/~~ theor.
Particle dimen-
sions of original
powder, 0 - S?
?
Pressed ppowder,
~
s~,ntered, hot-
rolledrecrysta1=
lized
Purity of orig-
'
l
na
powder,
[fs]
s
Same material
99,98%
withoutrecrystal-
,
lization
Curve
t l a
a e
s s
7
s
Mocon-
5
f0
i5
2C
centratio
voL %
I
Phase I
uoz I
-to uoz
4
nto
I uoz
Mo
uoz
Mo
Fig, 5
Fig, 4, Temperature dependences of flexural strength _of uranium dioxide and tensile strength
of molybdenum.
Fig, 5. Thermal stresses in phases of cermets at 293?K vs the temperatures at which they
cease to relax during cooling,
which its concentration i"s high lose strength, Furthermore, molybdenum present in cermets, as suggested
with some justification by us to explain the conductivity [1], must contain uranium and oxygen from the di-
oxide as impurities, The, strength of such a solid solution at high temperatures may be higher than that of
pure molybdenum, Since impurities penetrate the molybdenum layers only to a certain depth, if the molyb-
denum concentration is low the layers will consist wholly of solid solution, but if it is high they will con-
sist largely of pure molybdenum, Therefore the influence of the impurities decreases as the molybdenum
concentration increases, and the strength at high temperatures thus falls.
1.
2.
3.
4.
LITERATURE CITED
L,
E,
Kakushadze' and R. B, Kotel'nikov; At. Energ. , 36, No. 1 (1974).
V.
I.
Tret'yakov, .Cermet Solid Solutions [in Russian], Metallurgizdat, Moscow (1962).
V.
I,
Rudenko, Poroshkovaya Met. , No. 4, 86 (1961).
E,
M.
Savitskii, The Influence of Temperature on the Mechanical Properties of Metals and Alloys
[in Russian], Izd-vo AN SSSR,, Moscow (1957).
5. G. P. Zaitsev, Fiz. Metallov i Metallovedenie, 2, No. 3, 494 (1956).
6. A. E, Koval'skii et al, , in: Data on Metallurgy and the Technology of Preparation of Cermet Solid
Solutions, Refractory Metals, and Materials Based on Them [in Russian], V. F, Funke (editor), Izd.
TsIItsvetmet, Pt, 2 (1963), p, 29.
7. A, K. Natanson (editor), Molybdenum [Russian translation], IL, Moscow (1959).
8. Yu. N. Sokurskii (editor), Materials for Nuclear Reactors [Russian translation], Gosatomizdat,
Moscow (1963).
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 Declassified and Approved For Release 2013/01/25 :CIA-RDP10-021968000400030001-7
9.
J. Belle (editor), Uranium Dioxide, USAEC, Washington (1961).
10.
A. E, Coldstith, A Handbook of Thermophysical Properties of Solid Materials, Vols, I-III, Perga-
mon Press, Oxford-Paris (1961).
11.
M. Burdiek and H. Parker, J. Amer. Ceram, Soc. , 39, No, 5, 180 (1956).
12.
B. L, Mordaik, Probl. Sovrem. Metallurgii, No. 4, 149 (1960).
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 Declassified and Approved For Release 2013/01/25 :CIA-RDP10-021968000400030001-7
RADIATION-INDUCED SWELLING OF OKh16N1'5M3B STEEL
V.
N.
B.ykov, A..
G.
Vakhtin,
V.
D.'
Dmitriev,
L.
G. Kostromin,
A.
Ya.
Ladygin,
and
V. I. Shcherbak
A large number of published papers have been devoted to studying the radiation-induced porosity of
the austenitic steels used as construction materials in.the active zone of fast reactors. Nevertheless, there
is still no reliable information as to the temperature and dose dependence of the radiation-induced swelling
of OKh16N15M3B steel, which is widely employed as a material for the fuel-element cans of fast reactors.
In this paper we shall present certain results obtained by the electron-microscope .analysis of radia-
tion-induced porosity in OKh16N15M3B steel irradiated with neutrons in the BR-5 reactor.
MATERIALS AND METHOD
The .samples for electron-microscope study were discs 3.5 mm in diameter and 0.4 mm thick, cut
from various parts of fuel-element cans irradiated in the BR-5 reactor with integrated fluxes of 4.3 ?.1022
,.neutrons/cm2 at 430-580?C.~ After preparation the fuel-element cans were annealed at 950?C for 10 min in
vacuo.
The preparation of the._objects for examination- under the electron microscope and the method of ana-
lyzing the results were described earlier [1]. We should only mention that, in contrast to the present in-
vestigation, the .film thickness there lay between 1200 and 2000_ A, the porosity being calculated by another
method [2].
Electron-microscope examination of the irradiated samples revealed the presence of inclusions,
Fig. 1. Microstructure of OKh16N15M3B irradiated with a dose of 36 displacements
/atom at 520?C (x100,000).
Fig. 2. Swelling of OKh16N15M3B steel irradiated with a dose of 30 displacements
/atom as a function of the irradiation temperature.
Translated from Atomnaya Energiya, Vol. 36, No. 1, pp. 24-26, January, 1974. Original article
submitted May 28, 1973.-
? 197d Consultants Bareaa, a diuisi.on o f Plenum. Publishing Corporation, 227 Rest 17th Street; New 1'orh-, N. Y. 10011.
No part of t{zi.s publication. may be reproduced, stored in a retrieval system, or transmitted, is nay (arm or by any means,
electronic, mechanical, p/zotocopyin.g, micro/liming, recording or otherwise, witho~ct written permission of the publis/ter. ,1
copy of this article is available /ram the publislzer for $I;S.00.
24
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 Declassified and Approved For Release 2013/01/25 :CIA-RDP10-021968000400030001-7
cpt? 1O22,neutrons/cmZ
t 2 3 4 5 6
0,4
~~ ~ 0,2 ~
1
100 10 20 JO 40 50
400 450 500 550 Tin C Dose, displacements/atom
' Fig, 3 Fig, 4
Fig, 3, Concentration, mean diameter (O), and maximum diameter
(p) of cavities in OKh16N15M3B steel irradiated with a dose of 30 dis-
placements/atom, as functions of the irradiation temperature.
Fig, 4. Dose dependence of the swelling of OKh16N15M3B steel at
525?C,
loops, and cavities, the dimensions and concentrations of these varying with the conditions of irradiation,
Figure 1 shows the microstructure of OKh16N15M3B steel irradiated with an integrated flux of 4.3 ? 1022
neutrons/cm2 at 520?C, The concentrations p~,, mean diameter dv, and total volume of the cavities DV/V
(the latter being taken as equal to the total swelling of the steel) calculated on the basis of the electron mi-
crographs are illustrated in Figs, 2-5. The size distribution of the cavities is shown in relation to the
dose of irradiation in Fig, 6.
Since the experimental results as to the radiation-induced swelling were obtained for samples ir-
radiated in neutron fluxes having a variety of spectral characteristics, the experimental data were com-
pared by reference to the number of displacements per atom (kt) (d/a) rather than by reference to the in-
tegrated flux, For this calculation we used a modified displacement model developed for the irradiation of
materials by fast neutrons [3]. For comparison, Figs, 4 and 5 also show the integrated neutron fluxes for
an energy spectrum corresponding to the center of the active zone in the BR-5 reactor,
Experimental examination shows that the radiation-induced swelling of steels depends on a large num-
ber of factors, primarily the irradiation dose and temperature, It is accordingly of particular interest to
devote more detailed attention to the way in which the total volume of the cavities depends on these factors,
Temperature Dependence of the Swelling, The total volume of the cavities first increases with tem-
perature (Fig, 2), reaching a value of 1.4y~ for 30 displacements/atom, and then falls, i, e, , the tempera-
ture dependence of OV/Vhas abell-shaped form with a maximum in the region of 500?C, The increase in
OV/V on raising the temperature from 430 to 500?C (Fig, 3) is associated with an increase in 'the dimen-
sions of the cavities, since the concentration of the cavities has a tendency to fall with rising temperature
in this region, On further raising the irradiation temperature to 560?C the concentrations and mean diam-
eter of the cavities both diminish, For an irradiation temperature of over 560?C the concentration and
mean diameter of the cavities hardly depend on the temperature at all - if anything they increase slightly.
The dependence of pv on T may be explained as being due to a fall in the rate of formation of the cavity nu-
clei with increasing temperature, owing to a reduction in the degree of supersaturation of the matrix with
point defects,
As regards the physical reasons for the change in the mean size of the cavities, the following should
be noted, It follows from the expression for the rate of growth of a vacancy-type cavity of radius r [4]
Izvs~
az - r ~(C?DU - CIDI) - DS exp 1 rkT ~ ~ '
where CvDv and CIDI are the diffusion flows of vacancies and interstitial atoms at the surface of the cavity,
Ds is the self-diffusion coefficient, y is the surface energy, and St is the atomic volume, that the quantity
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 Declassified and Approved For Release 2013/01/25 :CIA-RDP10-021968000400030001-7
-- - -~
p~, cm_3
b r-?
~pt? 1O22,neutrons/cm2
2 3 4 5 6
o
a
faa~
f0 20 30 k0 SO
1012 ~
0
Fig, 5 Fig, 6
Fig, 5. Dose dependence of the mean diameter and concentration of the cavities
in OKh16N15M3B steel at 525?C,
Fig, 6. Size distribution of the cavities in OKh16N15M3B steel irradiated with
doses of 36 (v), 30 (O), and 12 (D) displacements/atom at 525?C,
dr/dt is determined by the ratio of two terms: the first describes the difference in the diffusive flows of
paint defects at the surface of the cavity, the second describes the thermal annealing of the cavity,
. On increasing the irradiation temperature from 430 to 525?C, the radius of the cavity may also in-
crease owing to the increase in the excess flow of vacancies to the surface of the pore as a result of the
increase in the mobility of the vacancies, despite the reduction in the concentration of these as the tem-
perature rises, At the same time, the second term in the equation for dr/dt, always negative for vacancy
pores, also increases with rising irradiation temperature, owing to the increase in the self-diffusion coef-
ficient,
The reduction in the supersaturation of the matrix with point defects and the intensification of the
thermal dissociation of the cavities may be used to explain the reduction in the mean diameter of the cavi- .
ties for OKh16N15M3B steel for irradiation temperatures above 525?C,
However, from 560?C the character of the temperature dependence of p~, and dv changes once again:
on further raising the temperature the concentration and mean diameter of the cavities hardly depend on
temperature at all, This kind of change in the kinetic characteristics representing the generation and
growth of the cavities may presumably be explained as being due to the stabilizing influence of the helium
formed as a result of (n, cz)-reactions in the steel. under consideration, Estimates based on the van der
Waals equation, in fact, showed that the stabilization of these cavities at 560?C required 2 ? 10-3 at?~ of he-
lium for a value of v = 1500 erg/cmZ and a van der Waals constant of 4.10-23 cm3/atom He, We note that .
recent work on the experimental determination of the, amount of helium in irradiated steels of the 304 [5]
and 347 [6] types has shown that the helium content of these materials (for comparable dose of irradiation)
greatly exceeds the calculated contents,- being equal to 3 ? 10-Z and (9-10) ? 10-Z at, ;~, i, e, , an order of mag-
nitude greater than is required for .the stabilization of the cavities in our own case,
Dose Dependence of the Swelling, The volume increment due to the formation of cavities in both
OKh16N15M3B steel and steel of the 316 type [7] varies with the irradiation dose in accordance with a power
law (Fig, 4), The graphically-determined power index for an irradiation temperature of 525?C is slightly
less for our present material than for steel 316 and equals 1,5. Extrapolation of the expe-rimental data to
a dose of 85 displacements/atom, corresponding to an integrated flux of 1023 neutrons/cmZ, shows that, on
allowing for the foregoing power index, the maximum swelling of the steel for this dose may amount to
6-7~. ,
The increase in LTV/V with increasing kt is chiefly associated with the development of new cavities
(Fig, 5), since the mean diameter of the cavities hardly depends on the dose at all,
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We see from the size distributions of the cavities illustrated in Fig, 6 (for irradiation with a variety
of doses at the same temperature) that the size dependence of p~, is of a complex nature, The reason for
this may lie in the effects of a whole series of factors (stress, the presence of gas, and so on) on the gen-
eration and growth of the cavities,
Analysis of the changes taking place in 'the size-distribution curves of the cavities with increasing ir-
radiation dose indicates that with increasing number of displacements per atom the character of the rela-
tionship between p~, and d remains almost constant; there is only a slight rise in these curves in the direc-
tion of greater concentrations and a_slight displacement in the direction of greater sizes, Thus the experi-
mental results indicate that, on irradiating OKh16N15M3B steel, the development of porosity takes place
chiefly by virtue of the growth of new cavity nuclei,
LITERATURE CITED
1. V. N. Bykov et al, , At, Energ, , 34, No, 4, 247 (19'73),
2. E, Wolff, Metallography, 2, 89 (1969). '
3, K. Ohmae and B. Hida, J. Nucl, Mater, , 42, 86 (1972).
4. D, Norris, Radiation Effects, 14, 1-37 (1972).
5. K, Robins, J. Nucl, Mater, , 33, 102 (1969),
6, A. Bauer and M, Kangilaski, J: Nucl, Mater. , 42, 91-95 (1972).
7. H. Brager and I, Straalsund, Trans. Amer, Nucl, Soc. , 15, 725 {1973).
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 Declassified and Approved For Release 2013/01/25 :CIA-RDP10-021968000400030001-7
UTILIZATION OF PULSED SORPTION COLUMNS FOR THE
DECONTAMINATION OF LIQUID RADIOACTIVE WASTES
F. V. Rauzen, E, I, Zakharov,
B, E, Ryabchikov, V. D, Konorchenko,
and E, G. Odintsova
UDC 621.039.325
An increase in the efficiency and economy of the decontamination of liquid radioactive wastes by an
ion exchange process can be attained by improvement. in its technology and instrumentation, Pulsed col-
umns with special partition plates, developed in the USSR [1-2], which are prospective ion-excfiange ap-
paratus, are applied in a number of hydrometallurgical processes, including the decontamination of waste
liquids [3, 4]. In industry, continuous ion-exchange installations operate with a performance of up to 300
m3/h, utilizing columns up to 3.4.m in diameter, The advantages of these apparatus are the small once-
only charging of the sorbent into the installation (5-15 times smaller than into filters), high specific per-
formance (Wp = 35-45 m3/mzh), small size, and simplicity of design and control [5j.
The purpose of the present study was an examination of the possibility for applying these apparatus to
the decontamination of liquid radioactive wastes,
A pulsed counterflow column with multilayer column packing (MCP) is a vertical cylinder inside
which downfall plates are mounted; settling zones for separating the sorbent and the solution are spaced
from top to bottom (Fig, 1),
Connections for the pulse chamber and the air-lift, which pumps the resin to the connecting pipe for.
the inlet and outlet of the sorbent and the solution, are found in the zones. Downfall plates with a 40-60~
transfer cross section in conjunction with the pulsations (reciprocating oscillation of the liquid in the col-
umn) provide a better distribution of the components over the cross section of large apparatus with little
longitudinal mixing,
During, operation of the column, the sorbent is fed into the upper zone, from which it enters the
packed section, where it moves owing to the difference in the densities of the phases (as a pseudoliquid) in
the flow of the solution, '
The saturated sorbent is collected in the lower settling zone, from which it is pumped by the air-lift
into the next device, as shown in the diagram, Draining from the upper zone, the solution proceeds to fur-
ther treatment,
A continuous-action sorption facility usually consists of three columns: sorption, regeneration, and
washing (Fig, 2), Separators are used to recover the solution conveying the sorbent in the air-lift, The
sorbent moves from .the separators -into the next column, but the transporting solution returns to the ap-
paratus,
During pulsation, the solution is intensively agitated by each plate; therefore the rate of diffusion of
the ions in the solution does not restrict the process and the rate of exchange. Because of this, the requis-
ite contact time of the sorbent with the solution and its charging time are reduced,
The velocity of the, solution in the column must be somewhat lower than that, for' which a small frac-
tion of the sorbent can be removed. This velocity (Vs) is 0.8Vo during sorption, where Vo is a characteris-
tic velocity of the sorbent, i, e, , the rate of fall of an average fraction of it in a motionles"s solution,
Translated from Atomnaya $nergiya, Vol, 36, No, 1, pp. 27-31, January, 1974. Original article
submitted April 20, 1972; revision submitted March 6, 1973
?1974 Consultants Bureau, a divisi.oa of Plenum Publi.sitiag Corporation; 227 Ii'est 17th Street, Nezo 3'ork, A'. 1'. 10011.
No part of t/tis publication. may be reproduced, stored t:n a retrieval system, or transmitted, in any form or bq any means,
electronic, mechanical, photocopying, microfilming, recording or otherwise, without wri.tteit permission of the publisher. ,~
copy of this article is auai.lable from. the publisher for $I;i.00.
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 Declassified and Approved For Release 2013/01/25 :CIA-RDP10-021968000400030001-7
Desorbed
solution
Concentrate
~ regenerated
solution
Fig. 1 Fig, 2
Fig, 1. Diagram of the framework of a pulsed sorption column in a test facility; 1)
connecting pipe for emptying the column; 2) lower settling zone; 3) connecting pipe
for recovery of the solution; 4) air rotameter; 5) pulse chamber; '6) valve; 7) pack-
ing zone; 8) column packing; 9) upper settling zone; 10) pressure stabilizer; 11)
pulser; 12) overflow box; 13) funnel; 14) sorbent and solution feeder-separators;
15) valve; 16) screen; 17) solution recovery line; 18) air-lift; i9) raw solution ro-
tameter; 20) pump,
Fig. 2, Diagram of acontinuous-action facility; 1) sorption; 2) washing; 3) regen-
eration columns; 4-6) separators for the recovery of the sorbent from solution; 7)
container for preparation of regenerating solution.
Provided that retention of the sorbent Sl (holdup) in the column for a flow rate ratio n = Qsb/Ql = 1 :50-
1 :500 is small and comprises 0.5-5`,~ (the sorbent occupies a very small part of the volume and the cross
section of the column), the actual velocity of the solution and the specific performance agree numerically.
Therefore, the specific performance of these columns is significantly higher than other apparatus- [6, 7]
.with aquasi-liquid layer of sorbent.
The rate of movement of the sorbent through the apparatus depends on the velocity of the solution:
Vsb = (Vo-Wp)cp, where cp is a coefficient which takes into account the retardation of the movement of the
solution, The coefficient equals one during sorption, when the retardation is small, and 0.6-0.8 during re-
generation and washing, when Sl = 10-309b. Under optimal conditions, the rate of fall of the sorbent is Vs
= 20-30 m/h, which corresponds to a residence time Tsb = 2-3 min/m of column height.
Calculation of Columns. The calculation of pulsed sorption columns is known from the methods ap-
plicable to the calculation of extraction columns [8, 9], since the behavior of the sorbent particles in them
is similar to the behavior of the dripping of the dispersed phase during extraction. Data on-the kinetics and
the statics of the processes, the hydrodynamics of the column, as well as a mathematical model of the ap-
paratus are required for the calculation. Knowing these parameters, one can calculate the diameter and
the height of the apparatus, and then design an industrial facility, using mathematical simulation techniques.
The kinetics and statics of a process have been studied in standard solutions which duplicate the com-
position of solutions from Moscow decontamination plants, in which the cations Na+ and Ca2+ and the anions
Cl- and SOq- comprise the main bulk of the contaminants.
Knowing the kinetics of a process, one can calculate the height equivalent to a theoretical plate
(HETP). From Fig. 3, it is seen that equilibrium is established in the system after X20 min. With a
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I I I I ~I I I I I I__1. D ] Z ,~ 4 5~ 6 7 8
10 20 JO 40 r~, min x, mg - eq/g
Fig. 3 Fig, 4
Fig, 3, Kinetics of the sorption ofcations on K17-2-8 resin (y is the concentration
of ions in the resin): 1) ENa+ + Ca++; 2) Ca++; 3) Na+,
Fig, 4. Determination of the number of theoretical plates (x is the concentration
of ions in the solution):. 1) sorption isotherm for the sum of Na+ and Ca2+ cations
on Kationite KU-2-8 in H+-form; 2) optimal operating curve; 3) operating curve
for test 3; 4) operating curve for test 2.
3 min/m sorbent residence time for these kinetics, there corresponds a calculated HETP = 1.5-2.0 m.
The distribution of cations between the sorbent and the solution is shown in Fig, 4.
The plotting of the operating curve of a process for optimal conditions showed that about 10 theoret-
ical plates (NT P) are required with sufficient impregnation of the sorbent for the complete decontamination
of the solution; the flow rate ratio in this case is n ~ 1 :200.
Thus, a column of height Hopr = NTP x HETP = 15-20 m is required for complete decontamination.
of the solution. Similar dependences are also obtained for other operations involving KU-2 and AV-17 res-
ins, for which the height of the column required for the decontamination of the solution was determined:
Description of Test Facility, The possibility of the application of pulsed columns for the decontami-
nation of liquid radioactive wastes, operational designs, and the optimal operating regions of new appara-
tus were tested in a laboratory facility with a performance of 100-150 liters/h. The processes conducted
in various apparatus (sorption and regeneration) in a continuously operative facility occurred at intervals
and were studied successively in two apparatus, during which the column functioned continuously in each
operation until the establishment of a stationary state,
The test facility consisted of two sorption columns, 76 mm in diameter and 4.8 m high each (since
the housing for the apparatus was not tall enough) operating on KU-2 and AV-17 resins,' sorbent feeder-
separators, regeneration solution preparation tanks, raw and treated water storage tanks, pump, com-
pressor, pulser, and rotameter rack. The framework for each column is shown in Fig, 1.
The columns were situated at different points in order to coordinate the self-flowing movement of the
solution. The sorbent feeder-separators were placed in pairs above the columns, They functioned alter-
nately as feeders and separators of the sorbent and the transporting solution.
In order to simplify the operation of the facility as a pulser, a valveless reciprocating pump with a
performance of 150 liters/h was utilized. In addition the pulse strength J = 600 mm/min in both columns.
In order to stabilize the level of the solution in the pulse chambers of the column, the pulser was actuated
by compressed air from a compressor through a pressure stabilization unit,
Pulsed Column Tests. After rolling and hydraulic tests with tap water, tests were conducted on the
decontamination of liquids and radioactive wastes processable by .the Moscow decontamination station. In
these tests, .the flow rates, concentrations, and radioactivity of the raw solution and the filters were mea-
sured after the cationite and anionite processes in the columns during sorption and regeneration. We de-
termined the cubic content of the resins before and after sorption. The ?sorption-desorption-washing of
resins? cycle was repeated four times. Each test of the sorbent took about 10 h, Regeneration and wash-
ing occupied 1-2 h. The first two cycles were conducted under conditions of free fall for the sorbent, The
operating conditions were altered in the next two cycles in order to increase. the residence time of the sor-
bent in the column.
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 Declassified and Approved For Release 2013/01/25 :CIA-RDP10-021968000400030001-7
TABLE 1. Decontamination of Liquid Wastes with a Low Level of Activity by Pulsed Sorp-
tion Columns and Filters
,
~
~
v
Concentration,.
Resin ca acit ,
P Y
f3-activity
decontamrn- i
~
; '?
~ ~
~
mg-eq/liter ~
mg-eq/g
anon coeffic_,
Test
~
n.~
? ~
a> o ~
~ ~
KU -2 AV - 17
~
NTP
No.
o ~
.v c
o ~
. j m ~
~
.o O
i\a~
Ca2+
CI-
_
S0~
o
~ ~ o
~ ~
~
~
~
N
Y
b0
`~-' "
Y
""' w
For comparison, tests were conducted simultaneously on the decontamination of the same solutions
in sorption filters 69 mm in diameter, with ion-exchange layers 1 and 2 m in height,
Averaged results of tests on pulsed columns and filters are presented in Table 1. From a compari-
son of the data obtained, it is seen that the performance of the columns is 4-5 times greater than that of
the filters. As one should have also expected, the decontamination coefficients in the first two tests of
the columns were found to be smaller than those for the filters, In succeeding tests, it was not possible to
bring the decontamination up to the same (or larger than for the filters) values. The capacity of the sor-
bent in the first two tests is much less, and in subsequent tests approaches the capacity of the filters.
Analysis of the data from hydraulic tests indicates that the performance of the facility is limited by
the throughput of the anionite columns and that one can increase it up to 140-160 liters/h, employing coar-
ser anion-exchange resin. The amount of sorbent found in the column during sorption is 0.5-1.0 liter in
the first tests, and 3-5 liters in the rest, which is much lower than in a filter with the same performance.
The first tests showed that the NTP during sorption of cations is small, and is 1.5-2, which corre-
sponds to an HETP = 2.1-2.7 m, This value is somewhat Larger than the theoretical value, .which can be
explained by the low performance of the cationite column, i. e, , the apparatus operated under nonoptimal
conditions. In order to obtain the prescribed decontamination parameters, a column or a cascade of col-
umns, H = 21-27 m in height, is required. Under these conditions, a pulsed column facility has a smaller
once-only charge of sorbent and higher specific performance than the filters, but its size is greater. In
order to reduce its size, it is necessary to reduce the HETP, which can be achieved by the use of sorbents
with better kinetic properties, for example, macroporous ones, or to increase the residence time of the
sorbent in the apparatus, for which one should increase the retardation of the resin. The next two cycles
were conducted under such conditions, For this purpose, a layer of quasi-liquid resin, 3-3.5 m in height,
maintained either with the aid of an air-lift or a special "transport" pulser, was established in the column
[2] . The layer of sorbent is subdivided by the plates, and the movement of its particles occurs during the
transporting pulses,
The results showed (tests 3 and 4, Table 1) that, with the same performance as in the preceding
tests, the decontamination coefficient and the sorbent capacity increased sharply, The discharge solutions
surpassed in quality those obtained in a facility with filters, The value of the NTP increased to 4-5, and
the HETP was correspondingly reduced to 0.8-1.0 m, Under these conditions, one can obtain the pre-
scribed properties in a column up to 10 m in height,
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 Declassified and Approved For Release 2013/01/25 :CIA-RDP10-021968000400030001-7 .
The regeneration and washing in a system with transport pulsation also proved to be efficient, Thus,
under these conditions, the columns possess advantages in all essentials,
At the present time, designs for columns with transport pulsation, which provide an -even greater
efficiency that allows one to reduce the size of the apparatus, to increase the capacity of the sorbent in
comparison with that of filters, and, correspondingly, to increase the concentration coefficient, are under
development,
Based on the results obtained, the raw data are provided for the design of a combined continuous ac-
tion Facility with a performance of 1 m3/h.
1. S. M..Karpacheva et al. , Chemical and Petroleum Processing Machine Design [in Russian], Izd.
Tsintikhimneftemash, Moscow (1971).
2. B. E. Ryabchikov et al. , Problems in Atomic Science and Technology.. Series: Machinery, Appara-
tus, Means of Mechanization and Automation of Industrial Processes. Pulsed Apparatus [in Russian],
No. 1 (41), Izd. TsNIIatominform, Moscow (1972), p, 63.
3. E. I. Zakharov et al. , Ibid; , p. 77.
4. E. I. Zakharov et al. , Ibid. , p. 84. `
5. S. M. Karpacheva et al. , in: Extraction and Sorption in the Metallurgy of Molybdenum, Tungsten,
and Rhenium [in RussianJ, Izd. Tsvetmetinformatsiya, Moscow (1971), p, 182.
6. A. A. Komarovskii and G. F. Mironov, in: Ion-Exchange Technology [in Russian], Nauka, Moscow
(1965), p. 114.
7. Ya. M. Zagrai et al. , Ion-Exchange Decontamination of Industrial Sewage by Cation-Exchange Res-
ins in aQuasi-Liquid Layer [in Russian], Izd. UKRNIITI, Kiev (1966).
8. S. M. Karpacheva et al. , Pulsed Extractors [in Russian], Atomizdat, Moscow (1964).
9: R. Treibal, Liquid Extraction [in RussianJ, Khimiya, Moscow (1966).
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 Declassified and Approved For Release 2013/01/25 :CIA-RDP10-021968000400030001-7
CALCULATION AND PREDICTION OF RADIOACTIVE
CONTAMINATION OF THE LOWER ATMOSPHERE
BY ATOMIC POWER STATION STACK DISCHARGES
In the selection of an area for any reactor installation, one primarily takes into consideration the lev-
el of possible radiation hazard both for the population in general and for individuals living near the unit since
the radiation effects on the population must be as low as possible. This can be achieved mainly by high-
quality planning, construction, and operation of an atomic power station. However, no minor role can be
assigned to the location of ari atomic power station, in the selection of which one must take into account all
the features of meteorological dilution of the radioactive products discharged into the atmosphere during
operation of the atomic power station that are characteristic of the proposed area of construction.
Experience with operating atomic power stations shows that the fundamental criteria and physical
premises which are the basis for estimating atmospheric diffusion in the operating area of an atomic power
station are valid. Dispersion of a contaminantdischarged into the atmosphere occurs, as is well known, as
a result of its entrainment in turbulent atmospheric vortices of various scales with the intensity of the tur-
bulence being determined by the thermodynamic state of the surface layer of the atmosphere. The direction
of contaminant transport depends on the predominant wind direction. All cases of contaminant dispersion
can be classified for the several categories of thermodynamic stability responsible for contaminant trans-
port to the ground and the formation of a surface concentration field. For each class of atmospheric sta-
bility, definite values of maximum surface-layer contaminant concentration and definite localization dis-
tances from the source are observed and can be calculated.
At the present time, there are a large number of papers devoted to the evaluation of the degree of
atmospheric contamination by discharges from various types of sources. However, if one is speaking of
rapid methods for calculating atmospheric contamination suitable for practical use and having the necessary
nomograms and curves, two such groups of techniques should be mentioned.
1. Methods of Sutton and Pasquill. The latter was subsequently further developed by Mead, Beatty,
and Bryant, and was incorporated by them into the procedures of the United Kingdom Atomic Energy Author-
ity and was recommended to the World Meteorological Organization as a method for calculating the disper-
sion of discharges from atomic reactors.
2. Methods for calculating industrial contamination of the atmosphere developed in the USSR at the
Main Geophysical Observatory (MGO), at the Leningrad Hydrometeorological Institute (LHI), and at the In-
stitute of Experimental Meteorology (IEM) which have been checked by a large amount of experimental ma-
terial and which were used successfully for planning and forecasting.
These two groups of techniques are based on two different approaches to the description of atmos-
pheric diffusion: statistical and semiempirical using analogy with molecular diffusion (and in some cases
a combination of both approaches). Each has certain advantages and deficiencies. The difference between
them is mainly that the statistical theory is based on the study of fluctuations as a statistical ensemble and
considers their influence on the nature of the entire field including the average flow while the semiempiri-
cal theory assumes the average flow to be a steady state and studies its features and its influence on other
properties. The use of one or the other method for calculations of surface-layer contaminant concentration
Translated from Atomnaya Energiya, Vol. 36, No. 1, pp. 32-37, January, 1974. Original article
submitted April 11, 1973.
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No part o/ this publication. may be reproduced, stored in a retrieval system, or transmitted, in nay form or by any means,
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 Declassified and Approved For Release 2013/01/25 :CIA-RDP10-021968000400030001-7
510
290'
4 10_y
6.10-~
B f0:~
J'1Q 8
7.107 ~
~?i
G 10~'^~
G40~ ~ I /,
4 x, km
Fig, 1. Surface-layer concentration field for a weightless,
long-lived contaminant calculated by the Sutton (a), LHI (b),
IEM (c), and MGO (d) methods.
leads to identical results if one takes into account the averaging period for the set of meteorological param-
eters used and the accuracy of each method: The best agreement with experimental data, and conse-
quently the best accuracy, is obtained with calculations of contaminant dispersion under averaged condi-
tions, i. e. , for those weather categories which are characterized by conditions of ,weakly developed con-
vection and an equilibrium (neutral) thermodynamic state in the surface atmosphere,
As an illustration, Fig, 1 shows the surface-layer concentration field of a weightless long-lived con-
taminantcalculatedfor acontinuous discharge of unit intensity by the methods of Sutton [1], LHI [2], MGO
[3J, and Bosanke-Pearson-Denisov [4, 5] (the last is a variation of the IEM method, for a nonsettling con-
taminant), The discharge height is 110 m, the thermodynamic stratification is slightly unstable and close
to equilibrium, and the wind velocity at anemometer height, or the average velocity in the 0-110 m layer,
is .5 m/sec. Table 1 gives brief characteristics of the methods for calculating contamination of~the surface
layer of the atmosphere by the different techniques. The, error in concentration determination by the IEM
and LHI methods is approximately 50~,. The accuracy of the MGO method is 30~. The relative error was
not determined for the Sutton method,
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 Declassified and Approved For Release 2013/01/25 :CIA-RDP10-021968000400030001-7
Fig. 2. Change in axial surface con-
centration of a weightless, long-lived
contaminant.
The following notation is used in Table 1: x, y, and z
are the axes of a Cartesian coordinate system. The wind di-
rection is in the x direction. The source is located at the point
(0, 0, H); His the stack height, m; q is the contaminant con-
centration in the surface layer of the atmosphere, 1/m3; Cy
and Cz are the Sutton virtual coefficients for turbulent diffu-
sion, mn/Z; u is the wind velocity, m/sec; n is the Sutton
stratification parameter; yz is the contaminant dispersion in
the y direction, m2; by and bz are empirical constants; qM
is the maximum surface concentration; xM is the distance
from the base of the source to the point of maximum surface
concentration; m; uM is the critical wind velocity, m/sec; A
is a climatological parameter; m and F are respectively pa-
rameters which characterize the emission rate from the stack
of the gas-air mixture and its gravitational settling; V is the
volume of discharged gas-air mixture per unit time, m3/sec;
LET is the temperature difference between the discharged gases
and the ambient air, deg.
As is well known, the vertical temperature gradient in
the lower layer of the atmosphere and the wind characteris-
tics, 'which are responsible for the thermodynamic state of
the atmosphere, and the intensity of vertical and horizontal
transport of a contaminant are quite variable quantities and can
change their values even over comparatively short periods of
time. The LHI, IEM, and MGO methods take into account the
variation in wind direction over periods of 20-30 min and
therefore make it possible to estimate "single-discharge"
values of the concentration, i, e. , a concentration averaged
over 20-30 min. The Sutton method assumes a constant wind
direction and consequently makes it possible to calculate in-
stantaneous values of contaminant concentration, i, e. , con-
centration averaged over 2-3 min. In the main, this explains the difference in concentration values ob-
tained from Sutton's method and from single-discharge averaging methods. In addition, the Sutton method
can lead to significant errors in the calculation of surface concentration at large distances from the source
because of the lack of justification for the assumed scheme for spatial variation of the turbulent diffusion
coefficients.
The Pasquill-Mead method [6-8] offers an opportunity to determine comparatively simply contami-
nant concentration in the surface layer of the atmosphere for instantaneous, single discharge, and contin-
uous discharge, for various categories of atmospheric stability, and at sufficiently large distance from the
source by using the tables and curves of Pasquill for averaged values of the lateral and vertical expansion
of a jet. Gifford~s model of a fluctuating jet [9], which is extensively used in studies of the features of
horizontal dispersion of a jet from a continuously operating source, makes it possible to identify the operat-
ing time of a source of finite duration (instantaneous, single discharge) with the time of sample collection
or the time of averaging the concentration from a continuously operating source. Consequently, the Pas-
quill-Mead method can also be used for estimating instantaneous, single-discharge, and mean annual con-
centrations from a continuously operating source. The error in the determination of concentration by the
Pasquill-Mead method is approximately 100`x. Figure 2 shows the change in axial concentration for in-
stantaneous (curve 1), single-discharge (curve 2), and mean annual (curve 3) averaging calculated by. the
Pasquill-Mead method for continuous discharge of unit amount per second of a weightless, long-lived con-
taminant at a height of 110 m and for equilibrium stratification of the atmosphere and an averaged wind
speed of 5 m/sec. For comparison, the dashed curves show the change in axial concentration from a con-
tinuously operating source calculated by the Sutton method with instantaneous averaging (curve 4), the MGO
and LHI methods with single-discharge averaging (curves 5 and 6), and for mean annual averaging in the
case of a circular wind rose by the method of the Institute of Applied Geophysics (curve 7), which is de-
scribed in [10]. The Pasquill-Mead equations are:
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 Declassified and Approved For Release 2013/01/25 :CIA-RDP10-021968000400030001-7
TABLE 1. Computational Formulas and Parameters Usedin the Various Methods
3 ~pT VDT=280
xM=20H; uM=0,65 / ~i r, Sr,SZare taken from the
appropriate curves in [3]
*The value A=0.12 is recommended iri the MGO method for calculating the dispersion of a contaminant dis-
charged into the atmosphere from an elevated source under conditions of developed turbulent exchange in the
central zone of the European portion of the USSR There are no instructions in [3] for neutral stratification con-
ditions,
-2.303 k2
168 e _
4inst = (instantaneous source)
Olzxu
4sing =
-2.303 hs
168 e
(prolonged source; from 20-
30 min to several hours)
-2.303 hie
e
9 - 1.161tlzxu (continuous source)'
Here, His stack height, m; his the vertical expansion of the jet, m; 8 is the lateral expansion of the jet
in the case`of an instantaneous discharge, deg; xis distance from the source, m; Bp is the lateral expan-
sion of the jet in the case of a prolonged discharge, deg; u is the wind velocity, m/sec.
The parameters B, Bp, and h are taken for various distances from the source for weather categories
D and C-D in accordance with the recommendations of Pasquill and Bryant [6, 8].
Analysis of Figs. 1 and 2 shows that within the limits of accuracy indicated for the various methods,
the calculated results of surface-layer concentration of a contaminant discharged from the stack of a con-
tinuously operating installation agree rather well for identical averaging periods. Consequently; any of
the methods can be used for estimating possible or existing atmospheric contamination from a source dis-
charging radioactive products. One should remember, 'however, that in each specific calculation it is nec-
essary to be guided by the necessary accuracy in the determination of the concentration and by typical dis-
persion conditions for the given region.
The coefficients of turbulent diffusion obtained by Sutton and used in calculations of contaminant con-
centration by his method are only valid for dispersion under neutral conditions and only for 2-3 min averag-
ing. The Pasquill-Mead method is attractive because of its simplicity; however, determination of con-
centration by this method, as acknowledged by the authors, may give errors of several orders of magni-
tude under extreme stability conditions (strongly developed instability and moderate and strong stability).
The MGO method. is mainly used for calculation of contaminant dispersion under conditions of developed
convective exchange and for the determination of single-discharge (20-30 min) concentrations. In general
form, the LHI method (for various stability conditions) uses a number of parameters that are difficult to
measure and cumbersome calculations which require a certain amount of preparation. The IEM method is
mainly suited for the determination of the deposition density of contaminants that settle out.
Averaging
time
_ ll2 Az
-
2-3
i
Czs2-n
v z
2 e zx2-n e
Cy=0,21
CZ = 0,12
m
n
(instantaneous)
4 (x, y, 0) = ttCyCZUxz-n
n-0,25
54 e2yz a 14,5Ht,rs
20-30 min
- H y2
(single discharge)
e bZS C 26~sz
bu = 0,08
9 (x, y, 0) = v2zc b~bZUxz
bt =_0,024
4 (x, y. 0) = 9msi (x/xM) Sz (y/x) r (u/um)
A =0,12
20-30 min
AmF
m=1
(single discharge)
9m- HZ f VOT
F=1
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 Declassified and Approved For Release 2013/01/25 :CIA-RDP10-021968000400030001-7
Q,iv,
ta'
to-" tu' to_z to-' to? .o' mz tc' to? z,i~2
Fig, 3. Dependence of gl(Tl)/gz(TZ), the ratio of concentra-
tions averaged over different periods of time, on Tt/T2, the
ratio of the averaging periods,
There is much experimental data obtained in the Soviet Union and abroad [l, 5] which indicates that
the magnitudes of surface-layer contaminant concentrations averaged over different periods of time are
related to the length of the averaging period, This dependence is given by
where ql and qZ are the surface-layer concentrations of the contaminant; Ti and TZ are the corresponding
averaging periods; k is a coefficient of proportionality which takes into account the stability of the basic
parameters that determine diffusion of the contaminant, mainly the stability of wind direction; k = 1 for
short-period concentrations (averaging over a day or less).
The relation presented is rather universal and can be used, for example, for predictions of possible
mean annual contamination of the surface atmosphere around an operating plant from measurements over a
finite sample-collecting period. In this case, the magnitude of the coefficient k is determined by the elonga-
tion index n/no of the mean annual wind rose where no is the f requency of any direction of the eight-point
wind rose for equal probability of all directions and is 0.125; n is the actual frequency of the predominant
wind direction (in this case, k = (1/2)n/no). Figure 3 shows the relation given in graphic form for k = 1.
The relation given makes it possible to use the simplest and most effective methods for calculating
surface-Layer concentrations of contaminants in order to predict the magnitude of atmospheric contamina-
tionfor any period of source operation and for meteorological information collected over an arbitrary time.
Since radiation safety standards regulate the mean annual maximum permissible intake of radioac-
tive isotopes into the body and the mean annual dose of external radiation, they are a basis for estimating
the allowable atmospheric contamination produced during normal operation of an atomic power station in
accordance with the mean annual meteorological characteristics of the given region. Having information
covering a period of many years about the frequency of various types of atmospheric instability and the
mean annual wind in a specific region, one can determine, as shown in (10] and by other methods as well,
the probable mean annual atmospheric contamination and calculate the possible external irradiation of
people and the intake of radioactive isotopes into the body through the lungs,
A comparison of calculated mean annual surface Layer concentrations with the maximum permissible
mean annual concentration is a basis for establishing norms for the amount of power station discharges or
for establishing the parameters for the ventilation stacks. However, such calculations can be carried out
only in those cases where the premises underlying the computational scheme are valid (primarily, a smooth
and level topography in the area where the source of the discharge is located and homogeneity of the ver-
tical thermodynamic structure of the surface layer of the atmosphere), Topographical features (mountains,
river valleys), high frequency of unfavorable meteorological conditions observed during formation of ele-
vated inversions at discharge height, or a combination of these and other conditions (which exists rather
often) are extremely undesirable at the location of an atomic power station.
Thus, in considering plans for the location of an atomic power station, it is necessary to know the
following from the viewpoint of ensuring radiation safety of the population in the surrounding territory dur-
ing normal operation:
1. Data about the projected atomic power station -power and type of reactor, composition and
amount of products discharged, height, rate, and velocity of discharge, temperature of the discharged
gas -air mixture,
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 Declassified and Approved For Release 2013/01/25 :CIA-RDP10-021968000400030001-7
2. Climatological data -mean annual wind rose, chart of wind velocities with corresponding-mean
annual frequency, mean annual wind velocity (preferably for each compass point), mean annual frequency
of elevated inversions (at discharge height), mean annual frequency of calms and fog, their average and
maximum duration, mean annual daily and maximum air temperature during the warm season.
The frequency of elevated inversions at discharge height can be obtained from temperature and wind
soundings taken at points no more than 100-150 km from the location of the atomic power station. The
other data can be taken from observations made at weather stations in the standard network which are lo-
cated as close as possible to the proposed location of the atomic power station and no farther than 40 km if
the topography is smooth and level, If the locality is hilly and cut by river valleys, one should determine
the most representative weather station in order to avoid using data from those stations which are lo-
cated in areas with a specific microstructure of the wind regime (for example, those in river valleys), In
this case, the distance to the reference weather station should be no more than 20 km,
In considering the problems associated with the estimate of possible atmospheric contamination by
radioactive products during the operation of an atomic power station, one also needs a plot (sketch map)
of the region around the location of the atomic power station out to a radius of not less than 200 stack heights
(20-50 km) on which the topography of the region should be-shown along with the location and absolute datum
for the reference weather stations closest to populated points, residential areas around the power station,
and industrial areas. It should be noted that when it is necessary to locate an atomic power station in areas
with complex topography and wind and temperature regimes, a 4-5 year set of weather observations must
be planned and carried out ahead of time at the proposed point of construction with obligatory temperature
and wind soundings of the 200-300 meter layer of the atmosphere (height of the soundings is determined by
the height of the ventilation stacks),
1.
Meteorology and Atomic Energy, N. A, Byzova and K. P. Makhon'ko (editors), Gidrometeoizdat,
Moscow (1972).
2.
D. L. Laikhtman, F, A. Gisina, and S, N. Kaplan, Tr. Leningr. Gidrometeorolog, Inta. , No. 15,
Izd, LGU (1963), pp, 37-47. ,
3.
Instructions on the Calculation of Atmospheric Dispersion of Harmful Materials Contained in Indus-
trial Discharges (SN 369-67) [in Russian], Gidrometeoizdat, Leningrad (1967).
4.
A.
I, Denisov, Izv, Akad, Nauk SSSR, Ser. Geofiz. , No. 6, 834-837 (1957).
5.
N,
L. Byzova, Tr. Inta. Eksperim. Meteorolog No, 15, Gidrometeoizdat, Moscow (1970).
6.
F,
Pasquill, Meteorological Magazine, 90, No. 1063 (1961).
7.
P,
S. Mead, WM.O-No. 97, TP41 (1960).
8.
P.
Bryant, Rep, AH,SB(RP) 842, UKAEA (1964).
9.
F.
Gifford, in: Atmospheric Diffusion and Air Contamination, A, S. Morin (editor), izd, Inostr,
Lit, , Moscow (1962), pp, 143-165.
10. N. E, Artemova and E, N. Teverovskii, At, Energ, , 31, No. 6, 573 (1971).
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 Declassified and Approved For Release 2013/01/25 :CIA-RDP10-021968000400030001-7
B. A. Briskman and R. B. Novgorodtsev UDC 539.1.01/39.1.08
In titling the review in this manner, we have in mind the measurement of absorbed doses* in materi-
als irradiated in nuclear reactors, In addition, since such measurements involve an extremely extensive
group of problems which have been discussed in a number of reviews mainly by foreign authors, we shall
confine our considerations only to those problems associated with the determination of the contributions to
absorbed dose from the separate components in reactor radiation fluxes. Obviously, .these problems are
most complicated and are of great interest for in-reactor measurements. Their solution is necessary for
the following reasons. First, in studies of the interaction mechanism in radiation physics, chemistry,
biology, and material studies, the question often arises as to whether the effect produced by the same ab=
sorbed dose from different types of radiation will by the same. Second, data on absorbed doses are needed
for the very broad range of materials being irradiated in nuclear reactors. Direct measurement of them
for each chemical element in an actual reactor is unrealistic. As a rule, a method is used in which data
obtained for some standard material is converted to the dose in the desired- material. To do this, it is
necessary to know the dose composition in the standard material and the neutron and v-ray spectra. Third,
for correlation of irradiation results in different reactors, it is necessary to use certain irradiation param-
eters which can be the total absorbed dose or its neutron component.t
A large number of papers on in-reactor dosimetry has been published. The IAEA reports [1-3], the
papers devoted to calorimetry alone [4-8], and the monograph on chemical dosimetry [9] should be con-
sidered as the most complete reviews in this field. The problem is discussed in an extensive chapter in
the monograph [10]. However, with the possible exception of the very complete and interesting book [3],
relatively little consideration is given to separation dosimetry in all the papers mentioned. Since the de-
termination of the composition of absorbed dose does not depend in principle on what method is used (calor-
imetric, chemical, etc.) for this purpose, we shall also consider the problem of separation dosimetry in
its most general form.
Just how is dose composition determined? Breaking up the entire reactor neutron spectrum into fast
(E > 0.1 MeV) and thermal neutrons (as a rule, the contribution from neutrons of intermediate energy can
be neglected under the condition that it is taken into account in the fast-neutron contribution), we have
Dt=D2-I-Df ~-Dtih (1)
where Di is the absorbed dose in the i-th material. We shall assume acharged-particle contribution to the
absorbed dose accompanies radiative capture of thermal neutrons or the interaction of y rays with matter.
The quantity Dih is determined from the known thermal-neutron flux including consideration of radio-
active decay scheme, capture v radiation, v radiation accompanying a decay, (3-particle energy, etc. Cap-
ture v-ray spectra are given in [ll]. A scheme for the calculation of absorbed energy of capture y radiation
from surrounding materials is discussed rather thoroughly in [12]. Equations for the calculation of self-
absorption in large, cylindrical samples are given in [13]. Examples of the calculation of Dih including
*For specialists in reactor engineering, more familiarly energy deposition although -these quantities are
not equivalent, strictly speaking, and have different physical significance.
fiFor brevity, we shall call the collection of methods for solving the problem separation dosimetry for re-
actor radiations.
Translated from Atomnaya Energiya, Vol. 36, No. 1, pp. 39-49, January, 1974. Original article
submitted October 31, 1972; revision submitted March 20, 1973.
? 1974 Conswltants Bureau,, a. di. vision of Pleaam Publishiag Corporation, 227 IT'est 17th Street, New York, ;V. Y. 10011.
No part o/ this publication ma.y be reproduced, stored in a retrieaal system, or transmitted, in any form or by any means,
electronic, mechanical; photocopying, mi.crof~lming, recording or otherwise, wit/wont written permission of the pnblislaer. :1
copy o/ this article is available /rom the publisleer for $1;1.00.
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 Declassified and Approved For Release 2013/01/25 :CIA-RDP10-021968000400030001-7
self-shielding are given in [3], in [14, 15] for aluminum without consideration of self-shielding, and in [16]
for polyethylene. It is shown in [15] that the dose in Al for radiative capture of thermal neutrons results
from the absorption of the primary v ray (E = 7.7 MeV), the secondary v ray during a decay (E = 1.8 MeV),
and the /3-particle energy (E = 1.23 MeV) in the following ratio, 0.30 :0.09 :0.61. In this case, DAl = 5.48
? 10-13th Mrad/h. In a number of cases, the dose from the reaction H(n, ~D with Ev = 2.2 MeV is im-
portant [17, 18].
The calculation of Di in an irradiated sample from the (nth, y) reaction in the jacket around the sam-
ple is similar to the calculation of Dih [5, 19]. The quantity Di can be determined more easily if heavy
charged particles result from the interaction of thermal neutrons in reactions such as 10B(n, cx)~Li, sLi(n,
cx)3H, and 14N(n, p)14C. It is logical to refer here to the dose from fission fragments, which is of great im-
portance in chemonuclear processes. This dose is calculated on the basis of nth, the uranium fission
cross section, fragment energy, self-absorption in uranium, and the range of fission fragments in the ir-
radiated material [20-22]. Thus it is only necessary to know nth (thermal-neutron flux density) fora de-
termination of Dih from input data,
It is considerably more complicated to determine the first two components of the dose since the fast-
neutron spectrum is known inaccurately~as a rule and there i.s thus far no experimental data on the spec-
trum of in-reactor radiation (with the exception of [23] and information on the performance of spectral mea-
surements in [24]). .Apparently beginning with [25] (although the first separation of effects from various
kinds of y + ~ + a radiations by calorimetric methods was described in [26, 27]), greatest consideration
was given to the measurement of absorbed dose rate or the dose itself in?two materials in which the y ray
and fast-neutron interactions differed rather markedly (the quantity Dth was taken into account in advance).
We write down the following relations (here, the additivity of dose is taken into consideration):
D,=Di-~Di;
We introduce the following notation: (D1/DZ)v = Ky, (D1/DZ)n = Kn; D1/DZ = K, where K.y is the ratio of
the y-ray energy mass-absorption coefficients ?a/p (this statement is only valid for sufficiently thin sam-
ples where self-shielding can be neglected) and Kn is the ratio of. the neutron mass-absorption coefficients
?n/p (or the neutron scattering integrals). Then the relative fraction for the neutron component fn material
1 is calculated from
The most complete and reliable information on values of ?a/p for monoenergetic v radiation is given
in [28, 29]. There is selected data, for example, in [3, 30]. Values of ?a/p for v-ray sources of complex
spectral composition are given in [3, 31]. In Table 1, taken from [31] with some exceptions, average val-
ues of ?a/p (cmz/g) are given for various y-ray spectra.
Values of Kn for elements with Z ~ 8 and for chemical compounds such as CHZ, HZO, etc, can be
calculated from data in [32] for sources of monoenergetic neutrons. For Z > 8, there is similar data [33]
for a number of elements in a narrower energy range. Values of Kn relative to the dose in graphite are
given in [5, 14, 34, 35] for various types of reactors and other sources of complex spectra with the con-
tribution from inelastic neutron scattering to the absorbed dose taken into account in [35]. The scattering
integral ratios Di/DC for neutron spectra described in [36, 37] are given in Table 2 (DC is the dose in
graphite).
Theoretical and experimental neutron dose data for BE PO and DIDO reactor spectra normalized to
unit thermal neutron flux is given in [38] for an extensive set of elements.
Without analyzing .the accuracy and correctness of the concept behind -the published values for Kn and
Ky, we only point out that determination of absorbed dose composition using the method described above
has been accomplished many times. Ionization, chemical, .solid-state, and calorimetric detectors are ex-
tensively used for the purpose.
Ionization Methods. The application of these methods reduces to the us.e of homogeneous ionization
chambers (IC) -made of materials differing in hydrogen content [3, 10, 18].
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 Declassified and Approved For Release 2013/01/25 :CIA-RDP10-021968000400030001-7
TABLE 1. 'Y-Ray Energy Mass.Absorption Coefficients for Various Radiation Sources
average energy, MeV
Element
o,ss I I.05 I I,zo I I,4o I I.15 I 4,a I o,o~ I i,no
energy range, MeV
0,f-4,0 ~ 0,1-4,0 ~ 0,10-8,2 ~ 0,42-10,0 ~ 0,13-10,0 ~ 0,4-iU.O ~ 0,3-7,0 ~ 0, 1--10,0
HIFAR
AsH
T
RPT I
fission
ex
U I L I
I II
II2
p
0,0486
0,0483
0,04G8
0,0449
0,0411
0,0309
O,O;i04
0,04'!0
0,024(1
0,0245
0,0238
0,0231
0,0240
O,OlE9
0,0255
0,02/18
0.0274
0,0272
0,0264
0,025G
0,0'266
O,OiSB
0,0253
0,0375
0,0242
0,0239
0,0233
0,0229
0,023E
0,048^_
0,0248
0,0343
0.0235
0,0229
0,0232
0,02'18
0,0237
O,O181
0,0'?34
0,0233
0,0274
0,0255
0,0241
0,0235
0,0253
0,007
0,0251
0,0257
0,042E
0,0339
0,0293
0,0270
0,0363
0,0311
0.0288
0,0342
0,0900
0,0633
0,0500
0,0379
O,OE33
0,0390
0,051-1
O,OGE4
0,420
0,0879
0,0743
0,0524
0,0761
0,0413
O,U735
0,0915
H
G
H2O
Al
Ar
r~~
~f n
Pb
U
Note, HIFAR-enriched U-Al alloy, D20 moderator and coolant (U, L are the upper and lower limits of the in-
tensity of the low-energy portion of the spectrum); BSR-I, BSR-II, IRT-highly enriched fuel, H2O moderator and
coolant. In all these reactors, the data is given for the edge of the core.; RPT measurements were in the graphite
thermal column (there is a large contribution to the RPT spectrum from capturey-radiation in graphite and steel);
fission is the 2'"'U fission spectrum; exp is the empirical approximation (E) =exp (-1,11E),
Two ionization chambers were discussed in [10] for measurement of dose composition: a tissue-
equivalent chamber and a C-COZ (or Teflon-COz) chamber. In this case, a system of equations similar
to Eqs. (2) was used:
T - 1.047 _~_ ~'; (5)
C =1.04!' ~- KN,
where T is the ionization in the tissue-equivalent IC produced by I' (rad) of y radiation + N (rad) of neutron
radiation in units of the ionization produced by 1 R of y radiation; Cis the ratio of the ionization in the
C-COZ IC produced by they + n dose to the ionization produced by 1 R of v radiation; K is the ratio of the
ionization produced by 1 rad of neutrons in tissue to the ionization from 1 R of y radiation. In [18], the fol-
lowing relation,
k 1 .Dpi L 1 .Do=10-'I,
Ko W~ Lo Wr
was proposed for the determination of dose composition in a sample where K and Ko are the y--ray mass ab-
sorption coefficients in the chamber material and in the irradiated sample; E and Eo are the kerma in the
chamber material and in the samples when in identical neutron fluxes; We and Wr are the average ion-
formation energy for secondary electrons from y rays and for protons and recoil nuclei respectively, elec-
tron volts per ion pair; Do and Do are the corresponding dose rates in the sample, rad/sec; I is the cur-
rent in the chamber, A/g. Values of W in various gases (air, argon, CO2, CZH2, C2H4) for electrons, v
rays, and a particles are given in [39]. Values of W are given in [2] for a considerably larger set of ele-
ments. As a rule, it is assumed that W for protons and recoil nuclei agrees with the W for tx particles al-
though averaging (for example, over the fission spectrum) leads to a difference amounting to ~6`~. The
values of E/Eo are ratios of neutron scattering integrals, which have been mentioned above. Some are giv-
en in Table 2. In [18], these quantities are given for muscle and bone (sample materials) and also for CH2,
CH, C, and aerion (a conducting plastic used for the construction of IC in [40]) for three kinds of neutron
spectrum. Such methods for the determination of dose composition were used in [41-43]. For the deter-
mination of fast-neutron kerma in biological samples irradiated in horizontal channels of the IRT reactor,
the same authors used IC made of polyethylene and polystyrene [44].
In Eq. (6), the value of W for y rays and recoil protons in a given material differs little (a few per-
cent at most). At the same time, Boyd showed [3, p. 193] that oxygen and carbon ions formed during ir-
radiation of a C-COz IC have energies of 1-1000 keV and the value of W for these ions is twice the value
of W for electrons (from measurements with the chamber mentioned and with a graphite calorimeter). The
sensitivity of such an IC is 0.1 ?A/mW for y radiation and 2.5 ?A per 1012 n-cm-2-sec-1 for neutrons with
energies above the nickel threshold. This difference makes it possible to use an IC and a calorimeter made
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TABLE 2. Di/DC for the Neutron Component of Dose
1/E spectrum over
0-1 MeV
87,9
16,4
0,75
0,42
0-2 MeV
.81,1
19,4
0,84
0,38
0-3 MeV
77,0
20,9
0,75
0,38
Homogeneous moderator
C
71,2
20,7
0,78
0,36
DZO
67.3
20,5
0,75
0,36
HRO
60,3
21,4
0,72
0,34
Graphite lattice
70,1
22,0
0,77
0,37
DIDO reactor; MkIII fuel elements
67,9
22,7
0, 73
0,38
Fission spectrum
57,7
21,9
0,71
0,33
of the same material [45] instead of IC made of different materials, Since such a conclusion is not in ac-
cord with the premises of [41-43J, the question of the variation of W in CO2 remains unsolved for mixed
radiation; this is also noted in [18].
Liquid IC appear promising for separation dosimetry. They consist of two electrodes separated by
a few millimeters and are filled with a dielectric fluid of high purity, Four dielectrics were used in [46-
48] for this purpose: pentane, hexane, heptane, and isooctane (the last proved to be the best). An elec-
. trochemical dosimeter [49] can also be used for this purpose, employing an aqueous solution of HzSO4 in
which hydrogen peroxide and pure hydrogen are formed through the action of radiation,
The chambers mentioned can also be used for the determination of the thermal neutron contribution
to the absorbed dose, Thus an IC with copper electrodes and filled with boron trifluoride was used in [50].
With the IC belong the Hurst proportional counters [51-53] in which the pulses from v-ray secondary elec-
trons have a considerably smaller amplitude than those from recoil protons, making it possible to separate
the dose from fast neutrons by discrimination against pulses from y radiation, Simultaneous recording of
D'Y and Dn is achieved by the modification of the Hurst counter proposed in [54].
It should be noted that in most cases the use of IC in nuclear reactors has an upper limit on dose rate
because of temperature effects,
Chemical Methods. These methods were reviewed in 1963 [9]. Papers published after 1963 are
mainly discussed below.
The use of chemical methods in separation dosimetry is based on the fact that in a number of mate-
rials bhe change occurring because of the effects of ionizing radiation does not depend, or is slightly de-
pendent, on LET, Such materials can. be used for the determination of the total absorbed dose from mixed
radiations over a broad spectrum. At the same time, a strong dependence of radiochemical effects on LET
appears in many materials. A combination of materials in the first and second groups provides a .solution
for the problem formulated. In this case, the relation used for the determination of the radiochemical
total yield is
Gef-G D"lD-i-GvDv/D,
where G.y and Gn are the radiochemical yields for v rays and fast neutrons respectively. The fundamental
difference between Eq. (7) and Eqs. (2)-(4) is in the replacement of the coefficients K3, and Kn by the ratio
of the corresponding radiochemical yields.
In the first group of chemical dosimeters is a cerium sulfate solution, the G of which for y rays and
fast neutrons is 2.32 [55., 56]. Its basic deficiency is temperature dependence of the response (for tem-
peratures above 35?C). The upper limit of the response is 100 Mrad. In the 0.1-3000 Mrad' range, ni-
trous ox-ide, N20, has been used repeatedly [2, 57-59]. In;its dissociation, NZ and OZ are formed, the
amount of which is a measure of the absorbed dose (GN = 10.0 f 0.2, Gp .= 4.0 f 0.4). For the proper
interpretation of the response of this dosimeter, one sh2ould take into account the contribution from the
laN(n, p)14C reaction, which amounts to 50`~ of the total dose in a number of cases ,[60].
Cyclohexane can be used for doses up to 16 Mrad. Hydrogen is obtained from its radiolytic dissocia-
tion (GH = 5.2 [61]). For the determination of dose composition, deuterated Cyclohexane, CsD12, is used
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simultaneously with CsHiZ [3]. One can obtain only a crude estimate of the dose components with this meth-
Aqueous solutions of glucose and maltose [62-64], for which Gy/Gn = 1.15-1.10, are used for doses
up to approximately 100 Mrad. At high dose rates (more than 104 rad/sec), however, it is recommended
[65] that one use dry glucose and analyze the gaseous products of radiolysis by gas chromatography. Aque-
ous solutions of oxalic acid are also used for the same purpose [66, 67]. Determination of dose composi-
tion is performed by means of oxalic acid solutions in HZO and DzO [66]. In this case, Gn = 3.3 was ob-
tained for protons and Gn = 2.8 for deuterons. These same solutions together with calorimeters were used
for carrying out comparative dosimetric measurements at the ISIS reactor [68].
For fluorocarbons, which are used in reactors in the 3-50 Mrad dose range, the values of Gn and Gy
agree within 10-15?,b [69]. Dyed polymer films, the optical properties of which change under irradiation
[70, 71], belong among the chemical systems having a yield slightly dependent on LET.
Above all, the ferrosulfate dosimeter (FSD), for which Gy/Gn = 2.0, ought to be placed in the second
group of chemical dosimeters. Although the FSD has not found application in reactors because of the low up-
per limit on measured dose (4 ? 104 rad), various modifications of it (the introduction of copper ions [72],
increasing the upper limit to 10 Mrad) make it possible to use this system in separation dosimetry. The
limitation on the use of the FSD for determination of the v and neutron components of the dose is also as-
sociated with a known uncertainty in the quantity Gn, particularly for recoil protons with E < 150 keV. A
theoretical calculation of Gnf for recoil protons was made in [73] on the basis of the concept of a local ra-
diochemical yield determined at each point of the proton range. Values of Gnf equal to 6.4 and 6.9 respec-
tively were obtained for two horizontal channels of the BR-5 reactor.
Chlorinated hydrocarbons used for this purpose (in particular, tetrachtormethane [74J for doses up to
approximately 2 Mrad and tetrachlorethylene [75] up to 0.2 Mrad) are practically insensitive to neutrons.
A combination of a deaerated FSD and a bichromate dosimeter was proposed [76] for doses up to ^-5 Mrad.
The use of plastic [77] and liquid [78] scintillators is based on principles similar to those given above.
In the second paper, the reduction of radioluminescence in liquid activators (n-terphenyl and 2, 5-diphenyl-
oxazole) was used in an experiment designed for the determination of dose composition. The radiolumines-
cence degradation effect was used with polycrystalline scintillator films [79].
In most solid detectors, dose composition is determined through a difference in hydrogen content.
Thus the use of polymer films of polyethylene and cellulose diacetate [80] and also of polytrifluorochlor-
ethylene [81] proved to be technically feasible. Such an approach also occurred in [82] where the gaseous
system CO2, CZH4 (or CzHz), and NZO was proposed. Free stable radicals formed by the irradiation of
amine salts of organic acids were detected by means of EPR in [83]. They contain up to 14~ of hydrogen
making it possible to separate out the neutron component of the dose. Because a number of solid-state de-
tectors have a response with a marked dependence on LET, such detectors are often combined with chem-
ical detectors. Thus a combination of liquid chemical dosimeters (FSD and cerium solutions) and thermo-
luminescent dosimeters (aluminophosphate glass, manganese activated) was used in [84], and in [85],a com-
bination of chemical dosimeters (oxalic acid and glucose) and solid-state dosimeters (the same glass, but
measurements in the megarad region were made through the change in optical density). At a lower dose
range, SGD-8 glass can be used [86]. In some cases, the combination of a chemical dosimeter and a cal-
orimeter is feasible [32].
As with IC, it is possible to pick out the thermal neutron contribution to the absorbed dose in a num-
ber of cases by means of appropriate additions to the chemical system. Thus, a glucose dosimeter with
1oB introduced in the form of boric acid was used in [87].
Solid-State Methods. In most cases, such detectors are used in the low-dose region of the biological
range. Nevertheless, present progress in the construction of microdetectors leads one to hope for con-
siderable increase in the limit on the dose measured with them. Their use in reactor dosimetry is de-
scribed most completely in [3, 88]. As a rule, these detectors are insensitive to fast neutrons and are
used for measurement of the y component of the dose or the thermal neutron contribution (for example,
through the formation of tracks in mice which are fixed after irradiation by etching [89]). The manganese-
activated phosphate glasses mentioned above [90] make it possible to measure D'Y up to 20-30 Mrad. In
the thermoluminescence region, it is convenient to use LiF. A TLD-700 (almost pure 7Li) and TLD-600
(highly enriched in sLi) pair provides a determination of Dy and Dth [91]. The sensitivity of TLD to fast
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neutrons when using a TLD in combination with atissue-equivalent IC for separation dosimetry was inves-
tigated [92] and the variation in y sensitivity of a TLD-700 during neutron irradiation has also been studied
[93]. No change was detected up to 1600 rad, An increase in TLD sensitivity to fast neutrons is achieved
by dispersion of the phosphor in hydrogenous materials, The semiconductor dosimeters are practically
solid-state IC. Their sensitivity to fast and thermal neutrons is provided by the deposition of a thin layer
of hydrogenous material on a silicon diode or by a coating of boron or uranium respectively [94]. In the
first case, the conduction current arises because of recoil protons, and in the second case because of cx
particles or fission fragments, .
A scintillation spectrometer with a stilbene crystal has been used for separation dosimetry [95). In
this case, the well-known method. of pulse-shape discrimination was used,
Calorimetric Methods, 'The most widely used method for separation dosimetry is the calorimetric
method [5, 14, 15, 34, 38, 66, 96-114]. The construction of the calorimeters is most diverse; adiabatic,
quasi-adiabatic, "kinetic," isothermal, "pedastal,"with temperature differential in a gas gap, in a solid,
etc. A review of such calorimeters and? methods is given in [3-8]. Despite the diversity in technique, the
principle of separation dosimetry remained the same; measurement of absorbed dose rate in different ma-
terials with subsequent application of Eq, (3), In this work, as a rule, the absorbers used in the calori-
meters were hydrogenous materials (CHZ, CH, HZO, etc,) in combination with heavy materials or with ele-
ments making up the chemical compound (C, COZ, DZO, Teflon, etc,) or an element and its hydride (Zr
and ZrHn),
In all the methods described above for the determination of the components of absorbed dose, two de-
tectors were most often used. To increase accuracy, a set of materials which made it possible to "over-
determine" an equation system such as Eqs, (2) was used in a number of cases (for example, in [34, 66,
99]). The use of least squares was recommended for its solution in [3], the proposed approach being a
particular case of the more general method for solution of overdetermined systems of linear equations
[115 ] .
As far as we know, none of the authors evaluated the accuracy of the method used with the exception
of [5]. This also explains why the published data is often quite contradictory, particularly the values for
Gn.
One of the authors of this review showed [116] the mean-square error in the determination of the rel-
ative magnitude of the neutron component m of the dose is given by
z I~? nc-1 ~ ~ Irv ~ ~ ~ k:? ~' z
Kn-liy na ~ ~ ~ lin-Ky ~ rn(K?-K~,)
We shall analyze the error ~ for two pairs of materials usually used in separation dosimetry of re-
actor radiations: CHZ-C and ZrHn-Zr (n = 1.42 [15]) in a light-water reactor, We discuss two versions;
1) the fraction of the neutron component in the absorbed dose for hydrogen is mH = 0.82 (usual for experi-
mental channels located in the reactor core); 2) mH = 0.5 (for channels in a reflector), For these mate-
rials Knl = 8.80, K.yl = 1.14 [34], Kri2 = 22.8 [15), and Kv = 1.013 (E = 0.6 MeV), In both cases, the es-
timate of the errors Egv and sgn are fl'~ and- f5`~ respec~lively, The results are given in Table 3. The
last two columns of the table give the error in measurement when using the "outstripping" effect about
which more will be said later, The value of K is obtained from
.fi'= DrtxalDB = 1 -~ Px 1 1
where pI.I is the weight fraction of hydrogen in the hydride of the desired element B; Ky = (DB/DM)y;
Kn = (DB/DH)n; EZmI, em2, and Ems are respectively the first, second, and third terms on the right side
of Eq, (g); m = D~H/DBH, Values of ~ and ~ were calculated for two values of the error sD in the
measurement of absorbed dose rate equal ~0 3,~ (in the numerator of the fraction) and 5?b (in the. denomina-
tor of the fraction), As is clear from Table 3, a completely acceptable value of 5-8~ for em is obtained
for irradiation in the reactor core only for the CHZ-C pair, For the same pair in the reflector the error
reaches 25-40~, and. for the ZrHn-Zr pair th'e error is 35-56`~ in the core and 180-300 in the reflector.
The values of em rise sharply for mH < 0.5. It should be pointed out that the estimates of eD amounting to
3 and 5~ are rather optimistic if only because they are related to measurements of energy deposition and
not of absorbed dose, The violation of electron equilibrium in the calorimeters used makes a realistic
evaluation of the error in the measurement of absorbed dose rate extremely difficult,
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TABLE 3. Error in Measurement of Neutron Dose Component
Material ~
Tfn
mH ~ Tc -x
n v
r
m~
emixio9
e;,2xiu4
ejn3xlo~ ~
e,nxio~
e;n3x1o4
e,,ixsa~
5
2
15
545
0
0
9
55
0
22,~i
4,9
25
1
1
G5
CII~ - C
1,1
,
,
,
,
g0,0
7,S
,
,
(p?=0,143)
O,S2
%rfI? - 'Lr
1,05
1,143
0,119
GO
O,GO
4130
3000
34,5
6G ~~
G3
91,0
(Pn == 0.0154)
C[?. - C
1,15
1,34
0,170
31,5
0,55
ICG7
.~
41,5
34,2
S,1
0,50
%rI~?
-Zr
1,05
1,04
0,027
1430
O,OG
30 200
178
,~
1730
SIi,O
?
81150U
JO
Analysis of the structure of the error in the determination of the quantity m shows that the error Egn
can be neglected when using a pair of materials which are markedly different in their interactions with
neutrons and v rays (Kn ? K.~. The main contribution to >rm is made by the error sg, i, e. , the error in
determination of the dose rates in the two materials. The quantity Egv is rather small but not always. In
fact, the CHz-C pair received widespread acceptance particularly because the ratio of the mass absorp-
tion coefficients ?a/p remains practically constant over a broad range of y-ray energies, and consequently
the error Egv can be neglected because of the absence of data for the v-ray spectrum. Sometimes a pair
consisting of polyethylene and a heavy metal (Bi [105], Pb [106]) is used. The use of such pairs, which at
first glance seem very suitable, leads to quite large values for trK Since Kn ? Ky in these cases, .Eq.
v
(8) takes the form
gym' ~ i~t (?~li V 1- &If)
Since the v-ray spectrum is poorly known, the quantity EK may be ~100~. In fact, in the range 0.3 < Ey
< 2 MeV, we have 0.93 < (?a/P)Pb/(?a/P)CHz < 7.74 and. 0.94 < (?a/p)Bi/(?a/p)CH < 8.01; one can note for
comparison that in the same ti-ray energy range 1.139 < (?a/p)CH2/(?a/p)C < 1.41. Then, for example,
we obtain sm = 300`6 with mCHz = 0.25 when using the pairs CHz-Pb or CHz-Bi.
A similar situation holds in the determination of the absorbed dose rate DH in hydrogen proposed by
some investigators [117] for correlation of the results of neutron irradiation of metals. In fact, EDH is
determined from
-_ KZ-I-(1-Px)z .sD -~'( K-1 1282 ~ (11)
and the quantity DH itself is determined from consideration of additivity with respect to the data for dose
rate in the element B and in its hydride BH:
_ K-1~-Px
Dx DB Px .
Here, as before, K = DBH/DB. It is assumed aDB - sDBH. Analysis of Eq. (11) shows how important
the correct choice of material pairs is for the determination of Dg. Numerical calculations for the two
pairs CHz-C (1) and ZrHI,88-Zr (2) for the two values 0.82 and 0.5 for mH yield
(EDH)2I(EDH)1 = 4.20 ? ?DZr~EDC' mH = 0.82;
(EDH)2I(EDH)i = 6-32 ? EDZr~EDC' mH = 0.50
(it is assumed epg = 0; the conversion from xnH to K is in accordance with Eq. (9); the subscripts 1 and
2 correspond to the element pairs), i, e. , for ~DZr - EDC the measurement error EDg when using the sec-
ond pair of materials is four to six times greater than for the first pair. When EPH is taken into consider-
ation, this difference becomes even greater,
The values given above are related not only to calorimetry but also to any dosimetric method or sys-
tem which is used for solution of the problem being discussed with the sole difference that other methods
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give poorer results in comparison with calorimetry, The consequences of such errors in application to
radiation processes using mixed radiations can be quite serious [116J.
One can therefore reach the following conclusions: 1) the error in determination of absorbed dose
composition is quite large; 2) the magnitude of these errors is mainly determined by errors in measure-
ment of dose rate or of the dose itself; 3) the existence of such errors is a chronic defect of the separa-
tion method for determination of dose composition; 4) since the error in the measurement of dose rate by
chemical methods is considerably greater than for calorimetric methods as a rule, we consider the use of
chemical methods for separation dosimetry of reactor radiations to be inadvisable, In technical reactor
dosimetry, their area of application may be the measurement of total doses (when Gn ~ G.y) and in mea-
surements of exposure doses in a number of cases (when Gn ? Gy), Liquid or gas chemical dosimetric
systems are irreplaceable in performing loop experiments [118].
Where should one look for an escape from the situation that has developed? If one considers the com-
plexity of in-reactor measurements and the effects of some factors which often cannot be taken into account,
one must not count on a significant increase in the accuracy of absorbed dose measurements, We consider
most realistic a fundamental rejection of direct measurement of dose rate preceding the determination of
composition, We have developed two methods for, simultaneous determination of absorbed dose rate and
composition for reactor radiations in which (and this is the point) the determination of the relative contri-
bution of the radiation components to the dose is not associated with a previous measurement of total dose
rate in different materials,
Relaxometer Method [119-121]. To determine dose composition one can use not only the difference
in energy mass absorption coefficients for a given form of radiation in two materials but also the differ-
ence in the linear absorption coefficients for two forms of radiation in a single material, This idea is also
the basis for the relaxometer method, During irradiation of a detector of given geometry, the temperature
field in it depends on the distribution of internal heat sources, i, e, , on the nature of the radiation flux at-
tenuation which, in turn, is a function of the absorbed dose composition, Thus the problem reduces to the
establishment of this last dependence and a temperature measurement at given points in the detector, The
error of the relaxometer method is considerably below the error of existing methods for mcH < 0.5, i, e, ,
for reactor channels in a reflector, and is 8-10`~ over practically the entire range of mCH2 va~ues,
"Outstripping" Method [121, 122], Unfortunately, the relaxometer method can be used only in sym-
metric radiation fields and the time required to make the measurement is.quite prolonged (approximately
3-4 h), This gave us the incentive to return to the use of the difference between ?a/p in two materials but
at a qualitatively new level, An attempt was made to .measure in in-reactor experiments not the dose rates
but their ratio K (to replace absolute measurements by relative ones), Furthermore, one can immediately
determine m and then obtain from an independently measured value of the dose rate the absolute value of
Dn. The experimentally determined dependence of the time to intersection of the temperature curves of
two calorimeter elements on the ratio of the energy deposition in them was used, The method was realized
in the design of an adiabatic isothermal calorimeter [123]. In the course of a single experiment, the ratio
K and the total- absorbed dose rate are measured, The systematic error in the measurement of K does not
exceed 0.6-0.9~ and the convergence is no worse than 1`~. Use of the "outstripping" effect made it possible
to reduce 8m by factors of 3.3-5.3 in comparison with existing methods (see Table 3) and to reduce mea-
surement time markedly,
Obviously, the use of the outstripping method significantiy facilitates the determination of absorbed
dose composition,
In conclusion, we turn to the choice of material for the determination of composition and dose rate,
As follows from. the examples given above, the use of a pair of materials - a chemical element plus its
hydride with maximum possible hydrogen content - is best, Taking this into consideration, the poly-
ethylene-graphite pair is preferable, The Zr-ZrHn pair, of course, is without competition for investi-
gations at high temperatures but is markedly inferior to the CHZ-C pair with regard to accuracy, At
times the supplementary use of polystyrene in combination with C or CHZ has been proposed [34, 104].
Polyethylene is inferior to polystyrene with respect to radiation stability, In those cases where an in-
vestigator is interested only in the total absorbed dose in a given material, its direct measurement by any
calorimetric method is undoubtedly the most precise, However, since knowledge of the dose in an exten-
sive set of materials is required in point of fact and the replacement of absorbers in irradiated calorime-
ters is extremely difficult (particularly because of their activation) (even in heat deflectors [124]), mea-
surement of the dose Ds in some standard material is most convenient, Conversion from Ds to Di in a
given material is accomplished by means of the expression
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Dt=Ds Kv-f-DS Kn.
In most cases, polyethylene can be such a standard material, This conclusion is based on the following
considerations.
1. As shown in a number of papers [5, 35], the quantity Kn = (Di/DCHz)n in the range 6 < Z < 82 and
over a broad set of neutron spectra (from a 1/E spectrum to afast-reactor spectrum) is independent of the
type of spectrum within 6-10~ and with ~n affecting am very slightly.
2. The quantity Ky = (Di/DCH~ v for Zef < 20 depends little on the effective energy of a reactor v-
ray spectrum (see examples for CHZ and C), For Z = 20, Ky varies by f2~ over the range 0.4 < Eav < 2,0
3. As shown above, the maximum accuracy in the determination of dose composition is obtained
when using the CH2-C pair.
4. Limitation of the accuracy in Ky to the value Z = 20 does not play an important part in many cases
since the magnitude of the total absorbed dose is mainly required in radiochemical studies where the radia-
tion yields for v rays and neutrons are approximately the same in most cases. In radiochemistry prob-
lems, low-Z elements are mainly considered, In radiation physics problems where elements with medium
and high Z are used only the neutron component of the dose is required, obviously, since y irradiation in
a background of neutron radiation does not yield significant radiation transformation, In this case, the
quantity K.y is not needed,
5. Graphite is aradiation-resistant material for use as a standard material in the measurement of
y-ray absorbed dose [125]. The change in its thermophysical properties under neutron irradiation is quite
small, which is important for calorimetry, and it has been thoroughly studied [126]. Polyethylene is con-
siderably less radiation resistant; however, its mechanical strength increases under moderate irradia-
tion, the working temperature increases to ^-200?C, and the change in thermophysical characteristics as a
function of absorbed dose has been studied rather thoroughly [127, 128]. We have shown previously that
the energy stored in polyethylene during irradiation does not exceed 0.7`~ [129]. Unfortunately, the Low
thermal conductivity of polyethylene increase the calorimeter time constant,
From our estimates, the error in the determination of total absorbed dose in elements with Z < 20
by conversion of data for absorbed dose rate and composition in polyethylene obtained by the outstripping
method is 5-15`16 in the range 0.85 > mg > 0.40 and 3 < sD < 5~. Under the same conditions, EDn is 7-11~,
which satisfies present requirements.
At the present time, work is being done on standardization of y--ray absorbed dose measurements
[86] and neutron flux measurements in reactors [130]. The time has come for standardization of reactor-
radiation dosimetry also. We hope that this review will prove to be useful in the performance of such
work.
The authors thank V, I, Ivanov, Yu. I. Bregadze, and V, V, Generalova for discussions and valuable
advice,
1. In-Pile Dosimetry, Techn, Rep. Ser, , No. 46, IAEA, Vienna (1965).
2, Neutron Fluence Measurements, Tech, Rep. Ser,, No. 107, IAEA, Vienna (1970).
3. Determination of Absorbed Doses in Reactors, Techn. Rep, Ser. , No. 127, IAEA, Vienna (1971).
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 Declassified and Approved For Release 2013/01/25 :CIA-RDP10-021968000400030001-7
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 Declassified and Approved For Release 2013/01/25 :CIA-RDP10-021968000400030001-7
63. S. V. Starodubtsev, V. V. Generalova, and G. V. Polyak, in: Radiation Physics [in Russian], Izd.
AN Latvian SSR, Riga (1964), No, 11, p. 27.
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65. E, P. Kovaleva, E, P. Petryaev, and E, P. Kalyazin, in: Dosimetry and Radiation Processes in
~~~ Dosimetric Systems [in Russian], FAN, Tashkent (1972), p. i67.
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Dosimetry [Russian translation], Gosatomizdat, Moscow (1962).
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V. Tkachenko, in: Radiation Dosimetry and Spectrometry of Ionizing Radiation [in Russian], FAN,
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 Declassified and Approved For Release 2013/01/25 :CIA-RDP10-021968000400030001-7
109. L, Bod and B. Radak, Bull, Boris Kidric Institute, 20, Chemistry, No. 2 (1969):
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(1968), p. 229. "
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5, 1180 (1972).
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130. R. D, Vasil~ev, Metrology of Neutron Radiation in Reactors and Accelerators. Abstracts [in Rus-
sian], VNIIFTRI, Moscow (1971), p, 16.
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 Declassified and Approved For Release 2013/01/25 :CIA-RDP10-021968000400030001-7
APPLICATION OF THE BUBNOV - GALERKIN METHOD
TO A MULTIGROUP CALCULATION OF A
TWO-DIMENSIONAL REACTOR
I. P. Kukharenok UDC 621.039.51.12
Using an example of a solution of an integrodifferential equation with three variables:
M (u, r, z) tp (u, r z) -{- c~N (u, r, z) ~ (u, r, z) = 0
we give a new method for obtaining calculation formulas of variational methods, on the basis of which cer-
tain concepts of tensor analysis are assumed. It is possible that this method has an analogy with the Dirac
method [1] and reduces to the following: at first, in simple spaces, to find reference operators; then,
using tensor transformations, to construct an operator that represents the reactor in the complex space of
interest to us. Evidently, the principal advantages of such a method are the uniformity of the procedures
and the quick operation. with a given number of trial functions.
The method is realized in a program for calculating a cylindrical reactor [2]. The spatial-energy .
distribution of neutrons is sought in the form of a triple series:
~ (u, re z)=tlRl/i~ (u) j~ (r) /Z (z)+
where the I-ij(u) are local functions (this is equivalent to using a group method for describing the energy de-
pendence). The functions fi(x) = fi(r), fl(z) can be as follows: a) given in the form of tables, i, e. , prac-
tically arbitrary; b) polynomials of the form fi(x) = b~igt(x), where the functions gi(x) are given in the
form of standard tables, and the coefficients bpi can be arbitrary; c) the coefficients bpi are eigenvectors
of the matrix that represents a multiregion one-dimensional one-group reactor-model in the space gi(x);
the diffusion coefficients, the absorption cross section of the thickness of the band of the model can be ar-
bitrary.
The principal constraints on the program are the following: no more than 26 groups, up to 25 trial
functions in the given group, and up to 40 bands. Presently the program is included in the complex of [3],
according to which macro- and micro-cross sections are calculated taking account of the blocking of reso-
nances, the number of processes, and the breeding ratios; we derive the reactor criticality by variation
of concentrations. This complex is in the code of a BESM-4 and is formulated just as the standard pro-
gram of an IS-2 system. The volume of the complex is 8500 words.
The calculation time of the program [2] for 26 groups, 5 x 3 trial functions, 10 zones is ~6 min.
Examples are given (the f~ are Bessel functions, and the fl are cosinusoidal functions) showing that for
such a number of trial functions, a calculation accuracy of the neutron fields of multizone profiled fast
reactors that is sufficient for practice is attained.
1. Ya. A. Skhouten, Tensor Analysis for Physicists [in Russian], Nauka, Moscow (1965), pp. 333-372.
2. I. P. Kukharenok, Preprint of NIIAR, P-103 (1971).
3. I. P. Kukharenok, Preprint of NIIAR, P-164 (1972).
Translated from Atomnaya Energiya, Vol. 36, No. 1, pp. 51-52, January, 1974. Original article
submitted January 15, 1973.
? 1974 Consultants E3ureau, a division o/ Plenum Publishin~.Corporation, 227 Rest 17th Street, New )'ork, IV. Y. IOOI1.
No part o/ this publication may be reproduced, stored in a retrievhl system, or transmitted, in any form or by any means,
electronic, mechanical, plaotocopyi.ng, microfilmi.ag, recording or otherwise, without written permission. o f t/te publisher. A
copy of this article is available from t/te pu.blisher for ~I;i.00.
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 Declassified and Approved For Release 2013/01/25 :CIA-RDP10-021968000400030001-7
MEASUREMENT OF THE ABSOLUTE INTENSITY OF
THE 278 keV LINE OF Np23s
L.
N.
Yurova, A. V. Bushuev,
V.
I.
Petrov, A. G. Inikhov,
V.
N.
Ozerkov, and V. V. Chachin
Using a Ge(Li) spectrometer, we have measured the absolute intensity of the 278 keV line of Np239
The line was observed in the spectrum of a uranium sample irradiated in the thermal column of the F-1
reactor at the I. V. Kurchatov Institute of Atomic Energy.
In order to determine the absolute intensity, we used the expression
2~a Ao II// _ 1 1I
Y2aoNp- N$aCtII ngnu~(1-0-RNP~u)@-~"NptOeK \1 NCd / ~
where Ao is the total number of pulses in the peak; N$ is the number of U238.nuclei in the sample; ?ctn is
the cross section for radiative capture of thermal neutrons in U238 (octn - 2.69 f 0.03); e is the efficiency
of the Ge(Li) spectrometer for Ev =.278 keV, which was determined by using a Hg203 source (No. 049 in
the IAEA collection); RCd is the cadmium ratio for the foils used, which was found experimentally to be
70 f 3; nvo is the neutron flux at the sample during the period of irradiation, estimated using gold foils
20 ? thick which were loaded together with the uranium samples.
The coefficient K takes into account the self-absorption of the 278 keV gamma radiation within the
uranium sample; the value of K was determined experimentally from measurements with a collection of
samples of various thicknesses. The activities of the gold foils were measured with a calibrated scintilla-
tion spectrometer.
An EVM M-220 computer was used to analyze the spectra obtained. We list below the values of the
absolute intensities (for comparison, the results of [l, 2] are listed as well):
Absolute Intensity
Results of [1]
0.141 f 0.007
Results of [2]'
0.145 f 0.004
Present results
It is clear that the results are identical within the lim
0.141 f 0.004
its of error.
1.
G. Ewan et al. , Phys. Rev. ,
L
108,
ITERATURE
1308 (1957).
CITED
2.
G. Ewan and M. Wahlgram,
Nucl.
Instr, and Meth. ,
99, 337 (1972).
Original article submitted March 20, 1973.
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 Declassified and Approved For Release 2013/01/25 :CIA-RDP10-021968000400030001-7
PARAMETERS OF THE RADIATION FIELD NEAR AN
APPARATUS USED FOR AGRICULTURAL IRRADIATION
The "Gamma-field" radiation engineering apparatus is used for- agricultural irradiation under natural
conditions in a vegetation process in order to obtain source material for selection, and -also to solve a num-
ber of problems connected with plant-growing. The apparatus was placed in a field. An earth embankment
was built at a radius Rz = 30 m from the apparatus, and a protective zone terminated at a radius R1 = 200
m. A 1660 Ci Cosy source was placed in a container at a height of 3.5 m from ground level. The radius
of the irradiation zone R3 ~ 25 m. We studied the regularities of the formation of the radiation field in the
irradiation zone as well as outside its boundaries, and developed the engineering technique for calculating
the optimal parameters for such types of apparatus. It should be noted that few studies of the radiation
field of such a geometry have been published.
The present experimental studies are used to obtain semiempirical reiations which enable one to cal-
culate the spatial distribution of the dosage rate within the irradiation zone and beyond its boundaries. The
dosage rate at the surface of the ground in the irradiation zone can be calculated with the formula
hE r
p (r) _ ~ v ya? 1.6.10-s \ l -;- , 2r ) R/sec (1)
where Qo is the activity of the source in photons/sec, ya is the mass absorption coefficient in cmz/g of pho-
tons with energy E,y in air, r is the distance in cm between the detector and the projection on the earths
surface of the point at the location of the source, and h is the height in cm at which the source is located.
The variation of the dosage rate beyond the boundaries of the earthen embankment is given by the following
expression:
P (r) =3.9.10-4Q (9.Oc-z,73.10-ar+2.0e-1.z ? i o-s*),
where r is expressed in cm, Pin ?R/sec, and Q in Ci, The first term in Eq, (2) is dominant in the region
r 50 m, the height of the embankment is almost independent of the radius of the -zone of irradiation
and is equal to the height at which the source is located plus two or three meters,
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 Declassified and Approved For Release 2013/01/25 :CIA-RDP10-021968000400030001-7
LETTERS TO THE EDITOR
CALCULATION OF INTEGRATED CROSS SECTIONS OF
COMPTON INTERACTION, SCATTERING, AND
ABSORPTION OF y QUANTA FOR STATISTICAL
MODELING OF TRANSPORT PROCESSES
O. S. Marenkov and V. N. Mitov UDC 539.121:72/'75
The energy dependences of the integrated cross sections of Compton interaction and actual scatter-
ing of yquanta with accuracy up to a constant multiplier are determined by the following expressions:
6(R)?(~- ~2 - ~6 / ln(1-F~)-f- ~6 -I-(1-I-~)-1-F(1-f-~)-2; (1)
Qs (Y)- q3 In (1-~-~)- S2 ('1-I-~)-1 -f- 9SS ~ (1-f-~)-2 -f- 2~2 (1-f-R)-3+
where /3 is twice the energy of yquanta in .units of electron potential-energy, The integrated cross section
of actual Compton absorption va(l3) = v(~ -QS((~).
For statistical modeling of y transport in a substance by the Monte Carlo method, the linear atten-
uation factors for Compton interaction, scattering, and absorption are calculated based on (1) and (2). In
the energy region a < 1, degradation of the energy of the quanta during "slowing" occurs comparatively
slowly, and repeated .application of Eqs, (1) and .(2) from the point of view of expenditure of computer time
becomes less economical owing to the presence of a logarithmic function, which-can be calculated using a
standard subprogram,
It is known that in the y-quanta energy region /3 < 1 for the functions rr((3), QS(~3), Qa(/3) we can obtain
expressions in the form of .series in powers of /3, Like powers of the expansion can be a source 'of approxi-
mation formulas.of polynomial type for calculation of the cross sections. We obtain power expansions in
general form, To do this we use the well-known expansions in powers of series of logarithmic and binomial
functions:
In (1~-~)= ~ (-1)h+l Sk ;
k=1
k=1
Series (3) converges in the interval -1 < /3