SOVIET ATOMIC ENERGY - VOL. 38, NO. 1

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• Document Page Count: 94
• Document Creation Date: January 4, 2017
• Document Release Date: September 25, 2013
• Sequence Number: 1
• Publication Date: July 1, 1975
• Content Type: REPORT

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,
\ ?
(
Russian Original .Vol. 38, No. 1, January, 1975
SATEAZ 38(.1) 1-88 (1975),
SOVIET
? ATOMIC
ENERGY
rc
q))
ATOMHAA 3HEPrI1R
(ATOMNAYA iNERGIYA)
TRANSLATED FROM RUSSIAN
, CONSULTANTS BUREAU, NEW YORK
. ?
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SOVIET
ATOMIC
ENERGY
Soviet Atomic Energy is abstracted or in-
dexed in Applied Mechanics Reviews, Chem-
ical Abstracts, Engineering Index, 1NSPEC?
Physics Abstracts and Electrical and Elec-
tronics Abstracts, Current Contents, and
Nuclear Science Abstracts.
Soviet Atomic Energy is a cover-to-cover translation of Atcmnaya
Energiya, a publication of the Academy of Sciences of the USSR.
An agreement with the Copyright Agency of the USSR (VAAP)
makes available both advance copies of the Russian journal and
original glossy photographs and artwork. This serves to decrease
the necessary time lag between publication of the original and
publication of the translation and helps to improve the quality
of the latter. The translation began with the first issue of the
Russian journal.
Editorial Board of Atomnaya Energiya:
Editor: M. D. Millionshchikov
Deputy Director
I. V. Kurchatov Institute of Atomic Energy
Academy of Sciences of the USSR
Moscow, USSR
Associate Editor: 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. Palei
V. B. Shevchenko
V. I. Smirnov
A. P. Vinogradov
A. P. Zefirov
Copyright C) 1975 Plenum Publishing Corporation, 227 West 17th Street, New York,
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Published monthly. Second-class postage paid at Jamaica, New York 11431.
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SOVIET ATOMIC ENERGY
A translation of Atomnaya tnergiya
July, 1975
Volume 38, Number 1 January, 1975
Recipients of the 1974 State Prize for the Design and Building of the VVER-440 Line of
CONTENTS
Engl./Russ.
Reactor Facilities for Nuclear Power Stations
1
2
ARTICLES
Ten Years of Operating Experience at theUSSR Fiftieth Anniversary Nuclear Power
Station at Novovoronezh ? A. N. Grigortyants, F. Ya. Ovchinnikov, V. K. Sedov,
L. I. Golubev, Yu. A. Akkuratnov, I. V. Prokopenko, I. I. Kustov,
N. A. Isakov, V. D. Dobrynin, A. V. Tsybul,nik, and V. I. Skrypnikov
2
3
Effect of Temperature Distribution on the Swelling of UO2 and UO2?Pu02 Cores
? V. I. Kuz,min and I. G. Lebedev
10
11
BOOK REVIEWS
N. I. Chesnokov, A. A. Petrosov, and A. A. Vinogradov. Optimization of Solutions
(Decisions) in the Development of Uranium Sites ? S. Ya. Chernavskii
15
15
ARTICLES
Hydraulic Resistance in Channels with Surface Boiling ? V. A. Knyazev
16
16
BOOK REVIEWS t
H. Muller, K. Meyersberger, and H. Sprinz. Special Methods of Analyzing Stable
Isotopes ? L. I. Petrenko
21
19
ARTICLES
Influence of Low-Temperature Irradiation on the Phase Composition of Uranium
Alloys Containing Small Quantities of Aluminum and Iron ? Yu. V. Bobkov,
I. A. Naskidashvili, V. V. Petrosyan, and Yu. N. Sokurskii
22
20
Radiolysis of Solutions of TBP in Contact with Nitric Acid. II. Processes of Oxidation
and Nitration ? E. V. Barelko and I. P. Solyanina
25
-'' '
23 ----
On the Magnitude of the Magnetic Field Produced in a Laser Plasma
? A. F. Nastoyashchii
30
27
Parameters of the Neutron Resonances of 241Am in the Energy Range from 8 to 30 eV
? T. S. Belanova, A. G. Kolesov, V. M. Nikolaev, V. A. Safonov,
V. Ya. Gabeskiriya, V. A. Poruchikov, S. M. Kalebin, R. I. Ivanov,
and 0. M. Gudkov
33
29
?ABSTRACTS
Recovery of Neutron Spectrum by the Method of Simulated Spectra Using Set of
Threshold Detectors ? G. M. Obaturov and A. A. Tumanov
38
35
Rating of the Capabilities of Photoneutron Installations for Determining Deuterium
Concentration Utilizing SNM Counters ? R. P. Meshcheryakov, S. V. Popov,
and E. S. Solodovnikov
39
36
Monte Carlo Calculation of the Distribution of Electron-Beam Energy Absorbed in
Matter ?A. M. Zlobin, E. N. Donskoi, and V. V. Khizhnyakov
40
36
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CONTENTS
(continued)
Engl./Russ.
Consideration of the Decay and Accumulation of Mixtures of Genetically Related
Short-Lived Isotopes in y -Spectral Analysis - Yu. A. Zaitsev
41
37
LETTERS TO THE EDITOR
Application of Albedo Equations - V. V. Orlov
42
39
Distribution of Fast Neutrons in Iron-Water Shielding - D. L. Broder, K. K. Popkov,
V. G. Taxoev, I. N. Trofimov, and S. A. Tsvetkova
44
40
Dose Distributions in Phantoms outside 1-1000-GeV Accelerators - N. V. Mokhov,
E. L. Potemkin, and V. V. Frolov
47
42
Measurement of gf(239po/af(2350 and crf(233U)/o-f(235U) Fission Ratios at 2 keV
- V. G. Dvyldisherstnov, V. L. Petrov, and V. M. Furmanov
49
43
Fast Neutron Spectrometry of (04 n) Reactions Using a Deuterated Scintillator
- E. M. Burymov, S. P. Korsunova, and N. N. Spendiarov
52
45
Initial Static Focusing in Small Linear Traveling-Wave Accelerators - A. D. Vlasov. . .
54
-N 46
Dependence of the Fission-Fragment Sputtering Ratio for Thin Layers of a Substance on
the Mean Energy of the Fragments - B. M. Aleksandrov, N. V. Babadzhanyants,
I. A. Baranov, A. S. Krivokhatskii, L. M. Krizhanskii, and V. V. Obnorskii.
57
47
Silver Chloride Track Detectors - N. P. Kocherov, N. R. Novikova, and N. A. Perfilov.
60
49
Change in the Optical Density of PMMA under the Action of Deuterons with Energies
4-150 keV - S. P. Kapchigashev, V. P. Kovalev, E. S. Baridaatov,
and V. A. Sokolov
62
50
COMECON NEWS
Provisional International Workteam - P. M. Tyulditin
64
51
Collaboration Daybook
65
51
INFORMATION: CONFERENCES AND MEETINGS
The Ninth World Energy Conference - Yu. I. Koryakin
68
53
Intensified Heat Transfer in Channels of RBMK Type Reactors - A. N. Ryabov
and V. N. Filippov
73
56
The Second All-Union Conference on the Metrology of Neutron Radiation at Reactor and
Accelerator Facilities - R. D. Vasiltev
76
58
MIFI Twenty-Sixth Scientific Conference Held - V. V. Frolov and V. A. Grigortev .
78
59
The Fourth All-Union Plasma Physics School - V. A. Papadichev
80
60
INFORMATION: SCIENCE AND ENGINEERING LIAISONS
Familiarization Trip by Soviet Specialists to Sweden - A. D. Amaev. .
83
61
BOOK REVIEWS
New Books from A tomizdat
86
63
The Russian press date (podpisano k pechati) of this issue was 12/23/1974.
Publication therefore did not occur prior to this date, but must be assumed
to have taken place reasonably soon thereafter.
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RECIPIENTS OF THE 1974 STATE PRIZE FOR THE
DESIGN AND BUILDING OF THE VVER-440 LINE OF
REACTOR FACILITIES FOR NUCLEAR POWER STATIONS
G. L. Lunin
E. S. Chernorotov
V. F. Ostashenko
V. V. Stekoltnikov
V. P. Denisov
A. B. Sukhov
N. I. Gorelov
P. M. Verkhovykh
N. M. Kantiev
L. M. Voronin
V. K. Kalashnlkov
A. T. Glushkov
Translated from Atomnaya Energiya, Vol. 38, No. 1, p. 2, January, 1975.
? 1975 Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011. No part of this publication may be reproduced,
stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying, microfilming,
recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $15.00.
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ARTICLES
TEN YEARS OF OPERATING EXPERIENCE AT THE
USSR FIFTIETH ANNIVERSARY NUCLEAR POWER
STATION AT NOVOVORONEZH
A. N. Grigortyants, F. Ya. Ovchinnikov,
V. K. Sedov, L. I. Golubev,
Yu. A. Akkuratnov, I. V. Prokopenko,
I. I. Kustov, N. A. Isakov,
V. D. Dobrynin, A. V. Tsybultnik,
and V. I. Skrypnikov
UDC 621.311.2:621.039
The Novovoronezh nuclear electric power generating station went into operation on September 30,
1964, when the first experimental full-scale power-generating mit, rated 210 MW(e), began to feed current
into the national grid.
The Novovoronezh nuclear power station is the largest nuclear electric power generating plant built
in our country. The total ratings of its four nuclear power generating units total 1455 MW(e).
Construction work on the NVAES (Novovoronezh nuclear power station) has been proceeding at
accelerated tempos. The first power generating unit with its VVER-210 reactor was built in the space
of 7.5 years; the second power generating unit with its VVER-365 reactor went on the line in 1969, a mere
5.5 years since ground was broken for the plant. The principal quantity-manufactured power generating
units with their VVER-440 reactors were installed and put into service in even shorter periods of time.
The building of the third and fourth power generating wilts was begun in 1968, and construction work was
completed within less than five years, with a one-year offset, the third being completed in 1971 and the
fourth in 1972. The equipment of the fourth power generating unit was installed within one year at most.
The construction program for the first echelon of the NVAES power station was completed successfully
when the fourth power generating unit began generating power for the national grid.
Startup and adjustment operations were carried out in stages, so that adjustment of subsystems could
be attended to several months in advance of the completion of installation and rigging operations, to be
followed by comprehensive adjustments of overall equipment and entire systems. This method was relied
on for startup of several other nuclear power stations, such as the Kola station, the Nord power station
(in GDR), the Kozlodui power station (in Bulgaria), and others.
In addition to the speedup in construction, installation, rigging, and startup and adjustment opera-
tions, the lead time for installation of capacity following the physical startup and power startup was also
shortened. This was particularly striking in the case of the quantity-manufactured units with VVER-440
reactors. The scheduled year-long period was cut down to six months when the third power generating
unit went on the line, and that period was cut down to a mere 83 days in the case of the fourth power gener-
ating unit.
Engineering cost indices for the power station (see Tables 1-3) have been continually improved in
the process of putting the plant into operation and expanding it.
Versatility of the Novovoronezh Nuclear Power Station. The power station operates as an integral
part of the Voronezhenergo power grid, mainly at base load with a slight drop in load at the end of the
week and on holidays or off-days, when the number of industrial consumers of electric power decreases
Translated from Atomnaya Energiya, Vol. 38, No. 1, pp. 3-10, January, 1975. Original article
submitted September 30, 1974.
2
? 1975 Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011. No part of this publication may be reproduced,
stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying, microfilming,
recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $15.00.
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TABLE 1. Characteristics of NVAES Basic Power Equipment
Equipment
_
1st power genera-
ting unit
2nd power genera-
ting unit
3rd and 4th power
generating units
Reactors
type
VVER-210
VVER-365
VVER-440
thermal power, MW
760
1320
1375
weight of pressure vessel, tons
185,4
209,2
200,8
dimensions of pressure vessel, m
4,4x11,1
4,4x12
4,27 x 11 ,8
operating coolant pressure, ligt/cm2
100
105
125
Steam generators
capacity, tons/h
230
325
455
saturated steam pressure, kg/cm2
32
33
47
heat transfer coefficient, k irri2??C
4,29
4,37
4,32
heat transfer surface area, m?
1290
1800
2500
U-tube diameter, mm
21x1,5
16x1,4
16x1,4
number of tubes
2774
3664
5146
dimensions of steam generator, m
Main circulation pumps
3,06x11,57
3,07x11,57
3,27x11,99
type
GTsgb1 -138
GT5EN-3 09
GTsgN-310
power intake, kW
1650
1530
2260
capacity, m3/11
5250
5600
7000
head, kgf/cm
average duration of maintenance-free operation, h
4
5000
5
20 000
, 5
15 000 *
Turbogenerators
output, MW
70
73
220
number
3
5
2
'From the time the third power generating unit was started up.
A constant shortage of electric power in the grid imposes stringent limitations on the length of reactor
fuel reloading shutdown periods, which occur in spring and summer months for the most part. In some
instances, particularly during the autumn and winter peak loads, a need is felt to stretch the in-pile time
of the fuel loading by cutting down on power output toward the end of the reactor campaign.
Operating the reactor on power effect lowers the fuel component of the net power costs. But a de-
crease in the amount of electric power generated and an increase in the capital component of the net power
costs restricts the time over which operation under that set of conditions can be justified economically.
When heightened wear on the equipment as the power generating unit is operated at reduced power
output levels is taken into account, we see that reliance on power effect must be treated as both forced
and provisional.
Dynamic tests showed the reactors exhibit excellent self-regulating properties in the face of such
major disturbances as tripping out of one or several main circulation pumps, outing of the turbogenerator
by disconnection from the power line, and so on. The self-stabilization capability of the reactor eases the
transition to the on-power level corresponding to the new allowable operating conditions.
Some experience has already been acquired at the power station in altering the power output of the
power generating unit over the course of a full day in order to cover the morning and evening peak loads
in the power system. For example, the second power generating unit operated under that set of conditions
throughout January of 1972. The power output of the power generating unit was maintained at the rated
level every day during the morning and evening hours of peak demand, and cut back 30% at nighttime during
that period.
The degree of adjustability of the nuclear power station is also characterized by the comparatively
short time it takes to bring the power generating unit up to full power after a prolonged shutdown period
for cooldown of the primary loop. The total time it takes to carry out a thoroughgoingcomplex checkout of
the reactor control and emergency protection system, to attain the primary loop parameter ratings, to
heat up the turbogenerators and synchronize them with the overall system, amounts to 15 h more or less.
The same time is required for reactor cooldown and to adjust the reactor to the appropriate state for
maintenance and repair operations.
The operating stability of the power station is of exceptional importance for normal electric power
supplies to the consumers. The stability exhibited by the NVAES is demonstrated to be quite high by the
data entered in Table 4. The total number of shutdowns experienced by the third, second, and fourth
power generating units over the 1970-1973 period was 57. The bulk of these shutdowns (43 out of the 57)
took place while the power generating units were being adjusted and run through their first paces. Once
adjusted to power level, the power generating units have been operating with excellent stability.
3
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Fig. 1. Novovoronezh nuclear power station. General view of third
and fourth power generating units.
? Fig. 2. Novovoronezh nuclear power station. Construction work on
the fifth power generating unit.
Basic Power Equipment. Reliable performance on the part of the basic power equipment is respon-
sible for the high operating stability of nuclear power stations. The power station staff and personnel care-
fully studies the performance characteristics of the equipment and the technological flowsheets, and
selects optimum operating conditions.
As demonstrated by separation tests and thermal hydraulics tests carried out in collaboration with
the F. E. Dzerzhinskii VTI, steam generators satisfy the necessary requirements in terms of steam
quality and steam capacity. But isolated cases of tube failure were observed in two steam generators (the
No. 3 steam generator in the first power generating unit and the No. 9 steam generator in the second power
generating unit). The failed pipes were blanked off in order to eliminate any leakage of primary-loop radio-
active water into the secondary loop.
4
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TABLE 2. Engineering Cost Data on Power Generating Units
?
Indices
1st power
generating
unit
2nd power
generating
unit
3rd power 14th
generating !generating
unit
power
unit
Electric power output. MW
210
365
440
440
Year construction was begun
1957
1965
1968
1969
Specific capital investments, rubles/kW
326
256
200
200
Physical startup
XII 1963
XII 1969
XII 1971
mi 1972
Power startup
IX 1964
XII 1969
XII 1971
XII 1972
Power rating achieved
XII 1964
IV 1970
VI 1972
IV 1973
Gross efficiency of power generating unit
27,6
27,6
32,0*
32,0*
In-plant power needs CP of power rating)
8,0
7,3
6,8
6,8
Total electric power generated by. 1974", million kWh
9193,1
8487,7
4010,5
2790.6
Planned output of electric power in 1974, million kWh
Annual installed power utilization factor achieved by
1357
2671
2945
2577
1974
0,87
0,81
0,55
0,72
Annual electric power net costs attained by 1974, kopeks
/kWh
0,925
0,643
0,834
0,716
Ti??irt?TiiTe condenser pressure 0.035 atm.
TABLE 3. Engineering Cost Data for NVAES Power Station Operation 1971-1975
Indices
1971
1972
1973
1974
(planned)
1975
(planned)
Electric power generated, million kWh
2027,2
5413,4
8674,7
9550
9900
Electric power net costs, kopeks/kWh
0,948
0,810
0,752
0,66
0,65
Installed power utilization coefficient
0,633
0,607
0,681
0,753
0,776
Operating experience with the steam generating units has demonstrated clearly that deposits on the
pipes, on the secondary-loop side, are of an oxide nature, are readily removable, and do not affect the
heat-transfer process. But rigid standards must be imposed on secondary-loop water management in
order to keep the pipes in good working condition. For example, feedwater for the steam generators was
excessively high during the first reactor campaign of the third power generating unit, because of the
unsoundness of some of the turbine condensor pipes. As a result, investigations carried out during fuel
reloading revealed, the piping suffered extensive contamination due mainly to iron oxides, so that the
steam generators had to be blown down. The water management regulations were revised to facilitate
removal of contaminants from the piping. A limiting hardness of 0.1 mg-equiv. was set, with turbines
outed immediately for repair and maintenance work on the pipes in the event that level was exceeded.
Special condensate cleanup systems for the secondary-loop water were proposed for the future, as is
the situation at most fossil-fuel electric power generating stations.
While the first power generating unit was in service, some shortcomings in the design of the main
circulation pumps deemed responsible for several unscheduled shutdowns were corrected. The various
subsystems of the pumps were redesigned and modernized, and the tubing arrangement entering the self-
contained cooling system for the bearings was simplified, so that the impeller system could be changed
with the object of minimizing fluctuations in head, and the stator blading cooling system was centralized.
At the present time, the obsolete GT5EN-138 pumps on stream in the first power generating unit
have been replaced by the more sophisticated GT5EN-309 pumps employed in the second power generating
unit.
A technique of vibratory acoustic diagnostics worked out by the NVAES staff has been in use to
monitor malfunctions of the main circulation pumps. Acoustic transducers have been installed on the
pumps, and secondary instruments have been located on the reactor control panels. The method in question
aids in pinpointing typical pump malfunctions such as excessive wear on bearings, beating of the rotor on
the casing, and so on.
Operation of 70 MW turbogenerator sets installed in the first and second power generating units of
the power station has demonstrated the possibility of stepping up the power output of each turbogenerator
set to 80 MW without any deterioration in efficiency. The unit power output of the turbines in the third
and fourth power generating units has been increased to 220 MW. The use of high power rating sets and
the increase in the performance parameters of the coolants has made it possible to obtain better perfor-
? mance per cost from the power generating units.
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TABLE 4. Data on NVAES Operating Stability
Indices
Before attaining power rating
After attaining power, rating
2nd power
genera-
ting unit
1970
3rd powei
genera-
ting unit
1972
4th power
genera-
ting unit
1973
2nd power generating unit
3rd unit
1971
1972
1973
1973
Number of shutdowns of power generating
units and electric power not generated in
10 kWh) because of equipment malfunc-
9
25
9
6
2
1
5
tions
222,4
220
55,7
81,4
1,6
19,8
86,0
Number of shutdowns of power generating
units and electric power not generated (in
106 kWh) because of personnel judgment
2
5 .
2
None
None
errors
0,7
2,0
0,4
0,3
0,5
Length of time power generating unit on
power, h
5651
7120
8072
5761
7884
7849
7123
It took the combined creative efforts of the power plant operating personnel and of the personnel
of the manufacturing plant to put these new turbines into service.
The operation of the nuclear power station turbines working on saturated steam demonstrated the
importance of the problem of coping with erosion, a severe problem for the turbine components. Specifi-
cally, erosion wear on components of the turbine flow passages in the third power generating unit turbines
was detected, and the unit had to be shut down for an additional period to effect repairs and eliminate the
problem. Work was begun, in collaboration with the personnel of the turbine manufacturing plant, on im-
proving the performance of the NVAES turbines by selecting and testing wear-resistant materials service-
able in the manufacture of turbine components, in plating turbine components, and in hardfacing those
components when buildup is called for during repair and maintenance work.
A first-stage rotor blade of the high-pressure cylinder broke off while the third power generating
unit was being run up to full power, and the noZzle guide vanes of the No. 9 turbine were ruined. The
shank of a rotor blade in the No. 10 turbine was discovered overstrained. Replacement of the blades caused
a two-month shutdown of those turbines. The insufficiently stable performance of the turbines was due
in large measure to the frequent damage suffered by the brass piping of the condensers.
As investigations revealed, the reason for this damage was the inadequate resistance to corrosion
on the part of the piping material. At the present time, preparatory studies are underway at the power
station on how best to replace this piping with other piping presenting improved resistance to corrosive
attack.
The design of the reactor control and protection components underwent some changes while the
reactors were in service, with modifications introduced by designers varying from one power generating
unit to the next depending on the experience accumulated and the uranium fuel enrichment being increased
to 3.6%. The number of assemblies of electromechanically controlled control and protection systems was
increased from 37 to 73 in the second, third, and fourth power generating units, and all of these assem-
blies combine the functions of control and scramming, so that the number of fuel assemblies could be
increased from 343 to 349 with scramming speed retained, and with enrichment of the fuel charge as an
added benefit. A system of fluid control of reactivity with the aid of boric acid has been instituted starting
with the second power generating unit, so that the power distribution throughout the core can be improved
and so that the electromechanical control and protection system can be backed up by a liquid scramming
feature. A rack and gear drive is being used in the power drives of the control and protection assemblies
in the third and fourth power generating units, in place of the earlier screw and ball nut kinematic pair,
and that has improved the speed characteristics of the power drives while enhancing their reliability. In
addition, automatic controllers (known as pulsed power controllers) tracking the parameters of the primary
and secondary loops and responding to any changes in those loop parameters have been introduced into the
systems of these power generating units. The pulsed power controllers also have the job of automatically
lowering reactor power output when several main circulating pumps shut off.
All scramming signals were divided into four distinct categories in terms of degree of hazard, in the
operation of the first power generating unit. In this protection arrangement, a complete shutdown of the
reactor through the simultaneous actuation of all of the control rods released downward would come about
only in instances where the heat-transfer reliability of the core deteriorated severely (e.g., tripping out of
main circulating pumps, an abrupt rise in reactor power output level, total dumping of turbine loads, and
the like). In any remaining cases, reactor power output would be lowered in response to a protection signal
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through sequential downward displacement of groups of control rods, to be terminated with the disappear-
ance of the protection signal. This improvement made it possible to cut down appreciably on downtime
due to actuation of the reactor protection system.
On the whole, the control and protection system fulfilled its functions reliably in proper response
to stimuli.
A procedure for maintenance of the equipment serving \TITER radioactive loops has been instituted
at the NVAES power station, and experience of great practical value has been accumulated in the process.
Inspection and maintenance technology for the reactor pressure vessel has been mastered for the first
time following a protracted service period, and this technology covers the pipe outlets for the principal
process pipes, the sheathed tubing of the power drives from the control and protection system assemblies,
and also the reactor innards (in-pile equipment and devices).
Reactor inspection in a program calling for complete unloading of the reactor core and extraction of
all of the innards from the reactor of the first power generating unit was executed for the first time in the
1970-1971 period. This breakdown inspection required the creative efforts of the entire staff for working
out a suitable repair and maintenance technology, methods of monitoring and inspection, devising protec-
tive devices and custom-engineered fixtures and accessories for remote-controlled in-pile operations,
learning new metal-working techniques (in particular plasma cutting, electric-arc cutting, and electric
erosion cutting of metals), including cutting of metals under water with the aid of carbon electrodes. A
biologically shielded container equipped with handholes and windows covered with special grades of glass
was designed and built to aid in inspection and repair of the reactor pressure vessel. Closed-circuit tele-
vision arrangements were employed liberally in inspecting equipment sets located in inaccessible and
difficult-to-get-at sites. As a rule, all of the repair, maintenance, and inspection operations were tried
out first on mockups simulating the actual conditions. Before this inspection and repair work was begun,
the primary-loop equipment was deactivated by applying special solutions.
Inspection of the reactor in the first power generating unit revealed some serious shortcomings in
the design of the reactor innards, which exhibited inadequate strength and inadequate resistance to vibra-
tion in the coolant stream. The new in-pile equipment, like the in-pile equipment in the other new reac-
tors of this nuclear power station, exhibit heightened ability to withstand vibrations, and feature modular
design which renders the job of monitoring and inspection easier. Cracks were detected on the adapters
of the main pipe connections in the process of inspecting the pressure vessel, and this led to the develop-
ment of a technology for inspection and repair of pipe connections which was implemented for the first
time in this program. Adapters made of 1Kh18N1OT grade steel were inserted and welded in place. The
adapters were press-fitted after cooling in liquid nitrogen, by using a special auxiliary fixture with a
pneumatic power drive developing 3 tons of force.
In 1971, the in-pile equipment of the reactor in the second power generating unit was extracted and
inspected in a similar repair and maintenance program, because of the rupturing of one of the quieting
tubes in the control and protection assemblies. All of the innards of the reactor in the fourth power gener-
ating unit were extracted in April, 1974, at fuel reloading time, for a shakedown inspection of the reactor
pressure vessel. Periodically scheduled inspections of reactor pressure vessels are on the agenda for
all the power generating units of the power station in future practice.
Defects in the weldments joining the sheathed tubes of the power drives of control and protection
system assemblies to the cover of the reactor pressure vessel were detected while the first power gener-
ating unit was in service, where the primary-loop hot water became mixed up with cold water fed into the
sheathes to cool the electric motors of the power drives. Thermal stresses generated in the metal con-
tributed to the appearance of flaws at those sites. After a special maintenance technology had been worked
out in 1967, the welded joints were replaced by flanged joints. This redesigning effort was also required
on the reactor of the second power generating mit in 1970. Operating experience with the mechanisms and
moving parts of the control and protection system serving the first power generating unit was incorporated
in the design of the VITtR-440 reactors of the third and fourth power generating units. Flanged sheath
joints were provided in those reactor assemblies, and the electric motors of the power drives are cooled
in these cases with the aid of a special intermediate loop.
A change in the design of some components of the electromechanical control and protection system
improved system performance. But since the presence of pulsations and temperature gradients affects
the operating characteristics and the state of the metal, designers are obliged to take into cognizance the
operating experience acquired with that system for attention to further improvements.
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A technology developed for replacing the primary-loop main circulating pumps was first worked out
at NVAES power station. With the active participation of the power plant personnel, a repair and main-
tenance machine shop was designed and built for the upkeep of main circulation pumps, with equipment for
deactivation, specialized ventilation, and all necessary accessories on hand.
Equipment maintenance experience accumulated over the 1964-1971 years with the first and second
power generating units made it possible to carry out successful redesign work needed while the power
generating unit was being put through its paces and put into service, and in the work on the quantity-manu-
factured power generating units incorporating VVER-440 reactors.
One of the crucial operations mastered in practice at the NVAES power station is reloading of nuclear
fuel, usually combined with inspection and maintenance of power equipment.
The power generating unit is shut down for short periods (ranging from 25 to 30 days) for fuel re-
placement, and involves such complex operations as failure of bonding in the primary loop, withdrawal
of spent fuel assemblies, and installation of new fuel assemblies. The characteristics of the new fuel
loading are calculated teqtatively on computers. The loading is done in such a way that the required dura-
tion of the next reactor campaign is rendered possible, and the acceptable physical parameters limiting
the reactor power output level are attained. Special attention is given to nuclear safety considerations in
the refueling process and to the performance of the new core. As a rule, reloading of about one third of
the fuel assemblies in each reactor is carried out once a year, during the summertime. But the presence
of four functioning power generating units imposes additional time restrictions, since refueling requires
serious preparation. The refueling operations proceed parallel with monitoring of the soundness and leak-
proofing of fuel element cans in operating fuel assemblies, with the aid of special techniques worked out at
the power station. The method based on measuring the degree of radioactivity of the air in which self-
heating of a fuel assembly deprived of water occurs is the method most widely relied on in practice. Air
activity increases when flaws are present in the fuel element can through which gaseous and volatile fission
products can exit. "Leaky" assemblies are replaced; the makeup of the fuel charge is corrected and the
design characteristics of the reactor core are also corrected appropriately.
The fuel is reloaded underwater, in all of the power generating units of the nuclear power station
except for the second power generating unit, by special machines designed for the purpose, and that
ensures complete safety when appropriate dosimetric health physics monitoring is attended to. In the
second power generating unit, the spent fuel assemblies are conveyed to a storage tank in a special
shielded container. A liquid neutron absorber, boric acid added to the water of the primary loop, is relied
on to maintain the deep subcriticality of the reactor during the refueling process. The neutron flux in the
reactor is monitored by a special refueling auxiliary system with sensitive ionization chambers installed
in the channels of the faceted belt in the reactor support cage.
Research work and pilot tests continually staged at the NVAES power station help, in addition to
scientific-research work and planning and design work carried on by institutions dedicated to those pur-
poses, in shortening the time required for successful adaptation and improvement of the basic equipment
and improvement of engineering cost figures.
Among the most important results achieved are something little short of doubling the power rating of
the newly intrOduced reactors without any increase in core size, and by the same token without any increase
in the size of the reactor pressure vessels. This became possible through improvements in the power dis-
tribution over the core as a consequence of working out a liquid control schedule first implemented in the
second power generating unit. Simultaneously, a ilowsheet and a technology for normal and scrammed
introduction of boric acid into the primary loop, plans for extraction of the boric acid, and plans for clean-
up of loop water and proper water management, were worked out. A modification of the fuel grid with
the dimensions of the fuel assemblies left unaltered made it possible to increase the number of fuel ele-
ments, and consequently to increase the average energy yield per fuel element with the maximum allow-
able load at which there would be no meltdown of the uranium dioxide in the center of the fuel element core
retained.
There are 12 special measuring channels in the central tubes of the reactors in the third and fourth
power generating units, and five such channels in the second power generating unit, for measuring varia-
tion factors in power generation over the core height. The neutron flux distribution in the channels is
measured by activation detectors or direct-charging sensors. At the present time, the possibility of
automating in-pile measurements with the use of direct-charging sensors and data processing through a
M-6000 computer is under study at the power station.
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Methods for monitoring the state of fuel elements present in the reactor core were worked out and
perfected while the power generating units were in service. In addition to radiochemical monitoring based
on the degree of fission-fragment activity on the part of the primary-loop water in the first power genera-
ting unit, a system designed to take continuous recordings of the background level of delayed neutrons in
circulation subloops has also been worked out for the first power generating unit. The system demon-
strated its feasibility and has been instituted in all of the power generating units of the nuclear power
station.
Radiation Safety. Under normal operating conditions, even when several leaky fuel elements turn up
in the core, the total activity of water is not greater than 10-4 to 10-3 Ci/liter. A sufficiently low level
of activity and a high degree of leak-proofing of the primary-loop equipment can ensure a low level of
radioactive contamination of the nuclear power station rooms, and also can place limits on the amount of
gaseous and aerosol waste vented from the production rooms of the four power generating units via the
two exhaust stacks. Data available support the inference that the power station exerts no harmful in-
fluence on the surrounding environment, and that the level of radioactivity of the environment is com-
mensurate with the natural background. Various devices and protection equipment used in service, in
deactivation operations, in inspection and maintenance of radioactive equipment, aid in maintaining a
level of personnel exposure doses below public-health regulation levels.
Even though the probability of a serious accident associated with release of radioactivity from the
primary loop is very small, the station personnel is carrying out intensive work on periodic checkups on
the state of the metal in the equipment and piping of the reactor installation during the period when nuclear
fuel in the reactor is replaced, and a practically complete volume of monitoring and inspection work is
carried out within the space of four to five years. An especially large volume of work in investigating the
metal in the pipings was carried out during the period of inspection of the first power generating unit.
Consequently, a basic summary of experience in the operation of the four power generating units
over the 10-year period that the Novovoronezh nuclear power station has been in service leads to the con-
clusion that significant progress has been achieved in power station equipment and technology in the case
of power stations using VVER type reactors, which are reliable sources of electric power and installations
that are safe for the nuclear power station personnel and staff, for the surrounding population, and for
the local environment.
A further expansion of the Novovoronezh power station is being planned with the scheduled introduc-
tion of a fifth power generating unit incorporating a VVER-1000 reactor, construction work on the reactor
and power generating unit as a whole having begun back in 1973. The latest engineering solutions result-
ing in lower costs and enhanced safety will be materialized in the designs for the fifth power generating
unit. Plans call for building a shielded concrete enclosure for the primary loop, which is intended to com-
pletely localize and trap radioactivity in the event of accidents involving leakage of coolant. The dimen-
sions of the core are being increased by a factor of almost 1.5 while the pressure vessel size remains the
same, and this is being achieved through modifications in the reactivity control system. The unit power
rating of the turbogenerators is being increased to 500 MW. After the fifth power generating unit starts
generating power at the outset of the next Five-Year Plan, the total power output of the NVAES will be
increased by another 1000 MW. Engineering cost figures for the operation of the nuclear power station,
which are even now comparable to the figures for the operation of fossil-fuel power generating stations
in the central region of the nation, are undergoing impressive improvements at the same time.
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EFFECT OF TEMPERATURE DISTRIBUTION ON THE
SWELLING OF UO2 AND UO2 ? Pu02 CORES
V. I. Kuztmin and I. G. Lebedev UDC 621.039.542.342
One of the processes influencing the efficiency of fuel elements, including those based on oxide fuels
of the UO2 and UO2?Pu02 types, is the radiation-induced swelling of the fuel. Experimental data [1, 2]
relating to the influence of individual parameters on volume changes taking place in these types of fuel
as a result of irradiation often disagree both quantitatively and qualitatively. Calculations of volume
changes in oxide fuels (especially those due to gaseous fission products) cannot as yet be carried out with
adequate accuracy. ?
The contribution of solid fission products to swelling was determined earlier [3, 4]. In the pre'sent
investigation we determined the influence of temperature distribution in irradiated UO2 and UO2?Pu02
fuels on the total porosity, and considered methods of calculating the volume changes taking place in cyl-
indrical oxide-fuel cores.
METHOD
We studied UO2 samples and mixed UO2?PuO2 fuel containing15 and20%Pu02. Sintered moldings were
placed in 01011 61\115M3B steel cans and these were then sealed hermetically by welding.
The samples were irradiated in an SM-2 reactor to various degrees of burn-up (0.4-17% of the heavy
atoms) (Table 1). The can temperature was measured during irradiation with Chromel?Alumel thermo-
couples, and the fuel temperature was determined by calculation. The maximum error in determining the
temperature of the fuel in the center of the moldings was not greater than 200?C. The burn-up was deter-
mined by y or mass spectrometry.
The swelling of the fuel in all samples except those irradiated with a linear power of over 400 W/cm
took place without any restraint on the part of the can. The gas pressure inside the samples at the end
of irradiation was no greater than 30 kg/cm2.
The samples were studied under the microscope. Using photographs of the microstructure and the
method of secants [5], the total porosity was determined in several regions at different radii. The volu-
metric proportion of pores with a diameter of over 0.8 p was determined in the optical microscope, and
that of pores with diameters of 0.02-0.8 p in the electron microscope by the replica method [6]. The con-
tribution of still finer pores with diameters of 15-32 A was calculated after studying the UO2 by trans-
mission electron microscopy [71.
RESULTS
The results of the porosity measurements for different degrees of burn-up are presented in Fig. 1,
which also shows the porosity of the non irradiated samples (for comparison). The porosity/burn-up rela-
tionship and the absolute values of the porosity corresponded to the existence of three temperature zones,
1000-1500, 1500-1700, and 1700-2500?C, respectively.
In zone 1 the porosity of the samples before irradiation equalled 5-9%; after irradiation to burn-up
values of 1-5% it fell sharply. The fall varied in different cases. The maximum reduction in porosity
occurred in samples irradiated to a burn-up of 1%, in which the grain size and the size of the original
Translated from Atomnaya Energiya, Vol. 38, No. 1, pp. 11-15, January, 1975. Original article
submitted December 24, 1973.
0 1975 Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011. No part of this publication may be reproduced,
stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying, microfilming,
recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $15.00.
10
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TABLE 1. Characteristics of the Samples TABLE 2. Values of the Coefficients CI and
and Conditions of Irradiation C
Composition of sam-
ple
0 ,tz
41
is. 70.,
0 .m
Temperatu
of fuel
0.)
usz.
gt,)
:35
>
o
0.
E
4)0
on the
sur-
face
in the
cen-
ter
?
UO2
5
700
1700
0,4
350
UO2
5
1230
1600
1,0
275
UO2
5
1400
2200
1,3
345
1.102
6
920
1550
2,7
250
UO2
7
1230
2350
2,9
440
UO2
6
1120
1800
3,0
305
UO2
8
800
1450
4,2
255
UO2
6
1050
2250
4,8
510
UO2
2,5
890
1460
5,3
320
UO2
9
780
1860
5,6
300
1102
8
1000
2050
6,6
415
UO2
9
1100
2500
8,1
500
UO2
8
1100
2000
12,8
310
(0,2 Pu-0,8 U)02
2
1100
1800
14,6
245
(0,15 Pu-0,85 U)02
3
1150
1750
17,1
310
Coeffi-
cient
Temperature range, ?C
1000-
1500
1500?V700
1700-
250>
dT dT
dr 7". I dr >
500?C/mm
? ci
cio
1,04
?KP0
0,32
15 ? Po
0,32
? Po
0,32
5 ? Po
pores before irradiation were minimal. The porosity of the
samples originally at a level of 5-7% remained unchanged
after irradiation to a burn-up of 6-7%, but after a 14-17%
burn-up it increased from 2-3 to 12-15%. The porosity of
the irradiated samples depended linearly on burn-up within
the range of measuring error (relative 15%).
In zone 2 the porosity depended, not on the burn-up
and original porosity, but only on the temperature gradient;
this applied over the whole range of burn-ups (within the
limits of measuring error). The absolute porosity in zone 2 for a gradient of under 400 deg/mm averaged
15%, and for a gradient of over 500 deg/mm it varied between 5 and 10%. In the zone of columnar grains
(zone 3) the porosity was nonuniform. In the central part close to the cavity the porosity was usually
higher, but the average value amounted to ?5%.
DISCUSSION OF RESULTS
The reduction in porosity observed in zone 1 after irradiation to a burn-up of ?1% might be
asociated with two processes: sintering, and the filling of pores with solid fission products. In the
samples irradiated to a burn-up of 1% the porosity diminished from 5 to 1%, while the volume of
solid fission products for a burn-up of 1% was less than 0.7% of the volume of the fuel with the theo-
retical density [4]. It follows that the reduction in porosity was in fact due to sintering. This con-
clusion agrees with the results of [8]. The increase in porosity for greater degrees of burn-up as-
sociated with the increasing dimensions and volumes of the pores due to gas accumulation. It is well
known that the evolution of gas from this zone in UO2 tablets with a density of 10-10.7 g/cm3 equals
20-50% for a burn-up of '.4% [9], while the evolution of gas from zones 2 and 3 equals ..-'100% for
a burn-up of over 1%. Growth of the pores takes place chiefly at the grain boundaries. In samples
in which the zone size equalled 15-20 ?, considerable porosity also appeared inside the grains (at
the subboundaries). The increment in the porosity of zone 1 observed for fuel with the theoretical
density (introducing a correction for the nonuniform burn-up over the sample cross section) equalled 0.72%
/1% burn-up within the range studied. In zones 2 and 3, saturation of the porosity changes occurred even
at a burn-up of ?1%. The reason for this may be understood by analyzing the structure of the fuel. The
pores in these zones lie chiefly at the grain boundaries and form a network of interconnected channels,
which lie mainly on the lines connecting three neighboring grains. For this geometrical disposition of the
pores, a considerable proportion of the surface of the grain boundaries is linked to the free surface, and
the gas passing to a grain boundary emerges freely from the fuel (Fig. 2).
The intragranular porosity in zones 2 and 3 also fails to increase on increasing the burn-up above
1%, since the rate of gas evolution at the grain boundaries is such that the krypton and xenon atoms so
formed reach the grain boundaries in a time corresponding to a burn-up increment of 0.2%. This is con-
firmed by calculating the velocity of the gas-filled pores in the field of the temperature gradient [10].
Saturation of porosity with increasing burn-up should also occur in zone 1 owing to the transition
of a considerable proportion of the isolated gas pores to the open surface. However, saturation of the
porosity in zone 1 was not in fact reached for a burn-up of 17%, owing to the low mobility of the gaseous
fission fragments. For the same reason it would appear that saturation in zone 1 will set in gradually,
from the region of highest temperature to the region of low temperature.
The swelling of the fuel in the samples under consideration took place gradually as the gas pressure
within the can increased, starting from 1-2 leg/cm2 at the onset of irradiation to 30 kg/cm2 at the end.
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15
14
12
6 8 10 12 ?16 18
Burn -up, u/0
Fig. 1
Fig. 2
Fig. 1. Porosity as a function of burn-up in various temperature zones; 0) zone
1; G) zone 2, dT /dr 500 deg/mm);x)
zone 3; A) original porosity.
Fig. 2. Linking of pores at the grain boundaries.
The external gas pressure on the fuel was insufficient to have any major effect on the porosity by virtue
of gas-induced swelling. In samples irradiated with a linear power of over 400 W/cm the can was to some
extent able to restrain the swelling of the fuel in the radial direction. However, porosity measurements
revealed no marked influence of radial restraint.
Zone Model for Calculating Volume Changes in
Fuel-Element Cores
The zonal character of the temperature dependence of the changes in porosity and also the quantita-
tive data relating to changes in the volume of the fuel due to solid and gaseous fission products enable us
to calculate the total volume effect in the cores of oxide fuel elements. If we consider part of the core
with an external radius of R as having a length 1 so short that any change in the linear power and temper-
ature along its length may be neglected, and assume that the burn-up is independent of radius, the in-
crease in the volume of the annular layer may be expressed as follows:
AV i =al (?+i) (CB
(1)
where Ci and Cio are coefficients characterizing the swelling in the zone limited by the radii ri and ri ?i;
and B is the burn-up. The radius ri corresponding to the temperature Ti for a specified linear power q1,
core surface temperature TR, and thermal conductivity A is determined from the equation
T
( i
4n S ), (T) dT
rl = R2 1 TR .
qt
The relative change in the volume of the selected part of the core is determined from the equation
AV/V.= 4ril qi 2 (CB --00) (T) d7:
i=1 Ti
(2)
(3)
The number of zones depends on the temperature distribution. If the temperature in the center is
no greater than 1700?C, the calculation is carried out with a two-zone model. For a higher temperature a
three-zone model is needed.
The coefficients C1 and Cio determine the terms in (3), respectively, dependent and independent of
the burn-up. The value of the coefficients Ci equals the sum of the changes in volume of the i-th zone of
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10
2
4 6
Porosity, lo
Fig. 3
8 10
-2
-4
7
J 5 7 9
Burn-up, ?10
Fig. 4
11
13
Fig. 3. Increase in the volume of a uranium dioxide core as a function of the
porosity of the moldings: and - ?- ? - ? -) Tsurf= 1000 and 1200?C; qi
=350 and 550W/cm, respectively.
Fig. 4. Increase in the volume of uranium dioxide cores as a function of burn-up
(K = 0.5). Curve nomenclature as in Fig. 3.
the core due to the solid and gaseous fission products on increasing the burn-up by 1%. The change in vol-
ume due to the solid fission products is the same for all the zones and equals -0.4% [3] or according to
refined data 0.32%. The change in volume due to the gaseous fission products is 0.72% for the first zone
as already indicated and zero for the second and third zones, since the porosity is then independent of
burn-up (Fig. 1). After summing the contributions of the solid and gaseous fission products, we obtained
the coefficients C1, C2, and C3.
The coefficient C1;0 equals the product KP0. The coefficient of radiation-induced sintering K deter-
mines that part of the original porosity Po which vanishes as a result of sintering under irradiation. The
values of the coefficients C2;0 and C3;0 equal the difference between the porosity corresponding to satura-
tion and the original porosity.
Table 2 gives the coefficients Ci and Cio used for the burn-up range 1-17%, an original porosity
of 2-10%, and temperatures of 1000-2500?C.
Substituting the corresponding values of Ci and Cio we obtain the following for the two zones:
.2B ( 100-P0 41r [
100 1 TC
1500
Ar ? 100?o' -0 3
(O. I2B?KPo) (T) dT (15 - P 0) X (T) dr] ;
1500
surf
and for the three zones:
(4)
1500 1700
av?100% =
0.32B ? . -
( 1??- ) 43 [(0 72B KP (T) dT?(7-P0) S (T) dr (5 - P0) X (T) dr]. (5)
iuu
Tsurf 1500 1700
Equations (3)-(5) may be used for a solid core if the external pressure is no greater than 30 kg/cm2.
Figures 3 and 4 give the results of a calculation of the volume changes taking place in the fuel for various
initial porosities, burn-ups of 5 and 10%, and linear powers of 350 and 550W/cm. In the calculation it was as-
sumed that the sintering coefficient equalled 0.25 or 0.5 and the temperature on the fuel surface 1000 or
1200?C. As indicated by Eqs. (4) and (5) and Figs. 3 and 4, an initial porosity is useful for reducing the
volume changes in all the temperature ranges. With increasing linear power the swelling is reduced, since
the volumetric proportion of zone 1 in the core diminishes, and that of the fuel in the plastic zone 3 (less
subject to gas swelling) increases. This may have important consequences, since the deformation of the
can is mainly due to the "rigid" peripheral layer of fuel.
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Thus we have analyzed the distribution of porosity in the cores of UO2 and UO2-Pu02 fuel elements
irradiated in the burn-up range 0.4-17% at linear powers of 250-510W/cm by quantitative metallography.
We have given a qualitative description of the mechanisms underlying gas-induced swelling and gas evolu-
tion in the oxide fuel. We have shown that gas swelling reaches saturation in the core zone for a burn-up
of -1% and temperatures of over 1500?C. In the core zone the porosity increases with burn-up at a tem-
perature of 1.4 MeV in
iron?water mixtures of varying composition; the values were obtained by an analysis of the results of
the calculations. The table makes it clear that an insignificant reduction in relaxation length is observed
as the source distance increases for iron?water mixtures containing more than 50 vol. % of iron. A (; :
similar effect was observed in an analysis of the penetration of fast neutrons from monoenergetic sources
in heavy media [7].
LITERATURE CITED
1. D. L. Broder et al., At. Energ., 27, No. 3, 217 (1969).
2. L. I. Abagyan et al., Group Constants for Reactor Calculations [in Russian], Atomizdat, Moscow
(1964).
3. M. N. Nikolaev and N. 0. Bazazyants, Anisotropy in Elastic Scattering of Neutrons [in Russian],
Atomizdat, Moscow (1972).
4. B. R. Bergeltson et al., Multigroup Methods for Neutron Shielding Calculations [in Russian], Atom-
izdat, Moscow (1970).
5. A. P. Veselkin et al., Atlas of Fast-Neutron Spectra [in Russian], Atomizdat, Moscow (1970).
6. Radiation Shielding Handbook for Engineers [in Russian], Vol. 1, Atomizdat, Moscow (1972), p. 287.
7. T. A. Germogenova et al., Fast-Neutron Transport in Plane Shields [in Russian], Atomizdat, Mos-
cow (1971).
46
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DOSE DISTRIBUTIONS IN PHANTOMS OUTSIDE
1-1000-GeV ACCELERATORS
N. V. Mokhov, E. L. Potemkin, UDC 621.039-78
and V. V. Frolov
We present here calculated results obtained from the FORTRAN program SHIPHA for the depth dose
En phantoms produced by ionizing radiation produced in iron shielding of various thicknesses by primary
protons with energies E0 =1-1000 GeV and mit flux density. In the SHIPHA program, calculation of the
spectral and angular distributions of protons, neutrons, r?-mesons, K?-mesons, muons, and y -rays
outside the shield is performed in the subprogram HAMLET [1, 2] over the energy range 10'2-1012 eV.
Calculation of the dose distribution in a slab phantom 30-cm thick outside the shield is performed in the
subprogram FANTOM [3].
The depth distribution for absorbed dose in the phantom is plotted in Fig. 1 for protons of energies
1, 10, 100, and 1000 GeV incident on an iron shield having a thickness H =1500 g/cm2. The distribution
is exponential in nature because the dose is -90% determined by neutrons with E 5 10 MeV. Calculations
show that the addition of a concrete layer -110 g/cm2 thick to the iron shield reduces the dose by roughly
a factor of five and leads to comparable contributions to the dose from low-energy neutrons and high-
energy hadrons.
The energy dependence of the dose equivalent and the absorbed dose in a phantom outside an iron
shield is shown in Fig. 2. There is a ',critical thickness,' of the shield at each energy such that the dose
in the phantom for shield thicknesses less than the critical thickness is greater than without a shield.
10')
2 10
70-5
10
2
0) 10
1:1 2.2
? 10-8Cr'
a)
2 CI hie
?-cY
?c) !3--'
du ill10-9
7,,
I5 az-
al
7
101 13? 10 ' 102 103 ler
0 10 20 ,cm Eo, GeV 0 1000 2000 3000 H, g/cm2
Fig. 1 Fig. 2 Fig. 3
70-io
Fig. 1. Depth distribution of absorbed dose in a phantom outside an iron shield for various
primary proton energies.
Fig. 2. Dependence of dose equivalent (1) and absorbed dose (2) at a phantom depth x = 5 cm on
the energy of protons incident on a shield of thickness H =1500 g/cm2; 3, 4) corresponding val-
ues without shielding.
Fig. 3. Dependence of dose equivalent (1) and absorbed dose (2) at a phantom depth x =12 cm
on thickness of an iron shield (E0 = 50 GeV).
Translated from Atomnaya Energiya, Vol. 38, No. 1, pp. 42-43, January, 1975. Original letter
submitted April 15, 1974.
? 1975 Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011. No part of this publication may be reproduced,
stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying, microfilming,
recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $15.00.
47
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Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050001-5
? Dependence of dose equivalent and absorbed dose on the thickness of a shield irradiated by protons
with an energy E0 = 50 GeV is shown in Fig, 3. Outside a shield of thickness ,.,3500 g/cm2, the dose is
determined entirely by the muon component and the absorbed dose and dose equivalent are the same.
The program SHIPHA makes it possible to calculate dose distribution in phantoms of arbitrary shape
outside heterogeneous shields of thicknesses up to 5000 g/cm2. The program calculates the component
contribution to absorbed dose and dose equivalent. The primary energies of broad, monodirectional beams
of hadrons are E0 = 0.05-1500 GeV. The reliability of the resultant data was established by comparison of
the calculated results from the HAMLET and FANTOM programs with experimental data in the energy
range 0.1-100 GeV. The disagreement did not exceed 10% for dose calculations in phantoms [3, 4] and
30-50% for calculation of the intranuclear cascade in the shield [1, 2]. The computing time required to
obtain results with an accuracy of 5-10% is ?1-1.5 h on the BESM-6. The program is useful for the estima-
tion of the radiation environment to be encountered by space vehicles and that around existing and planned
accelerators.
LITERATURE CITED
1. L. R. Kimelt and N. V. Mokhov, Izv. VUZ, Fiz., No. 10, 18 (1974).
2. G. I. Britvich et al., IHEP Preprint 74-86 (1974).
3. V. T. Golovachik et al., IHEP Preprint 73-29 (1973).
4. V. T. Golovachik et al., IHEP Preprint 74-58 (1974).
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Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050001-5
MEASUREMENT OF of(2 39Pu)/of(23 5 U) AND (23 3U)/af(23 5U)
FISSION RATIOS AT 2 keV
V. G. Dukhsherstnov, V. L. Petrov, UDC 539.1.083
and V. M. Furmanov
Evaluations of the effect of uncertainties in nuclear data on the design parameters and economics of
fast breeder reactors [1, 2] indicate the need for refinement of the energy dependence of the effective
cross sections for basic reactor materials. This refers primarily to the fission and radiative capture _ _
cross sections in U and Pu isotopes, for which the measurements still fail to show sufficiently good agree-
ment.
Useful information for the formulation of a recommended library of evaluated nuclear data can be
obtained from measurements of the fission cross sections in U and Pu isotopes relative to the fission
cross section of 235U. In the neutron energy range 1-5 keV, however, there is an extremely limited set
of experimental data from "direct" measurements of the of(239Pu)/01(235U) [31 and o-f(233U)/crf(235U) [3, 4]
ratios or from simultaneous measurements of of(23913u) and af(235U) [5, 61. Furthermore, the differences
in the af(239Pu)/o-f(235U) ratios in [3, 5, 6] are significantly greater than the errors cited by the authors.
This paper gives the results of "direct" measurements of the quantities o1(239Pu)/af(235U) and 7f(233U)
/uf(23 5U) in the scandium neutron beam (2 keV) of the reactor at the Obninsk nuclear power station.
METHOD
Measurements of fission cross-section ratios were carried out in the "signal" spectrum of the
scandium beam of the reactor at the Obninsk nuclear power station [7] by means of double fission cham-
bers. Each chamber (239Pu/235U and 233U/235U) had five aluminum-foil electrodes ?0.1 mm thick on two
of which a layer of fissile material was deposited on both sides in the form of a spot 14 x 9 mm in size.
The total amount of fissile material in a single chamber was ?3 mg:1 mg of 239PU dioxide (-99.8%) or
of uranous?uranic oxides of 233U (-99.9%) and 2 mg of uranous?uranic oxides of 235U (-90%). The elec-
trodes were mounted on plastic insulators with a spacing of 1.5 mm inside aluminum housings having a
wall thickness of 1 mm and filled with argon to a pressure of ?1.5 atm. The chambers were used in the
current-pulse mode [8]. Calibration measurements with the double fission chambers were made in the
neutron beam from a crystal monochromator [9] yielding energies of 0.051-0.056 eV. The admixture of
neutrons with energies 0.193-0.235 eV in the spectrum of the diffraction beam was estimated by the authors
[9] to be 10% of the flux of neutrons in the main group with an average energy of 0.053 eV. The results
TABLE 1. Basic Experimental Characteristics
Parameter
Chamber 239pu/235u
Chamber 233U/235U
trionochromatori
beam
scandium
beam
monochromator
beam I
scandium
beam
Signal/background ratio
Signal counting rate, cts/sec
Ratio of chamber counts
Measured cross section ratio in "signal" spectrum
of scandium beam
Cross section tatio for energy range 1.5-2:3 keV
(2-keV peak)
? 33/48
? 38/17
2,245+0,011
0,578+0,014
0,562+0,015
0,7/1,9
0,6/0,7
0,791+0,012
? 60/53
? 30/13
2,355+0,028
1,41+0,04
1,41+0,04
?2,3/1,5
?1,7/0,5
3,460+0,052
Translated from Atomnaya Energiya, Vol. 38, No. 1, pp. 43-44, January, 1975. Original letter
submitted June 7, 1974,
0 1975 Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011. No part of this publication may be reproduced,
stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying, microfilming,
recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $15.00.
49
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Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050001-5
Fig. 1. Data for
of(239Pu) /af(23 5U)
range 1-3.5 keV.
?) [17]; - ? -
0) present work.
of(233U) /0-f(235 U) and
in the neutron energy
4) [4]; Or) [3]; C)) [51;
? -) [18]; - - -1 [16];
in the neutron beam of the monochromator were normalized to values of the ratios 0-f(239Pu)/af(235U) and
af(233U)/o-f(235U) which were 1.651 0.03 and 0.96 ? 0.02, respectively, and which were calculated from
evaluated data [10, 11] and from experiment [12-141.
RESULTS
Table 1 gives the main results of the experiment in the neutron beam from the crystal monochro-
mator and of the experiment in the scandium beam for a reactor thermal power of -11 MW at the Obninsk
nuclear power station. To obtain the fission cross-section ratios for the energy range 1.5-2.3 keV, a
correction was introduced for the nonmonoenergetic nature of the spectrum of the "signal" scandium neu-
tron beam in accordance with data of the BNAB system of constants [151. The mean-square error given in
Table 1 includes the statistical error of the measurements and the uncertainties in the calculated normal-
izing constants and corrections.
Figure 1 shows data for af(233U)/o-f(235U) and o-f(239Pu)/af(23U) in the neutron energy range 1-3.5
key. It is clear that the present data for o-f(233U)/7f(235U) agrees within the limits of error with the results
of spectrometric "direct" measurements with double fission chambers for the slowing-down time in lead
[3] and from a high-altitude nuclear explosion [4]. The data obtained for 01(2391:v)/Q1(235U) agrees satis-
factorily with experimental results [51 where af(239Pu) and af(235U) were measured "simultaneously" in a
lead cube and also with previous evaluations [16, 171. The values of the ratio af(239Pu)/o-f(235U) calculated
from the data in [51 and [61 for the spectrum of the 2-keV peak of the scandium beam are 0.556 and 0.713.
The first result is in good agreement with the data in Table 1.
In conclusion, the authors thank V. S. Golovkin for providing the opportunity to carry out the calibra-
tion measurements in the beam of the crystal monochromator and Yu. A. Kazanskil and S. P. Belov for
consideration of this work.
LITERATURE CITED
1. P. Greebler, B. Hutchins, and C. Cowan, in; Proceedings of the IAEA Symposium "Nuclear Data
for Reactors - 1970' Vol. I, June 15-19, Helsinki (1970), p. 17.
2. S. M. Zaritskii, M. N. Nikolaev, and M. F. Troyanov, Neutron Physics [in Russian], Vol. 1,
Naukova Dumka, Kiev (1972), p? 3.
3. W. Lehto, Nucl. Sci. and Engng., 39, 361 (1970).
4, R. Albert, Phys. Rev., 142, 718 (1966).
5. A. A. Bergman et al., Nuclear Constants [in Russian], No. 7, Izd. TsYaD, Obninsk (1971), p. 37.
6. Yu. V. Ryabov et al., At. Energ., 24, No. 4, 351 (1968).
7. E. N. Kuzin et al., At. Energ., 35, No. 6, 391 (1973).
8. Kononov et al., Prib, i Teich. Eksp., No. 6, 51 (1969).
9. V. S. Golovkin, V. N. Bykov, and V. A. Levdik, Zh. Eksp.Teor. Fiz., 48, No. 4, 1083 (1965).
10. Parker's Evaluated Data from the 1968 UK Nuclear Data Library, DFN-65, DFN-66.
11. B. Leonard et al., Evaluated Nuclear Data File of National Neutron Cross-Section Center, 239Pu
Data, Mat.-1104, April (1970).
12. R. Gwin et al., Nucl. Sol. and Engng., 45, 25 (1971).
50
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13. L. Weston et al., Nucl. Sci. and Engng., 42, 143 (1970).
14. R. Freemantle, Rep. AEEW-M-502 (1965).
15. L. P. Abagyan et al., Group Constants for Reactor Calculations [in Russian], Atornizdat, Moscow
(1964).
16. W. Davey, Nucl. Sei. and Engng., 26, 149 (1966).
17. W. Hart, Rep. AHSB(S)R-169 (1969).
18. T. Beyer, Atomic Energy Rev., 10, No. 4, 529 (1972).
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FAST NEUTRON SPECTROMETRY OF (a, n) REACTIONS
USING A DEUTERA TED SCINTILLA TOR
E. M. Burymov, S. P. Korsunova, UDC 539.125.5.164.07
and N. N. Spendiarov
A deuterated scintillator (octadeuteronaphthalene) was used in the work of Burymov et al. [1] for
neutron spectrometry in an investigation of inelastic scattering. Neutron spectra from (a, n) reactions in
the nuclei 9Be, 19B, and 11B were measured by the same method in the present work.
The experiment was performed at the 120-cm cyclotron of the Research Institute for Nuclear Physics,
Moscow State University with 23- and 25-MeV a-particles. Target thicknesses were 1-2 mg/cm2. A crys-
tal of octadeuteronaphthalene 30 mm in diameter and 20 mm high at 23 cm from a target was used in con-
junction with an FEU-29 photomultiplier for neutron spectrometry. The spectrometer was calibrated with
14.1-MeV neutrons.
300
200
100
0
100
0
15
70
Ed, MeV
15
20
Ed, MeV
100
50
80
60
40
20
0
15 20 Ed, MeV
15
20 Ed, MeV
E,-23Me
95-40?
15 20 Ed, MeV
60
40
20
0
Fig. 1. Pulse-height spectra of scintillator recoil deuterons
from elastic scattering of neutrons produced in the reactions:
a) 9Be(a, n)12C; b)19B(a, n)13N; c) itwa 014?
DI [Ed) energy of
recoil deuterons; N) number of counts per analyzer channel;
arrows indicate the position of the peak in the recoil deuteron
spectrum corresponding to the energy of emitted neutrons when
the final nucleus remains in the ground state; En (9/8)Ed].
Translated from Atomnaya Energlya, Vol. 38, No. 1, p. 45, January, 1975. Original letter sub-
mitted March 15, 1974.
0 1975 Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011. No part of this publication may be reproduced,
stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying, microfilming,
recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $15.00.
52
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Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050001-5
Neutron spectra are shown in Fig. 1 for several angles of emission. The time of measurement for
each spectrum was ?10 min at a beam current of ?0.1 A.
Since a single-crystal arrangement was used, great difficulties arose because of the y -background.
It was particularly great for signal pulses smaller than pulses corresponding to neutrons with energies of
18-20 MeV (this energy decreases somewhat as the angle increases). In those cases where the neutron
energy exceeds this value, peaks are observed in the experimental spectra corresponding to the ground
state in the final nucleus (see Fig. 1) and to the first excited state in particular cases. This is observed
in all the reactions studied at angles less than 90?. In the 12C(oz, n)150 reaction, the maximum neutron
energy is 13.5 MeV for Ea = 25 MeV andtherefore peaks cannot be observed.
Thus we have demonstrated the possibility of using a single-crystal deuterated neutron spectrometer
for studying (a, n) reactions in cases where the neutron energy is sufficiently high (greater than 18-20
MeV).
LITERATURE CITED
1. B. A. Benetskii, E. M. Burymov, and I. M. Frank, Yad. Fiz., 8, No. 5, 920 (1968).
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INITIAL STATIC FOCUSING IN SMALL LINEAR
TRAVELING-WAVE ACCELERATORS
A. D. Vlasov UDC 621.384,64.01
In a traveling-wave linear proton accelerator, an effect occurs which is similar to static electric
focusing in cyclotrons [1] and in linear accelerators with drift tubes [2, 3]. A momentum directed towards
the axis acts on a particle upon entrance into the accelerating wave and a momentum oppositely directed
upon leaving the wave. In a low-energy accelerator for acceleration of protons from 50-100 keV to several
MeV, the focusing action of the initial momentum can be sufficient to allow one to get along without special
focusing devices.
The longitudinal electric component of the accelerating wave has the form
E, cos ( --v? ?o)t = Em cos cp,
dz
where z, go, and v are the longitudinal coordinate, phase, and velocity of a particle; t is time; co = 2A-c/X
(here, X is wavelength and c is the velocity of light).
We assume the wave amplitude Em and the equilibrium phase (pp are constant along the accelerator,
and we confine our considerations to the equilibrium particle. The longitudinal motion of this particle
occurs at a constant acceleration eEm cos (pp/mo and is described by the equations
(1)
eEm COS (pp W052
Z? t0 ?
2m0 2eEm COS (Fp
Here e, mo, and Wo are the change, mass, andrest energy of the particle; 0 = vp/c; vp is the equilibrium
velocity. If, for example, Em. = 3 Mv/m, cOp = 30? and the protons are accelerated from 100 keV (f3i
= 0.0146), zi = 38.5 mm. For a final proton energy of 750 keV (pf = 0.0400) and 5 MeV (Of = 0.1028), the
accelerator length, Zf?Zi, is 0.25 and 1.86 m, respectively. If X = 2 m, the duration of particle accelera-
tion, c(tf?ti) /X will be 4.6 and 16 cycles of the hf field. For X = 24 m, we obtain only 0.4 and 1.3 cycles,
respectively.
The number of phase oscillations is given by
2Wo sin (pp
n.43.? 1/Ff? 1/N) ?
aeEmk cos2 (pp
For the parameters given above and X = 2 m, it becomes 0.5 and 1.62, and for X = 24 m we obtain
the values 0.2 and 0.5.
The transverse momentum acting upon a particle when it enters or leaves the field of the wave (Eq.
(1)) is calculated by integration of the equation of transverse motion
d2x dzx
oP2 ?
m? dt2 dZ2
ex aE.
2 az ?
For the equilibrium particle, these momenta are
dx ex, , eET,icos cp x
/ ?1L ?
21'il di- 2ft0f2
Translated from Atomnaya. Energiya, Vol. 38, No. 1, pp. 46-47, January, 1975. Original letter
submitted June 6, 1974.
0 1975 Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011. No part of this publication may be reproduced,
stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying, microfilming,
recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $15.00.
54
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Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050001-5
and their effect is given by the matrices
where
? 1 . 0 \ f _
i,f
The transverse motion of the equilibrium particle in the field (1) is described by
ci2x , sneEm sin cpp x
nEm sin (pp x 60 11 .1 x
Eix= p , E0x= R2
(2)
are the transverse components of the accelerating wave and the intrinsic Coulomb field of the accelerated
beam (I is beam current; R is beam radius) is
8
60 = 0
Elx ztEmR2 sin (pp
We obtainE '5- 1.1 for Em = 3 MV/m, cpp = 300, X = 24 m, and I/R2 5- 33 mA/cm2. The expression
for E 2x corresponds to a cylindrical beam. The radius of the beam varies and the beam breaks up into
bunches in proportion to acceleration. Both these factors lead to a variation in E. For simplicity, we
assume below E = Emax = const.
Expressing 8 through t, we bring Eq. (2) into the form
d2x px
dt2 4t p=2?0.)tg cpP.
The solution of this equation and its derivative are given by
MI I (T) K (T) ,
, dx dt
x = tit dz = 4z
where T = )rpl. Setting T = Ti = 1./71 in these equations, we determine the constants of integration A and B
through xo and x = (dx/dz)o, and obtain a matrix equation connecting the instantaneous values x and x/
with the initial values,
Here,
( 27,)= (bb 21 b11b 2122) (r4o )
b11=---T [K0 (1 ? 1) 11(T)+ 10 (t) 1C (T);
4z .x
b12 ? ? [If 1 (1111 (t) ?It (ti) IC 1 (TA;
_ 1
b21? ? [K0 (xi) 1.0 (x) ? 10 (xi) Ko (T)1;
P.
-?
by, ? ? [K1 (TO /0 (T) it (T) Ko (T)1.
The matrix determinant bilb22?bi2b2i = i//3 differs from unity and decreases like 13-1. The variation
of x and xl over the length of the accelerator including the initial and final momenta is given by the matrix
product
x\ (I 0\ (b11 b12) (I 0 (all av2,1 \
/ 621 lin ,/ki.1/\xJ a
A. 21 an/
xi ? i? /
Multiplying the matrices, we find
? ait=xf [Ko (Ti) Ki (1i) /2 (ti) ? [10 (x0+ (xl) 1C1 (Tf) ;
T T
a12 (K1 (TO /I (Tf) ? it (TO K2 (TM;
Ti
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a21=1 ito (Ti)
?ripi [ Ki (Ti) ] L + if (rf) 1 [10 (To 4_ ii(Ti) 1 r,?0(tf) K (stf) ]) ;
Ti 't
a22=-Ti-L-4 {KJ (Tii)L 11-/i(g)[K, (Tf) K1 (1f) } ?
pf Tif Tf
The matrix elements a ij depend on the four quantities Th pi, pf, and zi, with the last depending
mainly on Em and (3 i since cos (pp 1. Thus for given Em, (31, and 13f the matrix elements a ij are deter-
mined by the values of Ti or, alternatively, by the quantity
EM? cos2 (Pp 4nW9
Ti ?
sm (pp
The optimal choice of Ti depends on the parameters of the injected beam and on the requirements
imposed on the beam at the accelerator exit. In particular, one can select Ti from the condition
Ko(ri)? Kt (xj)? 0.
?r ?
Then, neglecting terms containing Ko(Tf) and KI(Tf) in the expressions
all= [10 (TO+ 11 (Ti) TfiCi (TO;
T ?
1
21 , ?
a12=4? A 1 (T1) 'till (TO;
Ti
EIO
Ki (Tf) 1 ,
a21= 4z10 f (Ti) Ti [K? (Tf) T
for a12 and a 22,
we obtain
(3)
? a22= 113fi (Ti) (Tf)J,-- 111!tf) .
f -
The root of Eq. (4) is T =1.33, to which Ti = 6630 MV corresponds. A wavelength A = 23.6 m is
required according to Eq. (3) when Ey/ = 3 MV/m, e = 1.1, cpp = 300, and the initial proton energy is 100
keV. If the protons are accelerated to 750 keV and 5 MeV, we, respectively, obtain Tf = 2.2 and 3.5 and
au= 0.50; 0.16, a12=0.174; 0.925 m,
021= ?0.36; ?0,044 m-1, a22 = 0.61; 0.64.
These elements a ij are not too large and therefore the transverse deflections and trajectory slopes
of the particles at the accelerator exit will .have acceptable values. If A = 2 m, we obtain (pp = 3?14' for the
same EM, E, and 13 i. The values of Ti,f and a ij obtained in this case are not much different from those
obtained above. Thus, although the choice T = 1.33 may differ from the optimal choice, the initial static
focusing is still sufficiently effective.
Having selected the quantity TE, it is necessary to choose an optimal wavelength and equilibrium
phase, which are related through Eq. (3). To do this, it is necessary to know the spreads in exit energy,
deflection and trajectory slope of nonequilibrium particles corresponding to various A. and (pp. The motion
of nonequilibrium particles can only be calculated by numerical methods on a computer.
LITERATURE CITED
1. A. A. Kolomenskii and A. N. Lebedev, Theory of Cyclic Accelerators [in Russian], Fizmatgiz,
Moscow (1962).
2. A. D. Vlasov, Theory of Linear Accelerators [in Russian], Atomizdat, Moscow (1965).
3. I. M. Ka.pchinskii, Particle Dynamics in Linear Resonance Accelerators [in Russian], Atomizdat,
Moscow (1966).
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DEPENDENCE OF THE FISSION-FRAGMENT SPUTTERING
RATIO FOR THIN LAYERS OF A SUBSTANCE ON THE
MEAN ENERGY OF THE FRAGMENTS
B. M. Aleksandrov, N. V. Babadzhanyants,
I. A. Baranov, A. S. Krivokhatskii,
L. M. Krizhanskii, and V. V. Obnorskii
UDC 546.799
In [1], it was shown qualitatively that the sputtering ratio K for thin layers of 241Am oxide is larger,
the higher the mean energy of the fission fragments. We assume that the difference, obtained in [2-5], in
the measurements on thin films of UO2 (K 104 atoms/fragment), on thick, coarse-grained baked disks
of UO2 (K 9 atoms/fragment) and on metallic disks of plutonium and uranium (K io atoms /fragment),
respectively, can be partly associated with the difference in the energy spectra of the fission fragments,
since the latter originated inside layers of different thickness and crossed the sputtered surface at dif-
ferent angles. The problem in this paper is the quantitative determination of the dependence of the sput-
tering ratio for thin layers of a substance on the mean energy of the fragments. Studies of a dependence of
this kind allow one to ascertain those interactions of the fission fragments with the atoms of the substance
(elastic or inelastic) which cause the sputtering, to verify the assumption expressed concerning the source
of the spread in the experimental data on K in [2-51, and also have a definite practical value.
In this paper, we investigated the sputtering of fine-grained layers, prepared from 238PU by elec-
trolysis of an aqueous solution and representing hydrated plutonium oxide, desiccated in air at room tem-
perature. A thin layer of 2 52C f served as a source of the fragments. The plutonium layer undergoing
sputtering and the californium layer were mounted parallel to each other, 7 mm apart, in a holder and
placed in a special vacuum chamber. A rotating disk was placed between the layers. Ten thin nickel films,
transparent to the fission fragments, and two thick foils, nontransparent to the fragments, were placed
in openings along the rim of the disk. The fission fragments from the californium layer passed through the
two protective nickel films, then the nickel-film collector in the disk, after which they fell almost perpen-
dicularly on the plutonium layer, sputtering it. The sputtered atoms of plutonium were collected by the
nickel collector-film located 2 mm from the plutonium layer. After 2 h of exposure, the disk was reversed
(without any change in the high vacuum) and another collector with a different thickness of nickel film was
placed between the layers of plutonium and californium, whereas the first collector-film was located above
a surface-barrier, silicon a-counter. The a-spectrum of the collector was measured with the aid of an
AI-256 multichannel amplitude analyzer with a precision-amplitude generator. The number of plutonium
atoms collected by the collector was determined from the a-particle counting rate and the presence or
absence of contamination from the 252Cf was monitored in the spectrum. At the same time, a correction
for the self-sputtering of the plutonium layer due to a-decay was taken into account. After the first series
of measurements (sputtering in all of the 12 collectors and production of a-spectra with each one), two
more series of measurements were conducted without impairment of the high vacuum, after which the
layer being sputtered was replaced. In all, four plutonium layers with different sputtering ratios (500,
185, 130, and 70 atoms/fragment) for a maximum mean energy of the fragments ?87 MeV were subjected
to sputtering. With a decrease in the mean energy to 23-27 MeV, the ratios for these layers were reduced
to 100, 33, 40, and 20 atoms/fragment, respectively. The dependence of the sputtering ratio on the mean
energy of the fragments for a layer with a maximum sputtering ratio of 500 atoms/fragment is shown in
Fig. 1. The errors indicated represent the variances obtained as a result of three series of measurements.
The curves for other plutonium layers have the same character; however, one should mention a certain
Translated from Atomnaya Energiya, Vol. 38, No. 1, pp. 47-49, January, 1975. Original letter
submitted July 1, 1974.
? 1975 Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011. No part of this publication may be reproduced,
stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying, microfilming,
recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $15.00.
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200 600 1000 1400 1800
1, g/cm2
Fig. 1
Fig. 1. Dependence of the sputtering ratio of a thin layer of hydrated plutonium oxide on the mean
energy of the fission fragments for 232Cf (1 is the thickness of the nickel films).
Fig. 2. Change in specific ionization dn/dx and specific energy losses dE /clx along a track x of
244Cm fission fragments [6]: 1, 2) for light and heavy fragments, respectively; 3) averaged values
of curves 1 and 2.
2200 2600
3000
12 16 20 24 X,mm
Fig. 2
100
75
50
25
0
tendency towards a reduction in the slope of the curves for plutonium layers with a lower sputtering ratio.
The film thicknesses range from 300-2800 pg/cm2, and the energy spectra of the fragments after passing
through these films were determined with a semiconductor spectrometer before the start of the experiment.
Three energy spectra for the fission fragments, which characterize the energy distribution of the frag-
ments after passage through 300, 1600, and 2800 ptg/cm2 nickel films, are given in Fig. 1. With a de-
crease in the mean energy of the fragments from ?87 to ?25 MeV, the sputtering ratio is reduced by a
factor of 3.5-6 as a function of the quality of the layer undergoing sputtering. Thus, for fine-grained
layers of a nonconducting substance, the sputtering ratio depends substantially on the energy of the frag-
ments, if this energy >20 MeV. One can assume as completely valid that the difference in orders of mag-
nitude in the sputtering ratios of thin, fine-grained films [2] and thick, metallic disks [4, 51 could be as-
sociated, completely or to a significant degree, with the difference in the mean energy of the fission frag-
ments. On the other hand, the rate of change of the sputtering ratio with a change in the energy attained
En this experiment can probably not explain the large difference in the sputtering ratios of the fine-grained
layers and films [1, 2], and the coarse-grained disks [3]. This apparently says something about the dif-
ferent mechanisms for the sputtering of coarse-grained and fine-grained samples. From the data obtained,
it also follows that the sputtering of coarse-grained, nonconducting layers by fission fragments with ener-
gies from ?20 up to ?100 MeV occurs mainly because of inelastic interactions of the fission fragments
with the atoms of the substance. This is seen from the character of the curve in Fig. 1, which agrees
qualitatively with the experimentally obtained curve for dE /clx as a function of x (Fig, 2, [6]), By investi-
gating the question of the possible effect of the size of the grains on the value of the sputtering ratio, one
can attempt to utilize the same reasons which were applied in [7, 8] for explaining the formation of tracks
from the fission fragments in different substances. Thus, one can assume that reflection of phonon waves,
low-energy electrons, and excitons, which arise during the passage of a fragment through a grain, occurs
at the boundary of the grain. These phenomena must inhibit the dissipation of energy, evolved within the
grain, beyond the limits of its boundary. At the same time, the amount of energy evolved in the grain by
the fission fragments and the degree of dissipation of this energy beyond the extent of the grain will depend
on its size. In fact, tracks from the displaced material were observed only in fine-grained films; the
maximum sputtering ratio (104 atoms/fragment) was measured for fine-grained layers. Hardly noticeable
tracks of a discontinuous character were observed in coarse-grained, insulating materials; the sputtering
ratio for such materials proved to be a minimum (5-10 atoms/fragment). However, there is also a dis-
crepancy. Thus, no tracks are observed in metals, although the sputtering ratio for metals proved to be
significant (103 atoms/fragment). It is of interest to investigate the dependence of the sputtering ratio on
the mean energy of the fission fragments in an experiment similar to the one described here for metals
and a coarse-grained, nonconducting material.
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LITERATURE CITED
1. B. M. Aleksandrov, A. I. Baranov, A. S. Krivokhatskii, and G. A. Tutin, At. Energ., 33, No. 4,
821 (1972).
2. M. Rogers, J. Nucl. Mater., 15, No. 1, 65; 16, No. 3, 298 (1965).
3. G. Nilsson, J. Nucl. Mater., 20, No. 2, 215 (1966).
4. F. S. Lapteva and B. V. Ershler, At. Energ., No. 4, 63 (1956).
5. M. Rogers and J. Adam, J. Nucl. Mater., 6, No. 2, 182 (1962).
6. F. Nasyrov, At. Energ., 16, No. 5, 449 (19-64).
7. M. Goland, J. Appl. Phys., 35, No. 7, 2188 (1964).
8. K. Jzui, J. Phys. Soc. Japan, 20, No. 6, 915 (1965).
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SILVER CHLORIDE TRACK DETECTORS
N. P. Kocherov, N. R. Novikova, UDC 539,1,073:546,57,131
and N. A. Perfilov
Since 1960, work has been done in France and FRG on the recording of charged-particle tracks in
flat single crystals of AgCl. The principle of operation of such detectors is based on the process of photo-
lysis in silver halide crystals. When a charged particle passes through the crystal, a chain of defects is
formed along its trajectory. If the crystals exposed to particles are thereafter subjected to photolysis
under the action of actinic light (X 410 nm), metallic silver will be produced at the defects caused by the
particles, and the resulting tracks will be visible under the microscope. The tracks found in crystals
are very similar in appearance to tracks in emulsions.
In 1972, Schopper [11 established that the latent image of the particles in AgC1 crystals is stable
when the crystals are illuminated with yellow light at the instant of exposure, whereas when there is no
illumination at the instant of particle action, the image regresses rapidly. This makes it possible to turn
the sensitivity on and off whenever necessary for the experimenter.
In the present paper we report on our experience in the preparation of such detectors and on their
properties.
The original AgC1 salt was recrystallized by dissolving in ammonia and subsequently precipitating
it by means of HC1. The resulting reagent was spectroscopically pure with respect to metals of valence
2 or higher.* The single crystals were grown by letting the solution flow between quartz and glass plates
[21 and had dimensions of 15 x 10 x 0.15 mm. In principle the area of the crystals can easily be increased
by a factor of 2-2.5. Crystals grown from AgC1 with no additives do not record even fission fragments.
Crystals alloyed with 0.1% CdC12 by weight record the a-particles from the decay of 252Cf but are insen-
sitive to protons. Crystals alloyed with 0.5% CdC12 record protons up to an energy of 4 MeV.t The rate of
regression of the latent image when the AgC1 is exposed to particles without being illuminated depends on
the specific energy losses of the recorded particles: the images persist for minutes in the case of protons,
for hours in the case of a-particles, and for days in the case of 252Cf fission fragments. If the crystal
is illuminated with yellow light at the same time as it is exposed to particles, the latent image is stable
for several months [1].
Fig. 1. Photomicrograph of tracks in an
AgC1 crystal irradiated with 660-MeV pro-
tons.
AgC1 detectors are transparent crystals: they
can be stored for long periods and used under conditions
of illumination with long waves measuring 440 nm or
more. An image appears when the crystal is illuminated
with light having a wavelength of X 410 nm (for example,
5-8 min of exposure to direct sunlight filtered through
ZhS-10 and P8-15 filters). Fixing the image is not
necessary. In external form (see Fig. 1) the particle
*The recrystallization of the reagent was carried out
by P. I. Chaikin and 0. V. Kesarev at the Scientific-
Research Institute of the Jewel Industry.
tDetectors capable of recording protons with energies
up to 14 MeV are being produced today.
Translated from Atonanaya Energiya, Vol. 38, No. 1, pp. 49-50, January, 1975. Original letter
submitted July 1, 1974.
0 1975 Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011. No part of this publication may be reproduced,
stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying, microfilming,
recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $15.00.
60
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tracks in the crystals exhibit almost no differences from tracks obtained with nuclear photographic emul-
sions (31
Thus, the use of silver chloride detectors in the case of prolonged exposure (of the order of weeks
or longer) makes it possible to turn the detector sensitivity on and off whenever necessary. The particles
recorded in the crystal will then be those which passed daring the time of exposure to yellow light plus
the time of regression. In the case of short exposures (seconds) without illumination, the crystals can be
developed after various intervals of time following the exposure, which makes it possible to obtain a series
of detectors with different recording thresholds, since the regression time of the latent image depends on
the specific energy losses of the recorded particles. It is possible to conduct biological experiments with
objects mounted on the surface of the detector: development using light does not destroy the objects and
makes it possible to observe them together with the tracks of the particles impinging on them.
?
Silver chloride detectors make it possible to study nuclear reactions on silver and chlorine without
any background from reactions involving light nuclei, which interfere seriously with the use of nuclear
emulsions; they can also serve as fast-neutron dosimeters, using the (n, p) and (n, a) reactions on silver
and chlorine [41.
Since these detectors are simple to produce and have convenient properties, it may be supposed that
they will come into wide use in the near future.
LITERATURE CITED
1. B. Schopper et al., in; Proceedings of the Eighth Conference on Nuclear Photography and SSTD, Vol.
1, Institute of Atomic Physics, Bucharest (1972), p. 350.
2. A. M. Levitskaya and A. M. Korolev, Zh. Teich. Fiz, 7, 760 (1973).
3. N. A. Perfilov, N. R. Novikova, and E. I. Prokoffeva, At. Energ., 4, No. 1, 45 (1958).
4. G. Henig et al., in: Proceedings of the Eighth Conference on Nuclear Photographyand SSTD, Vol.
1, Institute of Atomic Physics, Bucharest (1972), p.384.
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CHANGE IN THE OPTICAL DENSITY OF PMMA UNDER THE
ACTION OF DEUTERONS WITH ENERGIES 4-150 keV
S. P. Kapchigashev, V. P. Kovalev, UDC 539.125.4:541.15
E. S. Barkhatov, and V. A. Sokolov
The measurement of the difference in the optical density of polymethyl methacrylate (PMMA) before
and after irradiation is a widespread method of determination of the absorbed energy of high-intensity
ionizing radiation. However, the principles of the variation of the optical density (AS) as a function of the
characteristics of the radiation (LET, type of particles, eta.) have thus far been insufficiently studied
[1-4].
In this work we investigated the change in the optical density of PMMA under the action of deuterons
with an energy of 4-150 key. The method of irradiation was described earlier [5]. Plates of PMMA 1 mm
thick were irradiated with deuterons of various energies and the change in the optical density observed on
an USV-1 spectrophotometer at the wavelength 300 nm. A comparison of the dependence of the effective-
ness of radiation chemical changes on the energy was conducted with an integral flux equal to 4.3 -1014
deuterons/cm2. It was preliminarily shown that in this region of fluxes AS is a linear function of the dose.
The ranges of protons of the investigated energies are hundredths of a micron to units of microns,
which is substantially less than the thickness of the PMMA plates. Therefore, the absorbed energy is
proportional to the total energy of the particles E. The effectiveness of optical changes in n can be de-
termined as n= AS/E. The values of n for various deuteron energies are presented in Table 1.
From Table 1 it is evident that the effectiveness of the optical changes decreases substantially with
increasing deuteron energy, which does not agree with the conclusions drawn in [3] with respect to low-
energy particles. The increase in the effectiveness of radiation changes with decreasing dose of deuterons
may be associated with a relative increase in the contribution of elastic nuclear collisions to the "ex-
change" of energy. Here it must be assumed that elastic processes are significantly (more than three
times) more effective in radiation chemical disruptions of polymethyl methacrylate than processes of
ionization and excitation. This phenomenon must be taken into consideration in the practical use of poly-
mer materials in the dosimetry of nuclear radiation.
TABLE 1. Dependence of the Effectiveness of Optical
Changes in PMMA on the Deuteron Energy
E, key
71.cm2/MeV
E,keV
ibemz/MeV
4
11+0,5
50
3,4+0,2
10
6,6+1,0
100
3,9+0,2
25
3,7+0,6
150
3,6
LITERATURE CITED
1. Ya. I. Lavrentovich et al., in: Dosimetry and Radiation Processes in Dosimetric Systems [in Rus-
sian], Fan, Tashkent (1972), p. 178.
2. K. Chadwick, Rad. Res., 44, 282 (1970).
Translated from Atomnaya Energiya, Vol. 38, No. 1, p. 50, January, 1975. Original letter sub-
mitted July 1, 1974.
? 1975 Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011. No part of this publication may be reproduced,
stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying, microfilming,
recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $15.00.
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3. G. Angstrom and L. Ehrenberg, in: Collection of Materials on the Symposium on Individual Prob-
lems of Dosimetry [Russian translation], Atomizdat, Moscow (1962), p. 188.
4. D. Neufeld and W. Snyder, ibid., p. 33.
5. S. P. Kanchigashev et al., At. gnerg., 34, No. 4, 299 (1973).
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COME CON NEWS
PROVISIONAL INTERNATIONAL WORKTEAM
P. M. Tyukhtin
On February 23, 1973, authorized representatives of seven countries with membership in the Council
for Economic Mutual Aid [COMECON] (viz. Bulgaria, Hungary, GDR, Poland, Rumania, USSR, Czecho-
slovakia) signed an Agreement on setting up a Provisional International Scientific-Research Workteam to
carry out reactor physics research on a VVER type critical assembly.
The purpose of this research will be to devise and work out exact methods for design calculations
relevant to VVER type reactors, and to elaborate straightforward and reliable techniques for handling
variant calculations in the process of designing and operating reactors, and also to expedite research on
perfecting methods in measurements and in the design of measuring equipment.
The experimental research facilities available to the staff of this provisional international workteam
will be a VVER type critical assembly (3R-6 assembly) located at the Central Physics Research Institute
of the Hungarian Academy of Sciences. This critical assembly was started up in November, 1972. That
was followed by a period during which technical staff personnel were trained on the job, and in February,
1973, the scientific staff of the provisional international research team got started on the experimental
research program.
A Scientific Council staffed by no more than three representatives from each country has been set
up to provide scientific guidance for the activities of the research team, but each nation signatory to the
Agreement has one vote. The Scientific Council schedules its sessions to gear into the work plan of the
Council, but no more often than three times a year. These sessions are held in Budapest. Four sessions
have been held to date: the first in February, 1972, the second in December, 1972, the third in June,
1973, and the fourth in March, 1974; a fifth session is scheduled for January, 1975.
The head of the research team is the director of the Central Physics Research Institute, L. Mal.
A detailed analysis of results achieved and of scientific topics being tackled is presented at meet-
ings of specialists organized every eight months. To date two such conferences have been held, one in
Czechoslovakia on the statics of VVER type reactors (October 31 through November 3, 1972), the second
in Poland and dealing with reactor neutron statics (October 2-5, 1973). A third conference on VVgR
physics is planned for December 2-7, 1974, in GDR.
Three subteams have been set up to expedite the solution of some concrete problems. In the period
intervening between sessions of the Scientific Council, these subteams and the research personnel working
on the 3R-6 assembly are responsible for organizing their own work.
Translated from Atomnaya Energiya, Vol. 38, No. 1, p. 51, January, 1975.
? 1975 Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011. No part of this publication may be reproduced,
stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying, microfilming,
recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $15.00.
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COLLABORATION DAYBOOK
A conference of specialists of COMECON member-nations on standardization of radiation equipment
and radiation protection equipment was held in Mamaia (Rumania) September 3-6, 1974.
Drafts of COMECON standards applicable to modular lead shielding units were discussed and agreed
upon, as well as draft recommendations on standardization referable to shielded dry boxes (types, basic
parameters and dimensions), shielded gloveboxes (types, basic parameters and dimensions); y-therap'eutic
equipment (basic parameters and general technical specifications); complex mobile laboratories for non-
destructive testing and quality control work (general technical requirements); systems and devices for
alarm annunciation and flaw detection signalling in radiation flaw detection work.
Technical reports on recommendations referable to shielded cabinets or safes, packing sets for
shipping radioactive materials safely, were heard and discussed. Draft recommendations on proper clas-
sification of laboratories for handling sources of ionizing radiations, and general technical requirements
applicable, were discussed; a draft glossary of terms and definitions in the field of radiation protection
work was discussed and agreed upon, as were a work plan on standardization of radiation equipment and
radiation protection equipment for the year 1975, and a draft plan on collaboration over the 1976-1980
period.
The 14th session of the panel on nuclear electric power generating stations of the COMECON Per-
manent Commission on Electric Power was held September 17-21, 1974, at Herceg Novi (Yugoslavia).
Urgent problems in the development of nuclear power in COMECON member-nations and in Yugoslavia
came under discussion at the session. Materials on "Equipment type related engineering safety valida-
tions for installation and operation of nuclear power station equipment at the commissioning stage" and
on "Standard types of operating report forms for nuclear power stations with VVER-440 power generating
units" were approved. These documents enable those countries to go ahead with preparations on a high
technical level to put nuclear power stations into service and to raise the safety levels of those power
stations. Close attention was given to the development of power station equipment for nuclear power
stations with VVER -1000 type reactors, in particular for expanding cooperation between COMECON mem-
ber-nations in that area. A report entitled "Training of operating personnel for nuclear power stations,"
approved by the panel, is of considerable practical interest. The use of this material will be helpful in
significantly improving the training of highly skilled cadres for service in COMECON member-nations,
The 1975 work plan of the panel, envisaging research on such pressing topics as protection of the environ-
ment, nuclear power station safety, and improvements in nuclear power equipment, was approved.
The fourth session of the KNTS-RB (Scientific?technical coordinational council on radiation safety)
was held in Leipzig (GDR), September 24-27, 1974. The council heard information on how the resolutions
adopted at the )0CVIIth Session of COMECON on environmental conservation were being carried out, re-
ports and communications on work being done under the collaboration program, including recommendations
aiding physical design calculations of biological shielding for power stations with pressurized-water reac-
tors; results of a comparison of personnel monitors and dosimetric systems; suggestions aimed at further
Translated from Atomnaya Energiya, Vol. 38, No. 1, pp. 51-52, January, 1975.
? 1975 Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011. No part of this publication may be reproduced,
stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying, microfilming,
recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $15.00.
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development of the topic "Determination of volume of monitoring operations in radioactive wastes burial
areas"; requirements on radiation safety when sources of ionizing radiations are used in medicine, and on
general radiation safety criteria when radioactive isotopes are used in industry, and so forth.
Proposals on the holding of a conference of specialists from COMECON member-nations, sometime
in 1975 in Czechoslovakia, to deal with radiation safety problems in the context of operation of nuclear
power stations and within the framework of a program of international intercalibration of whole-body y -
radiation spectrometers, were discussed at this session. A draft drawn up by the IAEA leadership, en-
titled "Estimates of collectivepopulation exposure dose," was discussed, and remarks were prepared
for submission to the IAEA Secretariat, with special attention given to the most important measures to
be taken by IAEA in 1975 in the area of radiation safety from the standpoint of the interests of KNTS-RB
in this matter.
The council reached agreement on a tentative agenda for the fifth session of KNTS-RB, and a draft
work plan covering the 1975-1976 period.
A conference of specialists on ways of developing techniques and equipment for cleanup of radioactive
aerosols and gases from air-vented wastes was held September 30-October 4, 1974, in Leningrad.
Results of research work on methods of trapping radioactive iodine by trapping from the vapor phase,
and methods for analyzing radioiodine content, were discussed. The use of adsorbers packed with acti-
vated charcoal and presenting not less than 1000 m2/g surface area and 0.4 m bed height, with a linear gas
stream flow velocity of 0.3 to 0.4 m/sec in the adsorber, is recommended as a suitable means of trapping
iodine. It was pointed out that active charcoal cannot be used if nitrogen oxides, ozone, fluorine, or other
powerful oxidizing agents are present in the off-gases, and the same applies to admixtures of materials
of high molecular weight.
Specialists reported on research findings in studies of methods for determining the total radioactive
iodine content in air and in off-gases vented from nuclear power stations, the iodine content by components,
and the content of individual iodine compounds. The use of special cartridges (columns) filled with acti-
vated charcoal such as used in gas chromatography was recommended for periodic determinations in the
absence of nitrogen oxides or ozone. Continued research on developing distinct methods for analysis of
iodine and alkyl iodides was deemed advisable, as well as further work on methods for testing the filtering
capacity of materials employed. Comparative tests on specimens of filtering materials, using methods
generally applied in the participating countries, were proposed.
Attention was focused on how to remove radioactive noble gases and aerosols from off-gases vented
from nuclear power stations, on how to design filtering devices, and also on how to work out a unified
terminology referable to radioactive aerosols.
Reports were heard from specialists on the results of work in studying regular patterns in the dis-
semination of radioactive materials exhausted from the stacks of nuclear power stations. Concrete data on
isotope make-up, on the concentrations and physicochemical characteristics of radioactive products, and
also on models of the propagation of radioactive materials under conditions typical of short-term and con-
tinuous discharges were discussed, with attention given to special features of the terrain and environment,
wind conditions and variation, and the influence of those factors on scattering and propagation of radioactive
pollutants through the ground layer of the atmosphere.
Elaboration of a unified procedure, common to all COMECON member-nations, for the tolerance
level of radioactive products in the neighborhood of industrial plants and power plants, was proposed.
Long-term trends in the field of deactivation of radioactive aerosols and gases were also discussed.
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The seventh session of the KNTS on radioactive wastes and wastes deactivation was held in Lenin-
grad, September 30 through October 4, 1974.
Proposals sumitted by the delegations of the various countries dealing with research in the field
of deactivation of liquid wastes, solid wastes, and gaseous wastes, and deactivation of contaminated
surfaces in rooms and on equipment, came wider discussion. This research is to be conducted over the
1976-1980 period. These proposals were recommended as a basis in the preparation of a draft program
for collaborative efforts in this area. The proposals envisage an extended range of efforts associated with
processing and burial of radioactive wastes, including:
processing of liquid radioactive wastes of low and medium levels of radioactivity;
improvements in techniques and equipment for processing, immobilization, and burial of radio-
active wastes of all levels;
improvements in methods and equipment for averting radioactive contamination and spills, and for
deactivating rooms and equipment in nuclear power stations, in research centers, and in nuclear
fuel processing and reprocessing plants;
investigations of techniques and equipment for cleanup of air discharges to get rid of radioactive
aerosols and gases;
investigations of the Danube River basin and of the Baltic Sea for detection and assessment of pollu-
tion by radioactive materials.
The results of developments in the technology of disposal of low-level and medium-level solid and
immobilized wastes by burial in salt mine caverns, and also problems in shipping such wastes from the
point of formation to the burial site, were discussed (in reports covering developmental work in this area
by the GDR delegation); also results of research on physicochemical processes involving interaction of
radioactive wastes and materials of underground strata, and the development of a method for preparing
wastes for subterranean burial; data on research on applications of electrodialysis to cleanup of low-level
radioactive wastes (research and development work by the USSR delegation). Resolutions were adopted
on the appropriate use of these developments in the particular countries, and on recommendations relevant
to the more advanced research trends in these fields.
The council discussed a reference catalog of ion-exchange resins produced in COMECON member-
nations and used in technological processes for cleanup and deactivation of liquid radioactive wastes. The
catalog provides the characteristics of ion-exchange resins, defined according to unified evaluation pro-
cedures adopted in the various countries.
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INFORMATION: CONFERENCES AND MEETINGS
THE NINTH WORLD ENERGY CONFERENCE
Yu. I. Koryakin
The ninth world energy conference was held September 23-27, 1974, Detroit (USA). Participating
in the deliberations were over 4300 delegates from 80 countries, and from virtually all the major inter-
national organizations concerned with power problems [COMECON (Council for Mutual Economic Aid),
the European Economic Commission of UNO, the International Institute of Applied Systems Analysis,
IAEA, the European Economic Commonwealth, etc.]. The Soviet delegation was headed up by the Minister
of Power and Electrification P. S. Neporozhnii. This conference also functioned as an anniversary
gathering: the year 1974 marked 50 years of the existence and activities of the World Power Organization.
The US President G. Ford delivered an address inaugurating the conference, with the presentation of a
World Comprehensive Energy Program entitled Interdependence Project.
It is worth pointing out that one special feature of the conference was its timing following immediately
upon the heels of the 1973-1974 energy crisis which broke out in the western countries and which affected
international relations and adversely affected the economic situation in a number of countries. The con-
ference consequently went beyond the framework of a simple forum for discussion of scientific and applied
problems in the field of industrial power, power resources, and power utilization. In the discussions,
reports, and floor discussion following presentations, allusion was frequently made to problems in the
area of philosophy, politics, sociology, psychology, international collaboration, and so on.
The slogan of the conference was "Economics and the environment in the light of future energy
needs," itself a reflection of the topics around which 229 papers presented centered, as well as in char-
acter with the discussion unfolding and the classification of the reports. All of the reports were grouped
under the following headings and panels:*
I Division. Population and Energy Resources.
Panel 1. Population growth and population distribution (5).
Panel 2. Power resources and power utilization (33).
Panel 3. Distribution of resources (14).
II Division. The Environment and Energy Availability.
Panel 1. Quality of the air (12).
Panel 2. Quality of the water (12).
Panel 3. Utilization of land resources (5).
Panel 4. Noise level (4).
Panel 5. Esthetics (2).
Panel 6. How the environment influences energy availability (20).
III Division. Restoration of Energy Resources.
Panel 1. Progress in the technology of energy resources recovery (20).
Panel 2. Effect of resources recovery technology on the environment (8).
IV Division. Energy Transformation.
Panel 1. Progress in energy conversion technology (22).
Panel 2. Effect of energy conversion on the environment (24).
*The number of reports submitted and presented on a particular set of topics appears in parentheses.
Translated from Atomnaya Energiya, Vol. 38, No. 1, pp. 53-56, January, 1975.
0 1975 Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011. No part of this publication may be reproduced,
stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying, microfilming,
recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $15.00.
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V Division. Energy Transport.
Panel 1. Progress in energy transport technology (18).
Panel 2. Effect of energy transport technology on the environment (3).
VI Division. Utilization of Energy.
Panel 1. Progress in energy utilization (21).
Panel 2. Effect of energy utilization on the environment (6).
There were 26 papers dealing specifically with nuclear power topics, and of these 23 dealt with reactor
electric power problems and the remaining three with applications of reactor heat in desalination of sea-
water, a possible role for nuclear power in the utilization of ocean resources and thermonuclear energy.
Most of the papers on nuclear power fall under the heading of "Energy conversion" in the breakdown above.
In practically all of the reports dealing with plans and forecasts for the development of power in the
various countries, a decisive role was cut out for nuclear power. In the discussion of topics pertinent to
nuclear power, attention was focused on macroscopic aspects of nuclear energy production: the scale of
development and forecasting patterns, the fuel problem, and the fuel cycle in nuclear power systems. In
contrast to preceding conferences of this series, here there was no place for descriptions of reactor types
and nuclear power stations, reactor engineering or reactor technology. Attention was instead centered not
so much on nuclear power stations per se as a means of generating electric power as on those factors and
conditions governing the functioning of nuclear power stations (uranium resources and uranium mining,
fabrication of fuel elements) or those factors and conditions accompanying the functioning of nuclear power
plants (disposal of radioactive wastes by burial, nuclear power station safety topics, environmental effects
of nuclear power stations). These topics also came under discussion in relation to the entire nuclear
power grid, not just in relation to individual isolated nuclear power stations. In general, it should be
stressed here that the systems approach applied in the discussion of energy problems typified this par-
ticular conference.
Reports dealing with nuclear power topics can be grouped arbitrarily under the following headings:
1. Scales and forecasts of nuclear power development.
2. The role of fast reactors and nuclear power structure in the future development of nuclear power,
and alternative structures for nuclear power station systems.
3. Experience in the building and operation of nuclear power stations, new types of nuclear power
stations and new projects and concepts.
4. Topics relating to the external fuel cycle in the nuclear power industry.
5. Nuclear power and the environment.
The role played by nuclear power in making energy available throughout the world in the future
(figured or extrapolated to the year 2000 as a rule) was the subject of papers presented by national dele-
gations and by international organizations. Various authors advanced predictions of the scale of nuclear
power development throughout the world, with a breakdown by large-scale geographic zones: Europe,
North America, the Pacific zone, Central and South America, Asia and the Far East and New Zealand,
Africa, and the Middle East, countries with a centralized planned economy. Data on some individual coun-
tries were also presented.
It was emphasized that the considerable rise in worldwide prices for crude oil, from the area of
21.2-23.8 dollars /ton to ?133 dollars /ton, brought on by the 1973 energy crisis, greatly enhanced the
economic competitiveness of nuclear power stations. This found expression in the economic feasibility of
using more expensive uranium (22 dollars/kg U308), in an appreciable drop in the economically justifiable
unit power output of power generating units of nuclear power stations, and consequently an expansion of
the demand for nuclear power stations on the world market, and also in the greater amount of interest
displayed in building major nuclear power generating stations. All of this, as might be expected, will
bring about an intensified tempo of development of nuclear power, and will lead to nuclear power making
a significant contribution to the generation of electric power around the year 2000. On the whole, the
fraction accounted for by nuclear power stations in total electric power generating capacity by the year
2000 will be anywhere from 50 to 83%, depending on the authors making the estimates. The last percentage
figure, corresponding to a total power output of 5.3 billion kW, is a limiting figure, and is apparently
on the high side. The range of 2.7 to 3.0 billion kW is judged more realistic.
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Fig. 1. Detroit. Cobo Hall (sports and exhibition complex housing
the World Energy Conference) appears in foreground.
Almost all of the authors pointed out the role that fast reactors can play in the future economy of
uranium resources. The development of fast reactors continues to be regarded as a general developmental
trend in nuclear power. But for the time being the inadequate level of technical preparation for widespread
commercial exploitation continues to lend a low profile to fast reactors in the forecast nuclear power
structure (accounting for no more than 30% of the total) at least till the end of the century. In other words,
until the end of the century nuclear power will remain predominantly ?thermal," and no appreciable savings
in uranium are expected from the use of fast breeders up to the year 2000 or thereabouts. Meanwhile, the
cost of building such reactors is simply exorbitant at the present state of the art. For example, the build-
ing of the fast liquid-metal LMFBR reactor with its 350 MW(e) rating, at Oak Ridge (USA), is now esti-
mated at a total of 1.736 billion dollars.
It was also emphasized in this context that the economic competitiveness of nuclear power stations
with LWR type reactors holds its own even in the face of a doubling of uranium prices, and possibly even
in the event that uranium prices increase three- or fourfold. There are on that account no special worries
as to the long-term viability of LWR type thermal reactors, even looking at them outside of the question
of fast breeders. But if effective fast breeders become a serious factor in the nuclear power picture within
the next 10 to 20 yr, then LWR type reactors can be used to process plutonium for those reactors with
uranium demand running into the millions of tons. This does not amount to a significant portion of the
estimated worldwide uranium resources. It is felt that LWR type reactors can be converted to converter
operation. For example, a reduction in burnup from 34,000 to 15,000 MWd/ton U in those reactors would
mean a 50% increase in plutonium yield.
Reports submitted by French and Canadian specialists were an exception. The former felt that in-
dustrial utilization of fast breeders by the turn of the century would be held back by a shortage of plutonium
and the rather long doubling time of the first breeders to appear on the scene (15 yr doubling time on oxide
fuel). Nevertheless, a doubling time of 10 to 15 yr is judged adequate, and there is no need to switch to
some other type of fuel (say, gaseous fuel).
The Canadian specialists are looking into CANDU type heavy-water converter-reactors as a possible
alternative to fast breeders. Even though the nuclear power systems of other countries hold open a major
role for fast breeders, these authors are of the opinion that the very considerable reserves of uranium and
thorium existing in the Earth's land mass and oceans (respectively, 2.5.1012 and 8.1 012 tons) make it
possible, in principle, to satisfy any anticipated needs for power for hundreds of years through reliance on
thermal reactors.
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We should take note of the growing interest in HTR type high-temperature reactors, even though their
place in the nuclear energy picture as of the year 1990 is not expected to be greater than 3%, and they are
not expected to account for more than 7% by the year 2000.
We also note a significant increase in the specific costs of nuclear power stations. As an example,
the average specific cost of nuclear power stations with LWR type reactors in the USA rose from 134
dollars/kW in 1967 to 397 dollars /kW in 1973. This increase reflects a greater realism in cost esti-
mates, but also stems from several other factors: escalation of prices for equipment and materials, an
increase in labor construction costs, inflation in general, and changes in the percentage of capital costs
during construction and delays in construction amounting to several years' time, tightening of safety and
reliability requirements imposed on nuclear power generating stations, greater complications and higher
costs in licensing and authorization procedures, etc.
The construction time for nuclear power stations has been stretched to 8-9 yr in the USA. The
assumption is that the cost of nuclear power stations will go on climbing in the future, principally on
account of additional safety requirements. It is also assumed that the cost of nuclear power stations cannot
be determined exactly, since a matter of 9 to 10 yr from the time the decision is taken to begin construc-
tion till the plant is ready for commercial operation. During that period of time the cost picture can under-
go far-reaching changes, as can public opinion, positions taken by leading organizations and governments,
labor costs, and labor productivity.
At the present time, the unit power output of LWR type reactors in the USA is limited to 1300 MW,
even though the existing technology even now makes it possible to raise the power output level to 2000 MW.
It is still unclear just what factors will act to limit further increases in unit power output (well above 2000
MW). It is assumed that savings in specific costs through raising the unit power level will be partially
lost because of the rise in construction costs due to the factors alluded to.
According to the view entertained by USA specialists, no new reactor types of any kind will be re-
quired for a two-component nuclear power system (LWR + FBR), since this system meets the require-
ments imposed on nuclear power, and exhibits significant technical and economic potentials.
The new nuclear power station projects presented for discussion at the Ninth World Energy Con-
ference are characterized by new construction methods (floating nuclear power stations and underground
nuclear power stations). The former variant is important in the case of the USA. Mass construction of
floating nuclear power stations on the continental shelf would be preferable in terms of the difficulties
involved in selecting and acquiring construction sites, supplying cooling water, and long-term construction
of ground-level nuclear power stations. Moreover, in the USA over half of the population lives in shore-
line areas. The first such nuclear power station with two PWR reactors each of 1150 MW(e) ratings is sited
three miles from the shoreline (at a depth of 12 to 22 m) near New Jersey, on an anchored platform behind
an artificial protecting breakwater barrier. The capacity of the Jacksonville wharf for the construction of
such nuclear power stations will be brought up to a level of four floating nuclear power stations per year
by the beginning of the 1980s. There is some hope that expensive protecting breakwaters will not be needed
eventually.
Underground construction of combined nuclear and fossil-fuel electric power generating stations was
described in a report by Swedish specialists. The need and the economic prerequisites for nuclear genera-
tion of space heat calls for shortening the length of space heat pipelines, which is in fact possible when
safety conditions are properly observed in underground construction of power stations.
Close attention was given, at the conference, to the fuel resources available to nuclear power and to
satisfaction of nuclear fuel needs. Over the last 10 years substantial progress has been observed in uran-
ium prospecting techniques. The average borehole drilling depth has been increased from 61 m in the early
19605 to 153 m in 1973, while drilling is being carried out at depths of 610 to 1200 m in a significant num-
ber of cases. At the same time, the uranium content in the occurrences brought to light decreased from
14.87 kg/m3 (before the 1960s) to 4.46 kg/m3 in more recent years, which in turn results in increased
uranium mining and recovery costs. If further uranium finds are possible at existing costs of 2.2 dollars
for each kilogram of U308, capital investments in uranium prospecting to ensure that uranium supplies will
be forthcoming during the service life of 1000 MW(e) nuclear power stations with LWR reactors will run
below 10 million dollars, which is not more than 2% of capital investments in nuclear power stations as
such. Consequently, an inevitable rise in uranium prices can be managed with, and the ultimate effect on
nuclear power costs will not be appreciable. Worldwide nuclear industry needs for uranium are estimated
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at 5.2 million tons up to the year 2000, while potential uranium reserves recoverable at economically
acceptable costs (up to 66 dollars/kg) are estimated at ?6.2 million tons. On the whole, the fuel problem
confronting the nuclear power industry seemed fairly bright in the light of the conference discussion.
As for one other major link in the external fuel chain, regeneration of spent nuclear fuel, here we
have to take note of various economic factors that are beginning to receive their share of attention and
which are stimulating international integration in this area. A capacity of 5 to 6 tons /day (or 1500 to 1800
tons annually) is deemed economically optimum for a nuclear spent fuel reprocessing plant. Such a plant
should be responsible for reprocessing fuel from nuclear power stations with LWR reactors of total output
placed at 50 to 60 million kW(e). But this is an expensive and cumbersome process. That accounts for the
joining of Great Britain, France, and FRG to cope with common "fuel recovery interests."
IAEA General Manager Z. Eklund, taking the floor at the conference, reported that international
integration of the external fuel cycle will be the subject of a forthcoming IAEA-sponsored conference
scheduled for 1977.
Radiation safety problems associated with nuclear power development are viewed as reliably re-
solved according to the conference proceedings. As for disposal of considerable quantities of radioactive
wastes by underground burial, this confidence is based on over 20 yr of positive experience in wastes
storage and on the remarkable progress achieved in recent years in wastes processing technology (im-
mobilization of aqueous solutions, particularly).
Several measures were carried out simultaneously at the conference: plenary and panel sessions,
seminars, lectures, round-table discussions, press conferences, technical excursions, and film showings
were held. In line with the expressed wish on the part of a large number of delegates to take part in the
floor discussions, time allotted to take the floor was held at a minimum, in 6-min slots or even 2-min
slots. Even general reports presenting review papers on basic conference topics were accorded only 10-
to 15-min time. A daily conference bulletin aided measurably in keeping participants abreast of the con-
ference proceedings and offered helpful comments.
The fairly high degree of optimism with respect to the present status and future development of
nuclear power was shared in common in all of the reports on nuclear power topics, and also in most of
the floor discussion.
The next scheduled World Energy Conference (the 10th) will be held in 1977 in Istanbul (Turkey).
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INTENSIFIED HEAT TRANSFER IN CHANNELS
OF RBMK TYPE REACTORS
A. N. Ryabov and V. N. Filippov
A conference on intensification of heat transfer in the channels of RBMK type reactors was held in
Moscow in early April, 1974, under the chairmanship of Academician N. A. Dollezhalf, to provide a forum
for discussion of the state of the art, results of experimental research, and pathways of future research.
The problem of how to increase the electrical power output of a power generating mit in a nuclear
power station with RBMK type reactors, while keeping the dimensions and the number of process channels
constant, and also keeping the same basic reactor equipment, came under discussion. An increase to 4500
kW in the thermal power output of the channels should not bring about burnout phenomena at elevated heat
flux densities and high coolant steam content levels. Moreover, the temperature of the outer surface of the
wall must be kept constant in the process channel, and the maximum temperature of the graphite moderator
must be kept within permissible limits. Head losses over the circulation loop must be limited to 18.6 bar.
Coolant parameters can be smoothed out over the transverse cross section of the fuel assembly in order
to optimize the positioning of fuel elements in that cross section, which should in turn allow an increase in
critical (burnout) heat flux density. Calculated and experimental studies on hydrodynamical optimization
of fuel element positioning in the fuel assembly have shown that the critical output level of the fuel as-
sembly can be increased by 5 to 10% simply by judiciously altering the arrangement of fuel elements in
the assembly.
Increasing the inner diameter of the reactor channel from 80 to 82 mm by decreasing the channel
wall thickness makes'it possible to raise the channel output level by 5-7% while holding to permissible
hydraulic drag values.
An increase in channel output to 4500 kW without any changes in the design of the channel and fuel
elements, and while keeping the fuel loading constant, can be attained by using what are known as heat-
transfer intensifiers in the core, so that the steam content at the exit from a maximally stressed channel
can be increased by as much as 50 to 70%, and as much as 30 to 40% as averaged over the entire reactor.
In this case the total flowrate of coolant through the reactor experiences an abrupt drop, leading to a re-
duction in electric power costs for the local needs of the plant itself.
On the basis of experimental data obtained in various scientific-research organizations throughout
the country, thermal hydraulic characteristics of a channel with 80 and 82 mm inner diameter were cal-
culated for power levels of 4200 and 4500 kW, in the case of fuel assemblies containing fuel elements 13.5
mm in diameter and various heat-transfer intensifiers. It was assumed in these calculations that the lower
fuel assembly would have 10 regular spacing grids, while the upper fuel assembly would have 20 spacing
grids with heat-transfer intensifiers. The local hydraulic drag ratio for spacing grids with intensifiers
was assigned a value of 2.5, in contrast to the 0,5 value assigned in the case of regular grids.
Calculations showed that the maximum steam content is -65% when the allowable loop hydraulic
drag is 18.6 bar, the channel inner diameter is 80 mm, the channel output level is 4500 kW, and the maxi-
mum heat flux density is 1.16 MW/m2. Preliminary results on thermal-hydraulic investigations pointed
to the possibility of actually attaining such steam content levels. "
Results of thermal-hydraulic investigations of channel models with different intensifiers, obtained
within the recent period at different scientific-research organizations, came under discussion. All of
these investigations were carried out according to a generally recognized procedure: the experimental
Translated from Atornnaya Energiya, Vol. 38, No. 1, pp. 56-57, January, 1975.
? 1975 Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011. No part of this publication may be reproduced,
stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying, microfilming,
recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $15.00.
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sections were heated by low-voltage current, and the onset of burnout was attained by raising the electrical
power smoothly until the temperature conditions of the rods were changed at constant pressure, constant
flowrate, and constant enthalpy at the entrance.
Results of experimental burnout studies using rod assemblies with spacer grids and local intensi-
fiers ? vortex generators or swirl vanes located in the grids and spaced 125 mm apart over the entire
height of the assembly ? were reported out. The intensifiers in question consisted each of six pairs of
segments of twisted variable-cross-section strips radiating outward at 600 angles.
The experiments were conducted on assemblies of seven cylindrical rod type fuel elements designed
to simulate 12 mm diameter fuel rods with 2200 mm heated length and situated in a regular triangular
grid array in a 45.5 mm diameter channel. The empirical data were obtained at pressures 68.5, 88.1, and
117.6 bar and mass velocities of 600, 800, 1000, and 1500 kg/m2. sec.
Comparison of the empirical data obtained both with and without intensifiers demonstrated that the
intensifiers account for as much as a 50% power output increment, and that hydraulic drag increases by a
factor of 1.5. In addition, the use of intensifiers has the effect of substantially broadening the range of
burnout-free performance of the assemblies while increasing the critical heat flux density level.
The effect brought about is accounted for by the existence of a rotational flow pattern on the part of
the coolant as it flows through the vortex swirler and by separation of moisture from the core of the stream
on the heat-generating surface of the rods. The twist in the coolant flow also contributes to intensified
agitation or turbulization and to vigorous mixing between adjacent cells, which in turn leads to leveling
out of the enthalpy in the transverse cross section of the assembly.
Results of an experimental investigation of burnout in rod assemblies with heat-transfer intensifiers
made in the form of strips twisted in spirals and spaced 165 mm apart, running the entire length of the
assembly, were reported out and discussed. The intensifiers filled the clearances between the fuel rods
over the entire heated length.
These experiments were carried out on assemblies of 21 fuel elements each 13.5 mm in diameter
and of 900 mm heated length. The rods were arranged on a square grid in a 90 mm channel. The experi-
mental data were obtained at 68.6 bar pressure, 800 to 1000 kg/m2 ? sec mass velocities, and 0.58 to 1.16
MW/in2 heat flux densities.
As the experimental data revealed, the critical steam content in this type of assembly is increased
1.5 to 2 times over that in an assembly with no intensifiers present. It is evident from the findings of this
investigation that the hydraulic drag presented by the assembly increases by -50% in this instance.
The conference also heard data on experiments designed to study the effect of the spacer grids
functioning as turbulizers or vortex generators on the temperature conditions of an annular channel and a
bundle consisting of four rods arranged on a square lattice array. The rod diameter in both experimental
sections was 13.5 mm in this case, and the heated length was 3500 and 7000 mm. The experiments were
conducted at pressures 68.5 to 127.3 bar and at mass velocities in the range from 1000 to 2500 kg/m2. sec.
In the experiments using an annular channel, turbulization of the flow was achieved through reliance
on transverse ribs of rectangular cross section, 1.0 or 2.0 mm in width. The rib spacing pitch was 375,
187.5, and 94 mm. It was found that the temperature conditions varied without burnout type discontinuities
as usually observed in a channel with no intensifiers present. But the rate of wall temperature rise with
increasing power level increased abruptly in response to this (by as much as 20 to 50 times). The greatest
power output gain was 35-40% when ribs were spaced 187.5 mm apart. The gain achieved was smaller in
experiments with four-rod assemblies.
One of the reports cited results of an experimental investigation on temperature patterns and burnout
behavior for an experimental section comprising a thermal-hydraulic model simulating a channel with a
full-size transverse cross section for the flow passages. The fuel assembly in question consisted of 18
heated tubes 15 mm in diameter and arranged in square and triangular lattices within a 79 mm diameter
channel. The heated length of the bundle was 1100 mm. Both spacer intensifier grids contracting the ef-
fective cross section of the assembly by 65% because of the lobes on them that were bent back, and regular
spacer grids set 175 and 350 mm apart, were employed in these experiments.
The temperature conditions brought about in the assemblies with the regular spacer grids were in-
vestigated at pressure 74 bar, mass velocities of 600, 1000, and 1500 kg/m2 sec, and with unheated water
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supplied at the entry to the section. The results of these experiments showed that qualitatively the same
temperature distribution at the walls occurred at all of the mass velocities studied, viz.: the highest tem-
perature was exhibited by surfaces located in the central cells, and the lowest by those in the peripheral
cells, while those located in intermediate triangular cells featured a still lower temperature, and the
lowest temperature of all was found in the intermediate square cells. Since the heat flux was distributed
uniformly over the rods, that would mean that heating of the coolant was maximized in the central cells of
this pattern.
The investigation of temperature conditions obtaining in assemblies with spacer intensifier-grids
showed that the coolant was heated the most in the peripheral cells. Heating was pretty much the same
in the central and intermediate square cells.
Burnout phenomena were studied at pressures 58.8 and 74 bar, and at mass velocities of 600, 1000,
and 1500 kg/m2 ? sec. The sites of onset of burnout, the rise of the wall temperature accompanying the
onset of burnout, and fluctuations in wall temperature in the region beyond burnout, were all recorded in
the experiments. In the assemblies with regular spacer grids, burnout generally occurred in the narrow
clearances separating the central cells, and the level of critical heat flux densities was in accordance
with calculations based on the V. S. Osmachkin formula. In assemblies with spacer intensifier-grids,
burnout occurred in the peripheral cells in the narrow clearances intervening between the peripheral rods
and the channel.
As the experiments revealed, the effect brought about by the use of spacer intensifier grids involved
a smooth rise in the wall temperature of the rods in the region beyond burnout, while the critical heat flux
density rose by 30-40% in the nucleate boiling region.
The findings of the hydraulics study revealed that hydraulic drag increases on the average by 2.9
times, when spacer intensifier grids with a pitch of 175 mm are installed, as compared to the situation
in assemblies with regular grids.
The feasibility of using a "multistage" channel, i.e., a channel separated in height into a number of
sections included in parallel as far as coolant flow is concerned, in a RBMK type reactor was reported.
In a channel of this type, the critical heat flux densities would be slightly higher because of the low mass
velocity, within the confines of each distinct stage, but hydraulic drag opposing the inflow and efflux of
coolant would be greatly increased.
One of the reports dealt with an analytical treatment of burnout at high steam content levels. A
procedure for solving the balance equation for liquid in the film was presented as an aid in calculating
limiting steam content. Equations were derived for use in calculating the limiting liquid flowrate through
the film and the wetting intensity. Theoretical data were found to be in adequate agreement with empirical
data.
The conference noted the fact that the scientific-research work and experimental design work on
intensified heat transfer in relation to the operating conditions of RBMK type reactors is making it pos-
sible to anticipate a 20-50% rise in the output level of the reactor technological (process) channel without
any substantial changes in the reactor design as a whole, and primarily through improvements in the
design of the fuel assemblies. It was recommended that steps be taken toward gaining practical acceptance,
on an experimental level, of intensifying devices in the form of improved regular spacer grids, without
any changes introduced into the design of the fuel elements as such
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THE SECOND ALL-UNION CONFERENCE ON THE
METROLOGY OF NEUTRON RADIATION AT REACTOR
AND ACCELERATOR FACILITIES
R. D. Vasilley
The Second All-Union conference on the metrology of neutron radiation at reactor and accelerator
facilities was held in Moscow on October 14-17, 1974. This conference was organized and sponsored by
the VNIIFTRI (All-Union scientific-research institute for physics and engineering and radio-engineering
measurements) on the initiative of the State Committee on Standards attached to the Council of Ministers of
the USSR, and of the State Committee on the Peaceful Uses of Atomic Energy of the USSR. Participating
in the conference were specialists assigned from over 80-odd organizations under the aegis of various
ministries, departments, and the Academy of Sciences.
The purpose of this gathering was to draw up a balance sheet of metrological research over the
three-year period elapsed between the first and second conferences. A collection of papers, in two vol-
umes, entitled "Metrology of neutron radiation at reactor and accelerator installations," published to coin-
cide with the inauguration of the conference, was distributed among the participants.
Review papers on topics pertaining to the metrology of neutron measurements at nuclear physics
installations and facilities were presented at a plenary session. Achievements of the past few years were
discussed and new problems on the horizon were formulated. The structure of metrological servicing of
measurements work at nuclear physics installations was discussed, as well as standard specimens and
physical standards for sources (fields) of neutrons, standard physical specimens of materials, counters
and instruments, nuclear physics constants, standard procedures to be followed in taking measurements,
calibration work, and certification, intercomparisons at different reactor installations, problems in-
volving how to go about determining errors and design of experiments, terminology in the field of neu-
tron radiation, It was pointed out that the beginnings of the organization of a system of metrological ser-
vicing of measurements at nuclear physics installations which would be unique for the various administra-
tive departments have been laid with the devising of a checkout pattern for equipment used in measuring
the characteristics of neutron fields at given installations, a state-sponsored special physical standard
of the unit of neutron flux density using an accelerator as the basic metrological tool (at VNIIFTRI), the
first standard specimens of neutron sources based on nuclear reactors [at the I. V. Kurchatov Institute
of Atomic Energy (IA) and at other organizations as well], and also by the production of a variety of
standard specimens equipment including inparticular sets of standard AKN specimens (neutron activation
sets) and NDS specimens (neutron tracking detectors). This system encompasses measurements of the
characteristics of neutron fields over the entire range of reactor neutron energies at flux densities to
10" neutrons /cm2 ? sec.
There were 43 reports presented at a panel session called to discuss "Equipment for measuring the
characteristics of neutron fields, and standard physical specimens; calibration and certification." These
43 reports can be grouped according to the following basic trends covered in them: standard sources based
on nuclear reactors and other equipment for calibration work, continuous-acting instruments and detectors
(gas-discharge counters, calorimetric detectors, multisphere spectrometers, neutron identifiers respond-
ing to pulse shape, etc.), detectors consisting of fissionable specimens and specimens susceptible to acti-
vation in boron and cadmium absorbers, activation specimens for analysis of materials, detectors and
targets composed of fissionable isotopes, and fission fragment recording devices. Separate aspects of
Translated from Atomnaya nergiya, Vol. 38, No. 1, pp. 58-59, January, 1975.
? 1975 Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011. No part of this publication may be reproduced,
stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying, microfilming,
recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $15.00.
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measurements of induced activity, and problems involving isotope analysis of fissionable materials, were
also discussed at the panel.
A panel on "Measurements of the characteristics of neutron fields at reactor and accelerator instal-
lations" drew 23 reports. Some of these dealt with measurements of thermal neutrons and intermediate-
spectrum neutrons, the remainder with measurements of fast neutrons. Investigations carried out at reac-
tors accounted for almost the totality of the reports presented, only two of them delving into problems
encountered in work with accelerators. Applications of individual types of continuous-acting detectors
were among the topics covered, as well as various methods of recovering the spectrum of fast neutrons,
and results of experimental separation of fast neutron spectra. A report was given on attempts at evaluating
errors in the results of spectrum restitution. Attention of researchers working with reactors and accelera-
tors continued to focus on indirect and combined measurements, as the modes of measurements providing
the most complete volume of reliable information on the characteristics of neutron fields. The number of
experimental physicists making use of standard procedures in measurements work is on the increase.
Attention was drawn to the way closer attention is being given to more precise estimates of errors. For
example, values on the low side are being reported less and less often because of the painstaking analysis
of sources of systematic error.
Six reports were presented at the panel session on "Intercomparisons at reactors." Interest in this
problem has been steadily on the rise. For example, in recent years several organizations have been par-
ticipating in six national intercomparison programs and one international intercomparison program, en?
compassing indirect measurements and combined measurements of spectral and dosimetric character-
istic. Of the most remarkable national intercomparisons reported on, we must mention intercomparisons
of results of measurements on a static reactor that were carried out during 1974. The organizations, using
equipment for measurements over the entire range of reactor neutron energies, took part in this inter-
comparison project. The intercomparison program included preliminary comparisons of methods for
restituting a test spectrum. The national and international intercomparisons made it possible to discern
the most highly improved procedures and equipment for taking measurements, and to recommend promis-
ing detectors and instruments for practical applications.
The panel on "Nuclear data for neutron measurements, and metrological aspects of nuclear data
research" heard contributions by the leaders of the Nuclear Data Center, V. N. Manoldain [FII (Power
Physics Institute, Obninsk], and Center for Data on the Structure of the Atomic Nucleus and on Nuclear
Reactions, F. E. Chukreev [I. V. Kurchatov rAE (Atomic Energy Institute), Moscow]. These reports
gave accounts of the trends evident in the work of their data centers in relation to the needs of experi-
mental physicists for data used in neutron measurements. They acknowledged the importance of the activi-
ties undertaken by the Nuclear Data Commission of GKAE (State Committee for the Peaceful Uses of
Atomic Energy), by coordination teams on measurements of nuclear physics constants, and nuclear data
needs of reactor facilities and in shielding and protection work, by topic-oriented teams on various types
of neutron constants, standards, and reference quantities. Reports were also presented on measurements
of the fission neutron spectrum of 252Cf, on approximation of cross sections of threshold reactions, on
metrological analysis of experimental determinations of particular constants.
The panel on ',Determination of errors and design of experiments" discussed problems concerning
calculation of errors in indirect measurements, and approximation of calibration dependences. Methods
for estimating the information content of various experiments involving measurements of cross sections
were discussed, in addition to problems involving design of experiments and neutron data estimates for
reactors.
A general discussion was held at the concluding plenary session. The efforts of the various organ-
izations to solve various problems in metrology in the field of neutron radiation were approved. It became
clear from the discussion on the floor and from the reports presented that metrological studies are being
carried out more and more often by several organizations jointly. Coordination of these efforts is being
aided by exchange of views and opinions such as that afforded by these All-Union conferences.
The next such conference is scheduled for sometime in 1977. Organization of this next conference
is also entrusted to VNIIFTRI.
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MIFI TWENTY-SIXTH SCIENTIFIC CONFERENCE HELD
V. V. Frolov and V. A. Grigortev
The twenty-sixth scientific conference of instructors, students and staff personnel of MIFI (Moscow
Engineering and Physics Institute), devoted to the 250th anniversary of the Academy of Sciences of the
USSR, was held in June, 1974. A total of 480 reports and communications were presented at 26 panel
sessions.
The panel on experimental nuclear physics listened with keen interest to reports submitted on experi-
ments dealing with searches for the W-boson at the IFVE (Institute of High-Energy Physics) accelerator
located at Serpukhov. It was shown that a W-boson having a muon decay mode v-10% does not exit with a
mass in the interval from 5 to 8 GeV, at the level -10-33 cin2 of the production cross section of a W-boson
in p-N collisions. These experiments also yielded data on interaction cross sections of particles at ener-
gies of 70 GeV in the case of large transverse momenta (1.8 to 2.8 GeV/c). A new ir-meson generation
mechanism was detected (the intensity of these p-mesons amounts to -2.5.10-5 the intensity of pions in
the range of transverse momenta investigated).
F. M. Sergeev gave an account of experimental research on pion-nuclear reactions using the method
of heavy-liquid bubble chambers. From among the physical results reported, we should mention the mea-
surement of angular scattering of pions of both signs on carbon nuclei in the 5-20 MeV energy range. This
experiment made it possible to establish the fact that the pion-nuclear interaction in the o-state exhibits a
repulsion nature.
A report by P. S. Baranov et al., delivered at the panel on experimental techniques in nuclear
physics, cited new experimental data on the time resolution of large-size scintillation counters when Soviet-
make FEU-85 and FEU-87 photomultipliers were made with quantity-manufactured plastic polystyrene
base scintillators. The time resolution of counters of dimensions 400 x 100 x 20 mm was < 1 nsec when
relativistic particles were recorded.
The panel on theoretical nuclear physics responded with greatest interest to a report by Academician
A. B. Migdal presenting recent results he obtained on the theory of the ir-condensation in nuclei, which is
related to the production of 7r-meson pairs in the nucleus.
The plasma physics panel heard reports by 0. A. Vinogradova, S. K. Dimitrov, A. N. Igritskii,
V. M. Smirnov, D. A. Panov, V. A. Leitan, and A. S. Lutstko, discussing engineering problems of
thermonuclear fusion reactors. Results pertaining to the further development and improvements of an
ion energy recuperation system proposed on an earlier occasion were discussed at this panel. The ions
in question escape from an open type reactor trap. On the basis of experiments centered around the LIN-5
research facility, an estimate of the efficiency of ion energy recuperators in the injector circuit of
Tokamak machines was arrived at: the expected efficiency of real direct conversion systems is 0.85 to
0.90.
D. A. Knyazev et al. evaluated the outlook for acceptance of various physical and chemical methods
of isotope separation, in a report presented before the panel on physics of separation processes.
A lively discussion was touched off at the heat-transfer physics section by a presentation by V. A.
Andreev et al. on investigations of heat transfer in the cryogenic temperature range. We should also
mention a fundamental contribution by L. S. Kokorev et al. laying the basis for physical representations
on the nature of burnout phenomena on heat-generating surfaces.
Translated from Atomnaya Energiya, Vol. 38, No. 1, p. 59, January, 1975.
0 1975 Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011. No part of this publication may be reproduced,
stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying, microfilming,
recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $15.00.
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The panel on applied nuclear physics responded with deep interest to reports on applications of acti-
vation radiography techniques in nuclear y -resonance work as part of a rounded integrated study of diffu-
sion and redistribution processes affecting impurity and alloying elements in the corrosion of zirconium
and zirconium alloys. The same methods were used in obtaining data on the parameters and nature of
oxygen diffusion, on the distribution of alloying elements in oxide type zirconium films, and in oxide films
of zirconium alloys.
The panel on electrophysical equipment heard results obtained in the development of new models of
linear electron accelerators for applied purposes. The first model was designed for a pulsed current of
accelerated electrons of 1 A, and is capable of producing 4 to 10 MeV energy. The second model operates
at a high frequency ('-'10 GHz), so that a successful compact design was feasible. This model has been
exhibited at the Exposition of Achievements of the National Economy (VDNKh).
The panel on solid state physics sustained a lively discussion stimulated by a report presented by
T. I. Kozin et al., outlining a method for recording long-lived plasma with the aid of a stabilized laser.
This method favors the use of noncontacting techniques in investigating low-density plasma with densities
down to ,4012 cm-2, and is distinguished by its comparative simplicity.
The concluding plenary session heard a report entitled "High-energy physics and the Academy of
Sciences of the USSR," delivered by Corresponding Member of the Academy of Sciences of the USSR A.
M. Bald in. The history of the design and development of accelerators in the Soviet Union was presented
in his brilliant and content-packed presentation, which also took up the role of scientists belonging to the
Academy of Sciences of the USSR, and particularly the role played by FIAN (Lebedev Institute of Physics
of the Academy of Sciences of the USSR) in establishing and developing Soviet accelerator facilities and
equipment.
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THE FOURTH ALL-UNION PLASMA PHYSICS SCHOOL
V. A. Papadichev
The Fourth All-Union plasma physics school was held in Novosibirsk, July 29 through August 7,
1974. Participating in this session of the school, an institution which has become traditional in its field
and which has been scheduled once every two years, were 200-odd Soviet and foreign scientists. The 1974
school agenda included studies of high-power beams of relativistic electrons and applications of beams of
relativistic electrons in research on controlled thermonuclear fusion, and also plasma confinement sys-
tems and heating systems for high-density plasma. One of the problems tackled by the school, as stated
in the introductory remarks by the chairman of the Organizing Committee, G. I. Budker, was analyzing
the present status of research in those areas. The lecturers included leading scientists of the USSR, USA,
and Italy. Seminars were also staged as forums for detailed and informal discussion of a variety of prob-
lems: high-density plasma, collective acceleration of ions as beams of relativistic electrons traverse a
gas, techniques and equipment in the generation of ultrahigh-power electron beams, coupling of high-power
electron beams and plasma. The floor discussions at the lectures, seminars, and in the corridors, con-
tributed to a more complete exchange of ideas and information and to the establishing of close contacts be-
tween the scientists of different countries and laboratories.
A report by M. S. Rabinovich [FIAN (Lebedev Inst. Physics)] contained a review of the status of
thermonuclear research efforts throughout the world. The present stage is characterized by a transition
to the construction of large machines and installations (one being scheduled for commissioning in Europe
in 1977, another for the early 1980s in the USA) at which work on basic scientific problems will compete
with efforts to solve technological and engineering problems in the design of a successful thermonuclear
fusion reactor.
Problems pertaining to the production of powerful beams of relativistic electrons for achieving
pulsed thermonuclear reactions found reflection in a report presented by M. V. Babykin [I. V. Kurchatov
IA E (Institute of Atomic Energy)]. Relevant calculations indicate that a beam of duration ?1 nsec and
energy 102 J (power level ?1016W) would be required for the purpose. Reliance on a heavy shell around
the D?T mixture would necessarily lead to relaxing the restrictions on the beam (10 nsec, 3-106 J, 3 .1014
W). Conversion to disk-shaped and conically tapered shaping circuits with peripheral switching and to
positioning of a cold-emission gini at the center would be feasible in efforts to generate such beams. In
order to arrive at the required high-power growth rate (>3 .1022W/sec), switching inductance would have
to be kept down to 2-3 nH, which would in turn call for the use of 100 to 200 discharge switches working in
parallel. Magnetic insulation would do for the transport of the pulse to the gun. Experiments on multi-
channel switching devices are to be continued at IAE, using the new Angara-1 facility (2.5 MeV, 400 kA).
Similar problems were approached in a report delivered by A. Kolb (Maxwell Corp. Laboratory,
USA). Beams with a high power growth rate can also be used to excite high power UV gas lasers. The
principal problems confronting designers of plasma machines are, in Kolb's view, how to develop a
compact storage device, pulse switching (appropriate discharge switches), and generation of the beam per
se through the use of a field-emission diode. Several types of multichannel discharge switches functioning
in water, oil, or compressed gas environments have been developed, with voltage to 5 MV and power
switched in the range up to 1012W. Trigger pulse height is generally 20-25% of the voltage switched. How
to decrease the inductance of the discharge switch is the very first problem to be tackled, although the
diode itself actually accounts for anywhere from 20 to 30% of the total inductance. The diode wave impe-
dance is closely described by the Child?Langmuir law when proper attention is given to the displacement
of the plasma interface in the diode, i.e., to changes in the nonconducting clearance in the diode. The
Translated from Atomnaya Energiya, Vol. 38, No. 1, pp. 60-61, January, 1975.
? 1975 Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011. No part of this publication may be reproduced,
stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying, microfilming,
recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $15.00.
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duration of the voltage pulse rise is 5 to 10 nsec in the new medium-energy (1-2 MeV) high-current elec-
tron accelerators developed by the Maxwell Corp.
Some interesting experiments on focusing of high-current electron beams were described in a report
by J. Witkowitsky (Naval Research Laboratory, USA). In these experiments, a hollow cylindrical beam is
admitted into a drift chamber with a side wall made of insulating material and with a central conductor
made in the form of wire running between the anode and the second end face of the chamber, to short back
current. The beam was focused on the wire, and later on (in the case of a thin wire) the beam was not
immediately destroyed but allowed to follow the wire, even if it was a bent wire. A yield of 109 to 1011
neutrons/pulse resulted when the wire was coated with deuterated polyethylene. The beam current was
?300 kA in these experiments, the electron energy 200 to 400 keV, and the beam duration 50 nsec.
A report by G. A. Mesyats (Institute of Atmospheric Optics, Siberian Branch, Academy of Sciences
of the USSR) dealt with the initial stage in the shaping of a plasma cathode and the current flowing through
the field-emission diode. The decisive factors here are, in the author's opinion, explosive emission and
the propagation of the cathode flare (plasma and cathode).
The results of work done on devising pulsed high-current electron accelerators at the Nuclear
Physics Institute, Siberian Branch, Academy of Sciences of the USSR (IYaF SO AN SSSR) were reported
by V. M. Logunov. The shaping circuits were shaped in these machines with the aid of a Tesla pulse
transformer; the usual circuitry with discharge switches in the line and a field-emission diode were em-
ployed later. At the present time, accelerators with a single 250 keV, 250 kA, 40 nsec and 800 keV, 110
kA pulse shaping circuit are being used. An accelerator with a double pulse-shaping circuit in water (1
MeV, 230 kA, 60 nsec) is being built, and a 1 MeV, 300-400 kA accelerator is being planned. All of the
accelerators are being used or are being designed for experiments on beam?plasma coupling in a variety
of configurations (multiple-mirror configurations included). Some interesting experiments using mockups
to simulate a rise in the electrical conductivity of the water with the aid of a pressure rise (a threefold
rise of 120 atm) and screening of microcorona points on the electrode surfaces as an electrolyte solution
is forced through the electrode surfaces into the electrode sheath region were also reported on. In the
case where both electrodes are screened, the electrical conductivity of the clearance rose by a factor of
2 to 2.5. But we should draw attention to some engineering difficulties impeding the materialization of the
last technique, when electrode areas in existing accelerators are appreciable.
A basic research trend in evidence at Sandy laboratory (USA) centers on the use of high-power elec-
tron beams in heating and compression of a high-density plasma formed from D?T mixture grains. A
report by J. Jonas entitled "Nuclear fusion in microparticles of a D?T mixture" cited research findings on
beam focusing in a diode. Calculations showed current densities up to 107 A/cm2 at 300 kA current and a
2.5 mm anode-to-cathode gap (2.5 MV); densities of around 5-109 A/cm2 were obtained on the Hydra
machine (1 MV). A holographic study was made of a plasma formed when a beam from an angular cathode
impinged on a hemispherical anode in the diode. It was demonstrated that the plasma was distributed uni-
formly and symmetrically about the anode, which corresponds to symmetrical compaction of the plasma
in this geometry. An experiment being planned will hopefully shed light on irradiation of a D?T mixture
particle in the form of a spherule in two opposite directions (i.e., two cathodes), as a way of achieving
uniform compression. The large 3 MeV, 1012W EBFA accelerator, with ?20 nsec pulse duration, is being
built especially for this experiment. This machine includes six pulse-shaping circuits in the form of 27-m
disks commutated by specially designed multispark discharge gaps with overvoltage features, and two
diodes positioned at the center of the system. Beam generation on the EBFA accelerator is scheduled for
the spring of 1975.
A seminar dealing with equipment and techniques for generating ultrahigh-power electron beams
covered accelerator parts and subassemblies. Some of the reports cited data on current distribution in
the channels of multispark discharge gaps and some of the time characteristics of such devices. There
was a report on experiments using foil type switches in work with inductive storage devices, on the de-
velopment of a high-current long-pulse electron accelerator at NIIEFA based on an inductive storage
device, etc.
New results from an investigation of high-current ion beams were cited in a paper by R. Sudan
(Cornell University, USA). A three-electrode gun (anode in the center) was employed to produce 2.5 kA
proton current of 50 nsec duration and 300 keV energy. It is expected that 50 kA current will be generated
on a larger (5 MV) accelerator, These beams will be employed in plasma heating, and injection into an
Astron machine to bring about an E-layer (as reported by H. Fleischmann), and in the formation of dense
ion rings in investigations of shock-waves propagating through solids, etc.
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The widest variety of aspects of the problem of beam?plasma interaction came under discussion:
traversal of plasma by the beam giving rise to instabilities, mechanisms of effective energy transfer from
beam to plasma, acceleration of ions in the system plasma?beam?magnetic field, beam?plasma interac-
tions in the presence of an rf field, etc. It should be pointed out that the experimental research conducted
in this area is of a highly complex nature requiring development of special custom-engineered equipment
and measurements procedures, and that considerable experimental material obtained to date still awaits
clearcut and definitive theoretical analysis for the most part.
A seminar on acceleration of ions when high-current electron beams are passed through a gas was
held over and above the school program, on the initiative of A. A. Kolomenskil (FIAN). The seminar dis-
cussions brought to light a trend toward increasing complexity in models of acceleration (two acceleration
stages, treatment of effects of ion motion, etc.), since none of the models proposed earlier has been
found to account satisfactorily for all of the available experimental data.
Some basic research findings obtained by Novosibirsk physicists in plasma studies were reported
out. This work is being carried out generally in a variety of nontraditional approaches; experiments on
plasma confinement in multiple mirror configurations, a rotating-plasma trap, high-intensity beams of
monatomic hydrogen, and theoretical work.
Reviews of theoretical concepts on beam?plasma coupling were provided by prominent specialists
(A. A. Rukhadze, N. Rostoker, L. I. Rudakov), and rested to a significant extent on original work done
by those authors. In addition to deepening analytical research, closer attention is being given to simulation
with the aid of high-capacity analog computers; the fruitfulness of this approach was demonstrated in
studies of strong Langmuir turbulence,
On the whole, the work of the school demonstrated that high-power electron beams and ion beams are
being used ever more widely in research on plasma physics and in other areas of science, and also in the
solution of various applied problems. Generation of high-power electron beams and ion beams, the study
of special features of the propagation of such beams through various media, are at this point of independent
interest as a branch of science in its own right. The school provided fresh confirmation of the feasibility
of such a rounded discussion on one or two of the most urgent and timely problems selected from among the
many problems confronting plasma physics research.
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INFORMATION: SCIENCE AND ENGINEERING LIAISONS
FAMILIARIZATION TRIP BY SOVIET SPECIALISTS
TO SWEDEN
A. D. Amaev
A team of Soviet specialists headed up by Academician A. P. Aleksandrov visited Sweden September
3-14, 1974. The purpose of this trip was to become better acquainted with the work being done in the field
of nuclear science and nuclear engineering and industrial plants producing reactor equipment. The dele-
gation visited the State Hydroelectric Power Authority, and also visited various nuclear power plants now
in the startup stage or under construction, at Oskarshamn and Ringhals; a plant engaged in the production
of heavy reactor equipment (reactor pressure vessels, steam generating units, etc.), the Uddkomb firm,
the ASEA-atom design and production center, where our specialists became familiarized with a new fuel
element and fuel assembly manufacturing plant for deliveries to water-cooled water-moderated boiling
water reactors and heat-transfer experimental test stands; hot laboratories, the R-2 test reactor, and
test stands in operation at the AB Atomenergi research center in Studsvik, the well-known Sandvik concern
and its tube manufacturing plant which puts out jackets for fuel elements, as well as steam generators
and reactor loops, and metallographic and metal physics laboratories.
On the request of the delegation, a meeting was organized, above and beyond the previously agreed-
upon program, with a group of specialists to discuss the nuclear power station safety programs being
implemented at the former Marviken nuclear power plant.
On September 12, A. P. Aleksandrov spoke at the Royal Academy of Engineering Sciences before
an audience of prominent specialists in nuclear science and nuclear engineering, and heads of Swedish
firms, on "A rational fuel cycle in the nuclear power industry," eliciting keen interest and numerous in-
quiries from the floor. This occasion also prompted a discussion in a smaller group which led to an ex-
change of views with Professor G. Hambreus, General Director of the Royal Academy of Engineering
Sciences, with A. P. Aleksandrov delivering a brief review of problems providing a basis for possible
future development of mutually profitable liaisons between the Soviet Union and Sweden. The close attention
given to the Soviet delegation is accounted for by the significant contribution made by the Soviet Union in the
foundation and further development of nuclear science and nuclear industry, as well as the traditionally
benevolent relations prevailing between ow two countries.
Sweden, as a highly developed nation, has practically everything it needs (in terms of natural re-
sources, its own developed and tested boiling-water reactor projects involving BWR reactors, the Westing-
house Corp. PWR reactors, scientific-research facilities with highly trained engineering and scientific
personnel, a high industrial potential with up-to-date equipment) for carrying out its plans for introduction
of nuclear power capacity expected to reach a total level of 15,800 MW(e) by the year 1990. But Sweden is
experiencing certain difficulties stemming from the resistance being put up by firms not connected with the
Nuclear Industrial Group, as well as oppositional public opinion in the country on the possible harmful
effects of nuclear power stations on the natural environment. These factors have been shaping scientific and
engineering policy in the planning of the activities of research centers, and also in the use of the former
Marviken nuclear power station for organizing full-scale highly expensive special projects within the frame-
work of the nuclear power station safety program (verifying core cooling performance in response to
scramming brought on by loop leaks or failure, service-life and endurance tests on specific pieces of
equipment and subassemblies, etc.). Experiments carried out at Marviken within the framework of the
nuclear safety program involve participation and financing by several other countries as well. The scien-
tific findings resulting are the property of those countries, and can be made available to still other coun-
tries only through mutual agreement and financial compensation of costs incurred in the program.
Translated from Atomnaya Energiya, Vol. 38, No. 1, pp. 61-62, January, 1975.
? 1975 Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011. No part of this publication may be reproduced,
stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying, microfilming,
recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $15.00.
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Taking into consideration the unstable petroleum pricing situation and the constant rise in petroleum
costs, Sweden is now engaged in cooperation with other countries in a costly safety program aimed at
expediting successful fulfilment of nuclear power station construction plans. The Soviet Union is also
interested in establishing contacts with those participating in this program.
Highly sophisticated manufacture of fuel elements and fuel assemblies for BWR reactors, including
fabrication of sintered fuel pellets, has been organized at the plant run by the ASEA-atom firm. A produc-
tion line for the fabrication of fuel elements with helium pressurized to 30 atm under the canning jacket is
being planned for PWR reactor equipment. This plant will not only meet the needs of Sweden's own industry,
but will also be able to deliver fuel elements as an export item. The firm plans to expand fuel element
manufacture. Point-of-entry monitoring of tubes for Zircaloy-2 jackets of Sandvik fuel elements will be
carried out using plant equipment manufactured by Swedish industry or imported from other countries.
Scrap, put at 0.2%, will be examined to update information in order to eliminate causes of scrap and im-
prove the reliability of the fuel elements at the same time. Fuel elements designed by ASEA-atom offer
some design advantages over fuel elements produced by our industry.
Operation of BWR fuel elements in the first power generating unit of the Oskarshamn nuclear power
station has turned up leakage in only two out of ?30,000 fuel elements in service after 12,000 MWd/ton
U burnup. Even though the burn up attained is about half of the level planned for, this result must be con-.
sidered highly encouraging.
Fairly extensive research has been pursued both at the Studsvik research center and at the Sandvik
plant, and this includes reactor investigations involving studies of the effect of heat treatment and machin-
ing on the mechanical properties of reactor materials and studies of the structural stability of zircaloy
tubing and fuel-element cans.
It is worth pointing out that the Sandvik firm is placing great hopes on utilization of the steel grade
12R72HV (15% Cr, 15% Ni, 1.8% Mn, 1.2% Mo, 0.4% Ti, 0.006% B) which the firm itself developed, for the
fabrication of cans of fuel elements in fast reactors. Extensive studies have been carried out on creep
behavior, long-term strength in the 550-750?C temperature range over a 50,000-h time span, a set of in-
pile tests demonstrating the superiority of this grade of steel over the American grade 347 steel devised
for similar applications. The positive effect brought about by additions of titanium and boron on high-tem-
perature long-term mechanical properties is now evident.
Further investigations on materials for the cans of fuel elements designed for thermal reactors are
being carried out from the standpoint of nuclear power plant safety. A lot of work is being done in the field
of reactor materials studies along the lines of joint programs involving several participating countries
(Belgium, Norway, Britain, FRG, France, and Italy),
A relatively small-size plant for the production of heavy reactor equipment, operated by the Udd-
komb AB firm, is fabricating as many as seven reactor pressure vessels per year to cover Sweden's
needs of one or two pressure vessels per year with the remainder available for export. This plant is
operated mainly as a source of goods for export (to FRG, France, other Scandinavian countries), and is
equipped with highly productive and very sophisticated outsize equipment for all types of machining and
welding operations and inspection of welded joints. The plant is engaged in a program of cooperation with
other foreign firms which serve as vendors for forgings and partially finished fuel-element jackets and
shell type products. The plant handles such operations as machining, hardfacing, welding, inspection,
and inspection of BWR and PWR reactor pressure vessels and steam generators for 1500 MW(e) nuclear
power stations. Reactor pressure vessels are being fabricated from American steel grades A553-B (for
BWR reactors) and A508-B (for PWR reactors). Special attention is being given to inspection techniques,
and costs for inspection work total 10%. Uddkomb AB intends to expand its plant capacity to keep up with
long-term orders already in and in view of the steady interest shown in its products by clients.
Here we may point out that the method currently in use at the Uddkomb plant of applying hardfacing
by means of wire electrodes is inferior to the strip hardfacing method now in use in the Soviet Union. The
Uddkomb firm is now interested in placing orders for reactor pressure vessels manufactured in the Soviet
Union.
In 1967, Sweden, Denmark, Norway, and Finland pooled their efforts in an attempt to design a re-
inforced-concrete pressure vessel for use in a water-cooled water-moderated boiling-water (BWR type)
reactor, and in 1973 Britain and France joined in financing this program (without any direct participation
otherwise). The "Scandinavian program," as it is known, embraces the solution of problems associated
84
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with the strength, design of the vessel cover sealing, thermal insulation, monitoring and safety of such
reinforced-concrete pressure vessels in reactor design. The design of a suitable reinforced-concrete
pressure vessel will be completed in 1975, after which steps can be taken to proceed to installation of a
reinforced-concrete structure in a 900 MW boiling-water reactor facility.
The heads of the ASEA-atom and AB Atomenergi firms displayed intense interest in the work being
done in the USSR an the development of a boiling-water reactor with a reinforced-concrete pressure vessel.
85
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BOOK REVIEWS
NEW BOOKS FROM ATOMIZDAT
(FOURTH QUARTER, 1974)
N. G. Gusev, P. M. Rubtsov,
V. V. Kovalenko, and V. M. Kolobashkin
Radiation Characteristics of Fission Products; a
Reference Handbook*
Reviewed by A. A. Moiseev
This reference manual cites quantitative data on the radiation characteristics of fission products
formed in nuclear reactors and in instantaneous fission events. These data include independent yields of
fission products of 233U, 235u, 238u,239Pu, 719
---Th bombarded by thermal neutrons, fast neutrons (En .14
MeV), and fission spectrum neutrons; total and differential activities and y -ray equivalents of fission
products and of mixtures of fission products; energy release by y -photons per unit dose rate; effective
dose and energy spectral composition of y-ray emission by a mixture of fission products. Data on the
buildup of 134Cs in nuclear reactors are also tabulated.
The integrated radiation characteristics of a mixture of krypton and xenon after these escape from
fuel elements will be of great interest to designers and engineering physicists engaged in calculating doses
from a cloud of inert radioactive gases.
Data are published for the first time on the radiation characteristics of a mixture of fission prod-
ucts from a reactor operated with shutdowns (shutdowns for partial refueling and for preventive main-
tenance) and also from reactors operated in a critical-assembly mode.
The empirical formulas cited in the appendices and the required coefficients for determining total
activity [Q(T, t), Ci/kW) and energy release by y -photons [S(T, t), MeV/sec ? kW] of a mixture of nuclear
reactor fission products as a function of the reactor campaign T and exposure time t will also be of great
practical interest. These relationships are analogs of the familiar Way?Wigner formula for a mixture
of instantaneous-fission products.
We can make a special point of the originality of virtually all of the material in the handbook. There
is no doubt that this will be a desk-top reference book for designers and production experts at nuclear
power stations, spent fuel reprocessing and recovery plants, for physicists and engineers concerned
with protection and shielding against ionizing radiations and radiation safety, and for students and in-
structors in physics and engineering colleges.
We can express some regret that the handbook has not space for data on the differential specific
activities and y -equivalents of radionuclides with a very short half-life. Data are cited for differential
characteristics only in the case of four campaigns (the most important ones, to be sure), and on total
characteristics we have appropriate values for nine reactor campaigns. This presents certain difficulties
when the task is to find solutions for some radiation safety problems.
Among the shortcomings of this reference manual, we can mention the fact that the authors resort
to the term "isotope" (or "radioisotope") instead of "nuclide" (or "radionuclide"), even though the former
*Atomizdat, Moscow (1974), 224 pp.
Translated from Atomnaya Energiya, Vol. 38, No. 1, pp. 63-64, January, 1975.
0 1975 Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011. No part of this publication may be reproduced,
stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying, microfilming,
recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $15.00.
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term can be correctly used only in those cases where we are referring to atoms of the same chemical
element. If we have in mind atoms of different chemical elements, then it would be more correct to hold
to the term "nuclide."
Yu. V. Kuznetsov, V. N. Shchebetkovskii,
and A. G. Trusov
Fundamentals of Cleanup of Radioactive Contaminants
from Water (edited by V. M. Vdovenko)*
Reviewed by D. S. Gol'dshtein
The appearance of this book in print is an important event for all those involved to some extent with
removal of radioactive contaminants, and particularly with cleanup of contaminated water and effluents.
The first edition appeared back in 1968 with the title Fundamentals of Water Deactivation.
The authors have done a good deal of work on their monograph, and have perused new publications in
the field as well as their own research findings. The first and fourth chapters of the book are those that
have been enlarged the most.
Extensive factual material is presented in 168 tables of data and 56 diagrams and illustrations. The
book contains an impressive bibliography (with 986 references).
An interesting feature of this monograph is its approach to the problem of water cleanup to get rid
of radioactive contaminants from the vantage point of concepts relating to the physicochemical state of the
radioactive isotopes present in aqueous media, and the behavior of those isotopes in the sorption systems
under study. The authors convincingly demonstrate the serious value of information on the modes of exis-
tence of radioactive isotopes in aqueous media for solution of practical water cleanup problems. The first
chapter reserves considerable space for a thorough airing of this information. The first chapter also
covers possible sources of radioactive isotopes in natural bodies of water, and processes involving inter-
action between radioactive isotopes and inorganic or organic components of natural waters. The chapter
ends with a prognosis, the first of its kind in the literature, of the probable physicochemical state of radio-
active isotopes in surface waters on land, in the seas and oceans.
The second chapter centers on the basic techniques of water treatment and water management, and
their effectiveness when used to remove radioactive contaminants from water. In addition to a description
of traditional methods at a contemporary level of sophistication, the text presents concepts relating to the
physicochemical essentials of processes involving coprecipitation of trace amounts of radioactive elements
with hydroxy-hydrate collectors. The chapter ends with information cited on new water cleanup processes
for use at water conduit stations in our country and in other countries.
The third chapter offers a concise but content-packed description of other-than-basic methods in wa-
ter treatment and water preparation: soda-lime softening and ion exchange, phosphate coagulation, elec-
trodialysis, distillation, flotation, biological cleanup, and so on. It is of the utmost importance that the
authors cover not only the better known methods, but also methods which have only begun to be accepted
in practice (back osmosis, electroflotation, electrocoagulation, desalination of brines by freezing out,
solvent extraction techniques, etc.), and the promise each of these methods appears to hold for practical
applications, and the cost feasibility of the methods, are estimated.
The fourth, and most capacious, chapter cites extensive information on the use of sorbents of the
most widely varied classes in cleanup flowsheets for removal of radioactive contaminants from water and
effluents. Here we have to take note of the successful (in our view) arrangement of factual material in this
chapter: the description starts off with presentation of basic information on the structure of each particular
sorbent, its ion-exchange capacity, and the reasons governing the range of practical application of the
sorbent, and ends with arrays oflactual data on the effectiveness of using the sorbents in particular flow-
sheets.
*Atomizdat, Moscow (1974).
87
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One very obvious advantage of this monograph is the conciseness of presentation of the enormous
wealth of factual material that has been accumulating up to the present writing. We should also point out
that all of the material is systematized and reviewed in a successful manner, and that each chapter ends
with a critical resume of the situation by the authors. The monograph can be approached as a valuable
scientific text and reference manual for specialists concerned with cleanup of water to remove every con-
ceivable type of (radioactive) contaminants. The authors' own data, obtained in the course of many years
of research on processes designed to remove radioactive contaminants from water, will be of unquestion-
able interest to readers and users of the reference.
The value of the book is also evident in the close attention given to current problems pertaining to
protection of the natural environment from radioactive contaminants emanating from whatever sources.
88
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