SOVIET ATOMIC ENERGY - VOL. 37, NO. 1

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• Document Page Count: 118
• Document Creation Date: December 27, 2016
• Document Release Date: February 20, 2013
• Sequence Number: 1
• Publication Date: January 1, 1975
• Content Type: REPORT

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Russian Original Vol. 37, No. 1, July, 1974 ,.~ r
January, 1975
SATEAZ 37(1) 679-790 (1974)
SOVIET
ATOMIC
ENERGY
ATOMHAA 3HEP1'IAA
(ATOMNAYA ENERGIYA)
TRANSLATED FROM RUSSIAN
n
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, INSPEC
Physics Abstracts and Electrical and Elec-
tronics Abstracts, Current Contents, and
Nuclear Science Abstracts.
Soviet Atomic Energy is a cover-to-cover translation of Atomnaya
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
A. A. Bochvar
N. A. Dollezhal'
V. S. Fursov
I. N. Golovin
V. F. Kalinin
A: K. Krasin
A. I. Leipunskii
V. V. Matveev
M. G. Meshcheryakov
P. N. Paler
V. B. Shevchenko
V. I. Smirnov
A. P. Vinogradov
A. P. Zefirov
Copyright ? 1975 Plenum Publishing Corporation, 227 West 1 7th Street, New York,
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0
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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 Energiya
January, 1975
Volume 37, Number 1
. July, 1974
CONTENTS
Engl./Russ.
ARTICLES
Fast Gas and Thermal Breeder Reactor's - S. M. Feinberg. I ............. .. 679 3
New Prospects in the Creation of Gas-Cooled Fast: Reactors with a Short Doubling
Time, using Dissociating N204 - A:' K. Krasin and V: B. Nesterenko ...::... 687 11
Radioactive Safety Barriers in Nuclear Power Stations - E. P. Anan'ev
and G. N. Kruzhilin .....................................:.... 699 22
Low-Temperature Specific Heat and Thermodynamic Functions of Uranium Beryllide
- 0. P. Samorukov, V.'N. Kostryukov, F. A. Kostylev, and V. A. Tumbakov. 705. 28
Increasing the Efficiency of Separating Isotope-Cascades by Using Steps with-More
than Two Emergent Flows - N. A. Kolokol'tsov ....................... 709 32
Production of Ultracold Neutrons in a Stationary (Steady-State) Reactor of the
VVR-K (Water-Cooled, Water-Moderated) Type - E. Z. Akhmetov,
D. K. Kaipov, V. A. Konks, V. I. Lushchikov, Yu. N. Pokotilovskii,
A. V. Strelkov, and F. L. Shapiro ................................. 712 35
Optimizing Resonator-System Power when Accelerating Widely Spaced Bunches
- V. L. Serov, Yu. F. Orlov, and A. I. Baryshev ..................... 716 39
REVIEWS
Introduction of the Method of Preplanting. Gamma Irradiation of Seeds and the
Commercial Kolos Gamma Apparatus into Agricultural Practice
- N. M. Berezina, A. M. Kuzin, and D. A. Kaushanskii ................ 721 43
Sputtering of Matter by Fission Fragments - V. A. Bessonov .................. 730 52
ABSTRACTS
Electrical Conductivity of a Mullite-Corundum Ceramic at Elevated Temperatures
during Irradiation - U. G. Gulyamov, N. S. Kostyukov and A. P. Sokolov..... 735 57
Analytical Calculation of the Range of Ions and the Partial Loss of Energy during
Retardation - A. P. Balashov ................................... 735 57
Destructive Effect of Hydrogen on the Cladding during Reprocessing of the Fuel
Elements of a Wate r- Cooled /Wate r-Mode rated Power Reactor
- A. T. Ageenkov and V F. Savel'ev .............................. 737--- 58
Linear Perturbation Theory for Fuel Burnup Problems in a Fast Reactor
- V. V. Khromov, A. A. Kashutin and V. S. Glebov .................... 738 59
40K Gamma Distribution at the Ocean-Atmosphere Boundary - A. S. Vinogradov,
K. G. Vinogradova, and B. A. Nelepo ............................. 738 59
The Development of Activity Generators for Industrial Radiation Circuits with
Power-Generating Channel-Type Reactors - A. Kh. Breger, S. P. Dobrovol'skii,
E. L. Ivanter, V. S. Petrov, N. I. Rybkin, A. M. Sidorov, and Yu. I. Tokarev. 739 60
Determination of Fission Product Charge Distribution Parameters
- A. B. Koldobskii, V. Yu. Solovtev, and V. M. Kolobashkin ............. 741 61
Spatial and Spectral Distributions of Gamma Rays Reflected by Shields of Light
Materials - D. B. Pozdneev and M. A. Faddeev ...................... 741 61
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CONTENTS
LETTERS TO THE EDITOR
Perturbation Theory of Various Approximations in Steady-State Neutron
Transport- V. Ya.. Pupko.....................................
Effect of Reactor irradiation on the Dissociation Pressure of Zirconium Hydride
- V. S. Karasev, V. G. Kovyrshin, and V. V. Yakovlev ................
Optimal Bringing of Reactor into the Stationary Regime of Fuel Reloading
- B. P. Kochurov ..........................................
Operational Measurement of Biomedical Proton Beam Intensity - Ya. L. Kleinbok
and V. M. Narinskii ... ................... .............. .
Lead Transmission Curve for 31-GEV Electrons - A. S. Belousov, E. 1. Malinovskii,
S. V. Rusakov, S. P. Kruglov, and V. D. Savel'ev ....................
Radiation and Thermal Tests of Electron-Emissive Neutron Detectors and Cables
with Magnesia! Insulation - I. Ya. Emel'yanov, V. I. Vlasov, Yu. I. Volod'ko,
S. G. Karpechko, L. V.Konstantinov, V. V. Postnikov and V. I. Uvarov .....
The Problem of Calculating a Polycell in P3-Approximation - A. D. Galanin,
V. V. Smelov and B. Z. Torlin ...... ...... ..................
Measurement of Certain Characteristics of 249Bk - V. M. Glazov, R. I. Borisova,
and A. I. Shafiev .................. ...... .................
Use of Radiative Capture of Thermal Neutrons to Determine the Ash Content of Coal
- L. P. Starchik and Yu. N. Pak...................................
INFORMATION
Black Holes Are Real - N. A. Vlasov ....................... ..........
INFORMATION: CONFERENCES AND MEETINGS
35th Session of the Academic Council of the Joint Institute of Nuclear Research
- V. A. Biryukov ..... ...................
Conference of the IAEA Group on the Design of Fusion Reactors - G. N. Popkov.....
Seminars and Exhibitions of the All-Union "Izotop" Corporation ................
INFORMATION: NEW INSTRUMENTS AND FACILITIES
KTN-2 Tantalum and Niobium Concentration Meter - B. E. Kolesnikov,E. D. Kokhov,
Yu. D. Lavrent+ev, and A. P. Khukhyanskii ....... .... ..........
Transportation and Reloading Container for Remote-Controlled Gamma Therapy
Instruments - V. T. Emel'yanov, V. M. Kondrashov, and M. P. Sinodov .....
New Gamma Flaw Detectors for Testing the Welding of Main Pipelines
- V. N. Khoroshev, A. I. Murashov, V. N. Polosatov, and N. I. Neustruev....
BOOK REVIEWS
Yu. M. Dymkov. The Nature of Pitchblende - Reviewed by V. L. Barsukov .......
B. I. Spitsyn and B. B. Gromov. Physicochemical Properties of Radioactive Solids
- Reviewed by E. V. Sobotovich ......................... ... .....
A. Thomas and F. Abbey. Calculation Methods for Interacting Arrays of Fissile
Materials Reviewed by I. A. Stenbok ..................... .... ..
The Russian press date (podpisano k pechati) of this issue was 6/27/1974.
Publication therefore did not occur prior to this date, but must be assumed
to have taken place reasonably soon thereafter.
Engl./Russ.
742
63
745
65
747
66
751
69
754
71
756
72
762
76
764
78
766
79
769
81
772
82
777
85
779
86
780
87
782
89
784
90
787
93
789
93
790
94
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ARTICLES
S. M. Feinberg* UDC 621.039.524.526
Designs for breeder reactors with molten metal coolants (sodium) have by now been widely dissemi-
nated. Several such experimental and experimental-industrial reactors are already working or are being
brought into operation in the Soviet Union, the United States, Britain, and France.
However, in addition to the sodium breeder reactors there are also plans for other aspects of breeder
development including those of the fast gas and thermal thorium types.
The current phase in the development of nuclear power is characterized by high rates of increase in
power. The time for doubling the powers of nuclear installations is estimated as 3-5 years in the 70's and
5-7 years in the 90's. The fast breeders with liquid-metal (sodium) coolants now being built or designed
doubling time of over 7-10 years and therefore cannot lead to any significant economy in natural
?3years
1015 0
Year
Fig. 1. Demand for natural uranium (GU)
in the development of nuclear power, using
various types of reactors (GU represents
the reserves and cost of natural uranium).
1) Total development of electrical power; 2)
nuclear power stations of all types; 3) LWR;
4) HWR; 5, 6, 7) breeder reactors with a
doubling time of 7, 5, and 3 years respec-
tively.
uranium until the beginning of next century. Figure 1
shows the natural uranium requirements for nuclear
power using thermal reactors of various types and fast
breeders with various doubling times (T2) [11. The eco-
nomy of natural uranium depends on changes in the
doubling time of fast breeders: only for short doubling
times (of the order of three years) can there be a sub-
stantial economy in natural uranium before the year
2000. It is true that after the year 2000, with the reduc-
tion in the tempo of nuclear-power development, the
effectiveness of fast breeders may undergo a relative
increase; however, thermonuclear reactors may also
enter into the power arena, and even independently of
these the conditions and demands relating to the develop-
ment of power may change so substantially as to make
present predictions entirely out of date. We can hardly
look into the future further than 2010. Finally it must
be emphasized that a moderate saving of natural uranium
may from the economic point of view be completely in-
sufficient in order to compensate for the probable in-
crease in capital expenditure associated with the transi-
tion to fast breeders. This means that any shortening in
the doubling time of fast breeders will also increase their
competitivenesg.
We should note that the demand for uranium, after
reaching a maximum, will rapidly start falling; this
means there will be an excessive production of secondary
Translated from Atomnaya Energiya, Vol. 37, No. 1, pp. 3-10, July, 1974. Original article sub-
mitted September 11, 1973.
?1975 Plenum Publishing Corporation, 227 West 17th Street, New York, N. Y. 1001]. 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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fuel in fast breeders. The excess will have to be directed into
thermal reactors, particularly "peak" nuclear power stations,
which will ensure an economically satisfactory coexistence of
breeders and thermal reactors. If 30% of the power is provided
u~ro^ioMo by fast gas breeders with a doubling time of 3.2 years, the re-
maining 70% of the power may be obtained from traditional ther-
m I Ai ne wMo u?(ee) mal reactors working in the Th and 233U fuel cycle. The power-
doubling time in such a combined nuclear power system will be
10 years.
It is thus clear that in the coming quarter of a century, fast
4 reactors will only be able to become the basis of nuclear power if
short doubling times, of the order of 3-years, can be achieved.
Ideology of "Hard" and "Soft"
Fig. 2. Arrangement of a reactor Neutron Spectra in the
with a "plutonium" active zone. 1)
First screen; 2) second screen; 3) Active Z o n e
first end; 4) second end. It is well known that there are two "limiting" solutions of
the fast breeder.
1. A physically large active zone with M2 > 1. The fission of 238U takes place mainly inside the active zone. The neutron
spectrum in the active zone is soft, i.e., the average energy of the neutrons producing fission is of the
order of 0.1 MeV [1].
2. A physically small active zone with M2 R2 in which the plutonium is slightly diluted with oxygen
(or graphite), construction materials, and coolant, so that Eln/I;f 1. The fission of 238U then occurs in
the reflector zone adjacent to the active zone. The neutron spectrum is hard (mean energy ^-1 MeV) [2, 31.
In the first case the conversion ratio is lower and more sensitive to the increment a = ac /af . How-
ever, the dilution of the plutonium in the active zone enables the thermal loading of unit plutonium to be in-
creased and an internal "capacity" (for storage space) to be formed for storing a large quantity of fission
products, i.e., for increasing the depth of burn-up of the nuclear fuel. It is also possible to choose a com-
position of the active zone for which K.. changes little in the course of burn-up.
The first scheme describing the physical make-up of the active zone we shall subsequently call "di-
lute" and the second one "plutonium." We emphasize that the "plutonium" scheme in no way limits the size
of the active zone or its unit power. In order to increase the size of the active zone (or its power), we
must set up several individual plutonium zones inside the large reflector. Depending on the "clearance"
(thickness of the reflector), physical interference may arise between the individual modules: for a clear-
ance of the order of 300-500 mm this interference becomes insignificant. We shall call an active zone
made up of systems of individual plutonium zones "multiplutonium." A plutonium active zone is shown in
Fig. 2, together with the surrounding screens (breeding zones). The latter are situated in a vessel with a
diameter of around 2 m. A multiplutonium active zone is illustrated in Fig. 3. The diameter of the vessel
increases to 3.5 m and the power by an order of magnitude.
Up to the present time there has really not been any convincing analysis of the relative advantages
and disadvantages of the fundamentally different solutions of "dilute, " plutonium, and multiplutonium active
zones in breeder reactors.
Progress in Breeder Reactors with Gas Coolants
Considerable advances may be achieved by using gas coolants: The possibility of excluding chemical
corrosion and so achieving higher temperatures, simplicity of cleaning, and convenience inuse, is attracting
more and more attention toward gas coolants. We must accordingly regard gas as a coolant offering
new prospects for the progress of fast reactors with a short doubling time.
Possible gas coolants include helium, carbon dioxide, and so-called dissociated gases of the N204
type [4]; the latter possess a great heat capacity at moderate temperatures. Most authors now accept
steam coolants as a potentially weaker rival.
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The fundamental advantage of helium lies in the fact
-1i -, U_U+J%no that its interaction with the cores of the fuel elements, their
U+10%Mo coverings, and other construction materials is minimal. The
PU02 direct influence of helium on the physics of the reactor is
U+10%110 greatly reduced by virtue of the construction materials and
Fig. 3. Scheme of a reactor with a
"multiplutonium" active zone.
diluents introduced into the active.zone and the reflector of
the reactor so as to ensure a high specific heat take-off from
the nuclear fuel.
The introduction of sodium coolant into the active zone
slows the neutrons and greatly reduces the conversion ratio,
by -0.15. The conversion ratio is also greatly reduced by
the incorporation of construction materials such as iron and
iron alloys in the active zone, and also by slowing-down in
oxygen, since, in order to achieve a deep burn-up, the nu-
clear fuel is usually employed in the form of its oxides.
Helium may be used if a large heat take-off is achieved
from each unit of nuclear fuel, since without this the doubling
time will be greater even for a very high conversion ratio.
It is well known that the limiting factor for gas coolants is
the consumption of power in circulation and the pressure
drop around the circuit. In order to achieve pressure drops
and circulation power consumptions acceptable from the
engineering point of view, a high pressure of the working
gas is essential. On the other hand, it is desirable to mini-
mize the length of the working part of the active zone in order to reduce the pressure drop and also as far
as possible to reduce the diameter of the high-pressure vessel, if a metal vessel is in question, so that as
far as possible the dimensions of the active zone and the side reflectors should be reduced.
It follows from Table 1 that a gradual, significant progress may be achieved in the parameters of
nuclear power stations by raising the temperature and pressure in the system.
If the diameter of the vessel is not very much in excess of 2 m, the creation of such a vessel with a
pressure of 300 abs. atm presents no serious technical difficulties. For diameters no greater than 4.5-
4.0 m the achievement of pressures up to 55 abs. atm, using multilayer steel vessels, should be perfectly
reasonable in the next 15-20 years. Raising the surface temperature of the fuel elements to 800?C is also
a reasonable expectation. It is very important to emphasize the relationship between the fuel efficiency of
atomic power stations and the doubling time. For example, if T2 increases from 3.2 to 4.5 years as a
TABLE 1. Effect of Gas Pressure and the Temperature of the Fuel Elements on the Main
Parameters of Nuclear Power Stations with Fast Gas Reactors Having a "Plutonium" or
"Multiplutonium" Active Zone
Main parameters of the nuclear
power stations
300
Power used in circulating the heliur
(6% of the thermal power of the
circuit)
Temperature, ?C
at the inlet
210
265
305
250
335
390
at the outlet
420
440
460
440
525
550
Steam pressure before the turbine,
30
80
180
65
180
240
abs. atm
Efficiency,
gross
30-31
37-39
42-44
35-37
42-44
46-48
net
27-28
32-34
38-40
32-34
38-40
42-44
Conversion ratio
2,05
2,05
2,05
2,05
2,05
2,05
T2, *years
3,2
3,2
3,2
3,2
3,2
3,2
*For a 30% burn-up of the plutonium in the active zone the time of the campaign is 0.7 of a year; the depth of
accumulation of plutonium in the metal reflector is 10 kg/ton and the time of the external fuel cycle is 0.5 of
a year.
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TABLE 2. Some Characteristics of the MSBR Reactor
Without separation
of protactinium
With separation
of protactinium
Conversion ratio
1.0538
1.074
Loss of fuel in reprocessing
0.006
0.007
Yield of secondary fuel per atom of fuel
4.96
7.95
passing into the reactor,
Number of secondary neutrons arising per
consumed atom
Specific loading, MW (thermal)/kg fuel
2.89
3.26
Cost of fuel cycle,' mills/kW ? h
0.45
0.33
Doubling time for the system of reactors, years
14
8.7
result of a 40% fall in the thermal loading of the active zone, the helium inlet and outlet temperatures may
be raised to 340-500?C, and this will enable the thermal efficiency to be raised to -407c.
Fuel Element Problems
Metallic uranium and uranium alloys limit the depth of burn-up to -2%. Hence this solution to this
problem is usually sought by employing uranium and plutonium oxides or carbides. Experiments indicate
that oxides give a depth of burn-up of -40%. It is reasonable to expect that in loose fuel elements (uranium
or plutonium dioxide poured into a steel can) with a can thickness of 0.1-0.2 mm, a burn-up of ^-50% may be
achieved if the can is freed from the internal gas, pressure of the fission products, and if damage by con-
tact with the coolant is prevented. This latter is achieved by using an inert gas as coolant, and the former
by creating an oxide core of low density (-7.5 g/cm3) and a more or less free outlet for the gaseous fission
products from the fuel elements (for example, by incomplete sealing at the ends of the fuel elements). Thus
it is essential to allow for the outflow of radioactive gaseous fragments, with a certain time lag, into the
coolant circuit. Experience shows that the same applies to reactors with liquid-metal cooling, and also to
thermal reactors by virtue of the dehermetization of the fuel-element cans on irradiation.
Estimates show that, on removing the radioactive noble gases and iodine, amounting to 0.5ic of the
flow of coolant, the radioactivity of the circulation circuit is less than 0.1 Ci/liter (the usual norm for
water-cooled water-moderated reactors), which is 2-3 orders of magnitude smaller than in a sodium
circuit.
At first glance it might seem that, for a very deep burn-up (30-50%), the fuel-element can will suffer
extremely serious radiation damage from the action of the fast neutrons. Calculations show that this da-
mage is not only no greater but (unexpectedly) less than in breeders with a sodium coolant.
In order to reduce the amount of diluents and construction materials as much as possible, metallic
uranium should be used in the reflector, while the active-zone fuel elements in the steel cans should be
filled with plutonium dioxide only. In this way the amount of oxygen in the reactor is reduced by an order
of magnitude. It has been suggested that on slightly alloying the metallic uranium with molybdenum or zir-
conium the depth of accumulation of fission products may be increased to 0.8-1% without serious swelling
of the fuel elements. Fuel elements with very thin cans may then successfully operate in the reflector
zones directly bordering the active zone (in layers 10 cm thick). In the case of the end reflectors, if from
constructional considerations it is impossible to avoid a simultaneous replacement of the cassettes of the
active zone and the end reflector, it may be necessary to use uranium substantially alloyed with molyb-
denum, having a 2-3% depth of burn-up.
In many designs of gas breeder reactors with a "diluted" active zone it is proposed to use steel tubu-
lar fuel elements with a diameter of about 6 mm and an artificially rough surface in order to intensify the
heat take-off by the gas coolant. The cores of such fuel elements are either formed from tablets composed
of the sintered dioxide of 238U and plutonium or else filled with spheres of the sintered dioxide [5].
The production of a high volumetric energy intensity in reactors of large dimensions, with longitudi-
nal cooling of the rod-type fuel elements, involves large pressure drops and substantial losses of power in
pumping of comparatively moderate pressures. It is precisely because of this that, with rod-type fuel
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Fig. 4. Schematic sec-
tion of a cell in the active
zone of a salt thermal
breeder. 1) Tubes for
fuel; 2) upward motion of
the fuel; 3) graphite; 4)
slot apertures; 5) down-
ward motion of fuel; 6)
moderator unit; 7) fuel
inlet; 8) fuel outlet.
elements 5-6 mm in diameter, it is impossible to obtain a gas reactor with
characteristics much better than those of liquid-metal breeeders. It is
essential to find fuel elements with more developed specific surfaces, and
more efficient ways of "impelling" the gas through the reactor. The way
out may be found by using spherical micro-fuel elements, so creating a
longitudinal -transverse motionof the gas in the cassette. A substantial
increase in the cooling surface area, an increase in the maximum per-
missible surface temperature to 1000?C, and the longitudinal and trans-
verse motion of the gas should allow the volumetric loadings to be in-
creased to 1000 kW/liter for a gas pressure of 150 abs. atm. It is con-
sidered quite possible to heat the gas to 750-800?C. This opens the attrac-
tive prospect of replacing steam turbines by gas turbines and completely
eliminating the construction of a second circuit for the circulation of the
working substance. Hence the development and adoption of micro-fuel
elements is of great interest for the further development of fast gas
breeder technology.
Problem of Safety
The problem of the safety of a fast gas breeder working at high
pressures and very great specific thermal loadings in the active zone is
of vital importance. If leaks occur through gaps of less than 100 cm2 in
the circuit, the shut-down cooling of the active zone may be simply ef-
fected by a multiple-loop consideration of the circulatory circuit (at least
four loops).' A complicated problem is that of ensuring safety in the
improbable emergency of an instantaneous and complete rupture of the
pressure vessel, leading to the flying apart of fragments of the latter.
In the overwhelming majority of designs it is proposed that the reactor
vessel be made out of prestressed reinforced concrete. The air-tight-
ness of such vessels depends in no small measure on the reliability of the
large steel flanges and the covers for the steam generators and gas blowers
built into the stressed concrete. Through these covers pass all the many
conduits for the water and steam pipes, power cables, and so forth. Thus
the weak points of these vessels include the steel covers and flanges with
diameters of > 4 m. In order to strengthen these dangerous components
for pressures of > 60 abs. atm it is proposed to use a system of double
cover plates. The vessel is thus transformed into a cumbersome, com-
plicated, and expensive construction: for a reactor with a power of 1
million kW (el.) and a pressure of 120 abs. atm the diameter reaches
- 32 m, the height - 40 m, and the volume around 30,000 m3. The cost
of such a vessel (together with the thermal insulation) is about 20% of the
cost of the whole nuclear power station, i.e., 40 million dollars or over.
In a design proposed for a fast helium breeder with small plutonium
active zones, another method of ensuring high strength of the vessels, based
on the idea of spreading the stresses, is employed: the high-pressure vessel is placed inside another hermetic
vessel of intermediate pressure, designed to accommodate the whole of the working substance if the main vessel
ruptures. The low pressure so created (' 10-30% of the working pressure of the first circuit) is sufficient to en-
sure proper shut-down cooling of the active zone when the emergency protective system operates. Undernor-
mal use the second intermediate-pressure vessel does not carry any pressure, and a low temperature is
maintained in this vessel by means of a cooling system and thermal insulation. This vessel is designed
solely for short-term operation under emergency conditions.
In order to prevent any serious flying apart of pieces of the high-pressure vessel after its possible
instantaneous rupture, strong horizontal bandages and supports for the vessel covers are provided, limit-
ing the flight of pieces of the high-pressure vessel to gaps 1-2 cm wide. These are designed for briefly
accepting the pressure of the working substance.
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Fig. 5. Arrangement of the circulation loop of a thorium
thermal breeder. 1, 4) Salt-liquid mixers; 2) decay of
Pa; 3) liquid bismuth; 5) removal of fission products; 6)
electrolyzer; 7) bred uranium; 8) purified salt; 9) reactor;
10) cooling liquid salt; 11) turbogenerator; 12) steam.
In our own opinion the provision of bandages to limit the flight of vessel fragments is unnecessary,
since no such construction should be required in the presence, of a continuous monitoring system, prevent-
ing the appearance of dangerous cracks in the high-pressure vessels. This constitutes a reliable guaran-
tee against the sudden rupture of the high-pressure vessel. Methods of nondestructive testing for verifying
the state of the pressure vessels are rapidly being perfected and will undoubtedly reach the required level
in the near future.
Problem of the Composition of the Reactor
The widely expressed opinion that an increase in the unit power of the reactor is the best way of im-
proving the economy of a nuclear power station is not, in our own opinion, entirely obvious. We consider
that it is perfectly possible to conceive powerful nuclear power stations consisting of a large number of
individual (module) reactors, relatively small as regards unit power, linked by a single gas circuit and
sundry provisions.
What makes us say this ?
The conversion ratio of a reactor with a physically small (plutonium) active zone reaches the maxi-
mum possible value for fast reactors. In this respect the situation in fast reactors is the opposite of that
in thermal reactors, in which an increase in reactor dimensions leads to a rise in the plutonium coefficient
and to a fall in the fuel component.
A. reactor with a large unit power requires a large (once-only) insertion of plutonium and an assimila-
tion of the technology of large high-pressure vessels, creates difficulties in the reservation of power during
reactor shut-downs for recharging, and requires the creation of a very complicated system of recharging
while running (one of the reactors of a nuclear power station containing, say, ten reactors maybe under-
going recharging at any given time). All this inevitably complicates the design, erection, and development
of the primary installations, and makes the creation of the new technology less rapid and more expensive.
In our own opinion, it is precisely the expenses and delay involved in mastering the new technology that
constitute the chief reason for its retarded introduction. It would therefore appear desirable to seek solu-
tions to the problem in such a way that the unit outlay of plutonium and the unit powers should be compara-
tively small and not require the development of new technology for the erection of large high-pressure
vessels. It does not follow that a reduction in the unit power of the reactor will lead to a considerable in-
crease in the capital outlay for an established 1 kW of electric power. The capital outlay on a nuclear
power station is mainly made up of expenditure on the coolant circulation system, on the cassettes, heat
exchangers, gas provisions, the equipment of the second circuit, the machine room, the reactor building,
the radiation shielding and purification system, the preparation of the site, and so forth. The cost of the
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high-pressure steel vessels and the actual reactors amounts to some 10-15%. Hence a nuclear power sta-
tion composed of several (5-10) smallish module reactors may not be inferior as regards capital outlay
to a nuclear power station of equal power consisting of a smaller number of reactors (one or two).
However, the module station will be simpler in its adoption, considerably more reliable in use, and
less subject to radiation hazard: ten modules of a total power of 1 million kW (electric) are sited in pockets
within a single reinforced concrete unit with a total volume of ^-20, 000 m3.
The actual economic efficiency of the "breaking down" of the nuclear power station, i.e., the use of a
larger number of individual small reactors, constitutes a problem for future study.
Hence we should at the same time consider versions with a multiplutonium active zone, assuming the
use of a high-power reactor, while retaining physically small active zones and all the physical indices asso-
ciated with these.
At the present time designs for fast gas breeders are being extensively published in the literature of
other countries. Common to all these is a scheme based on the reactor having a high-power, diluted active
zone of -103 MW (el.). Preferentially proposed is an integrated composition in a large stressed reinforced
concrete vessel with a coolant pressure of 60-120 abs. atm. It is proposed to make either ordinary fuel
elements in the form of steel tubes with a uranium and plutonium dioxide core or else micro-fuel elements
with graphite-based coatings. The reflectors are of the same type, but without plutonium; the conversion
ratio is between 1.3 and 1.5.
According to the published economic estimates, the capital outlay is either 10-20% smaller than that
of sodium breeders or else approximately the same. The cost of the fuel cycle is 20-30% higher than that
of sodium breeders, and differs little from the values obtained for water-cooled, water-moderated reac-
tors. The doubling time is considerable: ten years or over.
A Soviet design based on a similar conception was recently published [2], except that in this case the
doubling time was 5-6 years and the helium pressure was rather greater (150 abs. atm). The shortening
of the doubling time is here motivated by considerations of the economy of raw materials, i.e., an increase
in the cost of mining uranium is envisaged.
Thermal Breeders with a Thorium Fuel Cycle and
Circulating Nuclear Fuel
When considering gas breeders, the importance and effectiveness of increasing the conversion ratio
is always emphasized; for a reactor consisting of a purely plutonium core surrounded by pure 238U this may
theoretically reach 2.5. Under practical conditions the conversion ratio hardly exceeds 2.0; in sodium
breeders being constructed or designed it is -1.5.
In contrast to this, it is possible to make breeders with a conversion factor of little over unity (1.05-
1.10) and a comparatively short doubling time. This arises from the fact that the doubling time for the
amount of nuclear fuel is mainly determined by two parameters: the conversion ratio and the energy inten-
sity of the nuclear fuel.
In thermal reactors the degree of dilution of the nuclear fuel is several orders of magnitude higher
than in fast breeders. Hence there may be tens of times more intensive a heat release in the former as in
the latter. Furthermore, there is also the possibility of taking a fundamentally new and important step:
that of abandoning solid fuel elements, the use of which is accompanied by a complicated process of radio-
chemical and metallurgical processing, and organizing the circulation of the nuclear fuel in the form of a
salt melt. A number of the reprocessing procedures of the thermal cycle then fall away. Regeneration is
considerably simplified, the necessity of ensuring a high degree of purification of the fuel from the fission
products is eliminated, and so forth. It is true that every power reactor is thus converted into a radio-
chemical undertaking, and of course the problem of burying and removing the radioactive waste is much
more complicated for radiochemical factories than for reactors.
However, in principle this problem is entirely soluble.
Thus if the (CR-1) of a fast breeder amounts to 0.5-1.0 and the (CR-1) of a thermal breeder to only
0.05-0.10, it is sufficient to increase the energy intensity of the fuel by a factor of 10 in order to achieve
the same doubling time.
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A reactor with circulating liquid nuclear fuel enables us to reach the required energy intensity.
Fuel Cycle of a Thermal Breeder
Since of all the known fuel cycles of thermal reactors only a composition of 233U an d 232Th yields a
conversion ratio of greater than unity, it is precisely this fuel cycle which is used in the thermal breeder.
Here a new nuclear raw material is introduced into the power system, thorium, and this greatly increases
the potential resources of nuclear fuel, since thorium is several times more widespread in the Earth's
crust than uranium. In designing fast breeders the possibility of drawing thorium resources into the power
system is also envisaged.
A great advance in the development of designs for such thermal breeders was the initiation in 1966
of a small experimental MSBR reactor with a thermal power of 8000 kW in Oak Ridge National Laboratory,
USA. This reactor has provided a great deal of working experience, on the basis of which designs of reac-
tors with powers of up to 1 million kW (el.) have been developed.
The active zone of the reactor is formed by a system of vertical graphite tubes (Fig. 4) serving as
moderator and construction material. A liquid salt melt containing 233U passes into the graphite tubes from
metal collectors placed under the active zone. The melt is carried up through apertures in the graphite
and then down through a central graphite tube into the outlet metal collectors. Passing through the active
zone, the 233U sustains' a chain reaction and heats the salt melt. The usual temperatures are 560?C at the in-
let and 700?C at the outlet.
At the reactor outlet the melt passes into metal heat-exchangers in which the heat is transferred to
the working substance of the turbine. The active zone is surrounded by a graphite reflector siting the
232Th. The leakage neutrons are absorbed by the 232Th, forming a new nuclear fuel, 233U. The radiochemi-
cal system, continuously acting in the bypass of the liquid nuclear-fuel circulation loop, separates the fis-
sion products and protactinium, and releases 233U and 232Th to make up the nuclear fuel (Fig. 5).
Table 2 shows the indices of the MSBR.
Thus fast gas breeder reactors may have a somewhat greater conversion ratio than liquid-metal fast
breeder reactors, and a coolant technology which is simpler in use. However, these reactors operate at
considerably higher pressures and require a great expenditure of energy in circulation. In principle such
reactors open the way for a transition to the single-loop system.
These reactors may exceed LMFBR (liquid metal fuel breeder reactor) as regards economic indices,
shorten the doubling time, and thus solve the problem of uranium raw material.
Thermal breeders with the circulation of liquid nuclear fuel and with a uranium-thorium fuel cycle
have a very small CR-1 factor (-0.05-0.10), but a high specific power, and may therefore also have a short
doubling time.
1. S. Feinberg, Cooled Fast Reactor, Techn. Rep., IAEA-154, Vienna (1973), p. 21.
2. N. Ponomarev-Stepnoi, ibid., p. 191.
3. P. Forteskye, ibid., p. 63.
4. A. Krasin, ibid., p. 89.
5. M. Donne, ibid., p. 267.
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NEW PROSPECTS IN THE CREATION. OF GAS-COOLED
FAST REACTORS WITH A SHORT,, DOUBLING TIME,
USING DISSOCIATING N204
A. K. Krasin and V. B. Nesterenko UDC 621.039.534
Future rates of development and economic characteristics of nuclear power will be largely deter-
mined by the rate of accumulation of plutonium. It is therefore an extremely urgent matter to solve the
problem of ensuring the automatic supply of plutonium for developing nuclear power systems involving
fast reactors [1], and also to create fast reactors with physical characteristics such as will advance the
construction of nuclear power stations with fast reactors having a doubling time of 4-5 years. This would
increase the rate of growth in the powers of nuclear power stations by a factor of two to four as compared
with that of thermal power stations.
Allowing for the small part now played by nuclear power in the total power-supply system of the USSR
and the need to increase this proportion substantially by the year 2000 (to 30-50% of the total output of all
power stations [2, 4]), it was shown in [4, 5] that in 1980-1990 the doubling time for the powers of nuclear
power stations should amount to 4-5 years, and in 1990-2000 to 6-7 years.
Further improvements in the physical characteristics of fast reactors may be achieved by raising the
specific thermal stress of the active zone and also the specific concentration of the nuclear fuel (for ex-
ample, by using low-alloy uranium, in which the nuclear concentration is twice as great and the thermal
conductivity an order of magnitude higher than in the case of U02). However, in such alloys there is a
characteristic radiation-induced swelling of the fuel elements. and a strict limitation is imposed upon the
maximum fuel temperature (no higher than 630-650?C). The heat-takeoff efficiency on the gas side may be
ductivity than U02 requires a reduction in the maximum temperatures of the coolant to 400-500?C for matrix
composites and 270-350?C for slightly-alloyed metallic uranium. This leads to a fall in the efficiency of the
nuclear power station by comparison with the accepted parameters of sodium breeder reactors. For nu-
clear power stations with fast reactors the main proportion of the cost of electrical power comes from the
capital outlay on equipment. For these stations it is therefore justifiable to seek simpler arrangements
for converting the heat, and also ways of reducing the equipment costs.
A most important problem is that of creating a strategy for the development of nuclear power such as
will ensure the minimum consumption of natural uranium. These problems may be solved if we make fast
reactors with a specific thermal stress of 800-1200 kW/liter, and use thermal parameters of the thermal
cycle of the atomic power station such as will allow the use of low-alloy metallic uranium with a maximum
fuel temperature of 650?C (allowing for overheating factors). It is therefore necessary to seek new coolants
and to develop new systems for heat conversion in atomic power stations working at temperature levels of
250-450?C [7].
raised by increasing the pressure; thus, in order to secure specific thermal stresses competitive with
sodium in the case of helium reactors, a pressure of 250-300 abs. atm must be used; for CO2 the corre-
sponding pressure is 220-250 and for N2O4 140-160 abs. atm [6].
Hence in fast gas-cooled reactors with a high specific thermal stress and a large external pressure
using U02-base fuel it is preferable to prepare the fuel not in the form of pure U02 but in matrix form,
which has a thermal conductivity two or three times greater, although the maximum temperature of such
fuel is lower (1300-1400?C) than in the case of UO2.
Thus in order to achieve high specific thermal stresses the use of fuel with a greater thermal con-
Translated from Atomnaya Energiya, Vol. 37, No. 1, pp. 11-21, July, 1974. Original article sub-
mitted April 1, 1974.
? 1975 Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011. No part of this publication ma}t 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 front the publisher for $15.00.
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L---- ------ 1
Fig. 2
Fig. 1. Scheme of a liquid-gas cycle with intermediate regeneration and one
take-off: 1) Fast reactor; 2) high-pressure turbine (HPT); 3) low-pressure tur-
bine (LPT); 4) electrogenerator; 5) condenser; 6) condensation pump; 7) supply
(feeding) pump; 8) regenerative preheater; 9) regenerator; 10) turbine of turbo-
PUMP
Fig. 2. Scheme of a gas-liquid cycle with intermediate regeneration: 1-7) see
Fig. 1; 8) regenerator; 9) turbine of feeding turbopump.
Thermophysical Characteristics of N204
One way of solving the problem of choosing a gas coolant for fast reactors may lie in the use of dis-
sociating nitrogen tetroxide (N204 2NO21 149 kcal/kg) in single-loop systems as both coolant and working
substance of the nuclear power station [7-12]. A complex study of the thermophysical properties of N204
has been carried out in the Soviet Union over a wide range of temperatures and pressures: specific heat at
1-170 abs. atm and 30-550?C, p-v-t properties at 50-5525?C and 8-150 abs. atm, enthalpy at 50-170 abs.
atm and 20-350?C, gas composition at 150-600?C and 1-7.5 abs. atm, viscosity at 30-500?C and 1-140 abs.
atm, and thermal conductivity at 30-540?C and 1-150 abs. atm [12]. Heat transfer has been studied during
condensation and boiling in a tube and on the flat surface; convective heat transfer has been studied at sub-
and supercritical pressures (1-150 abs. atm, 30-550?C) [11-14]. The thermal-radiation resistance has
been studied under static conditions for pressures up to 50 abs. atm and temperatures of 300-550?C [11,
12]. The corrosion resistance of construction materials has been studied in N204 under static conditions
at 150 abs. atm and 700?C and under dynamic conditions at velocities of 30-50 m/sec, pressures up to 50
abs. atm, and temperatures up to 550?C [11, 12, 15, 161.
The main factor involved in the favorable thermophysical properties of dissociating gases lies in the
existence of large thermal effects in both the first and second stages of the reaction. In experimental in-
vestigations into the isobaric specific heat of N204 = 2NO2 at temperatures of 160-300?C and pressures of
115, 130, and 150 abs. atm, large maximum values of the specific heat were obtained: Cp115 = 12 kcal/kg
-deg, Cp130 = 8 kcal/kg -deg, Cp150 = 5.0 kcal/kg ?deg (Fig. 1 in [11]); the mean specific heat for the range
160-280?C equals 1.95-2.1 kcal/kg -deg, - For comparison, we remember that for the same range of tem-
peratures the specific heat of CO2 is 0.28, that of sodium 0.31, and that of helium 1.243 kcal/kg -deg. Thus
the mean specific heat of N204 2NO is seven or eight times greater than that of CO2 and 1.6 times greater
than that of helium. In the temperature range 200-500?C the average specific heat of 2N02 2NO + 02 is
0.75-0.8 kcal /kg ?deg [11]. Owing to chemical reactions taking place in the dissociating gas, in a noniso-
thermal flow, in addition to the molecular specific heat there is also an additional transfer of a consider-
able amount of heat in the form of chemical enthalpy by concentration diffusion.
In experimental investigations into the heat transfer associated with the turbulent flow of a dissociat-
ing gas 2NO2 2NO + 02 in a heated tube at supercritical pressures of 115-160 abs. atm, temperatures of
250-530?C, thermal fluxes of 106 kcal/m2?h, and Reynolds numbers of Re = 2.105, heat-transfer coeffi-
cients equal to 18,000-38,000 kcal/m2 ?h - ?C were obtained.
In 1971 experimental work was carried out in relation to the convective heat transfer of a dissociat-
ing gas N204 2NO2 at temperatures of 165-180?C, pressures of 130-160 abs. atm, Re = (3-5) ? 105, and
thermal fluxes of (3-5.5) ? 106 kcal/m2 ?h in an experimental tube 4 x 1 mm in diameter and 800 mm long
(with electric heating), the diameter corresponding to the equivalent diameter in a. group of fuel elements
(deq = 2 mm). In these experiments, in half the length of the part under consideration a heat-transfer
coefficient at 150,000 kcal/m2 ? h ? ?C was attained, the lowest heat-transfer coefficients occurring in the
exit section and being about 30,000 kcal/m2 ? h ? ?C. On heating the dissociating gas N2O4 2NO2 from 165
to 280?C the maximum temperature of the can was no greater than 450?C.
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Thermophysical calculations carried out for fast gas-cooled reactors at pressures of 120-170 abs.
atm and a gas outlet temperature of 250-300?C showed the possibility of achieving a specific thermal stress
of 960-1200 kW /liter for specific thermal loadings of the, surface of the fuel elements equal to (3-5)-106
kcal/m2'h and a diameter for 6-7 mm. For dissociating gas pressures. of 130-160 abs. atm, gas outlet
temperatures of `450-500?C, and preheating temperatures of 230-270?C, efficient heat release may be se-
cured from active zone of a fast reactor with a thermal power of 3000-3400 MW for an average energy
stress of the active zone equal to 1000-1200 kW/liter and a maximum temperature of the fuel-element cans
(allowing for the overheating factor) of 650-680?C.
Experimental work has been carried out on the condensation of recombining gas 2NO2 + 02 2NO2
N2O4 in vertical and inclined tubes, and on vertical and horizontal surfaces at pressures of 1-9 abs. atm
and thermal loads of 0.64.103-3.1 ?104 kcal/m2 ?h. Average heat-transfer coefficients of 1000-1500 kcal
/M2- h ? ?C were obtained for mass velocities of yW = 2-40 kg/m2 ? sec. A considerable intensification of
the heat-transfer processes (by a factor of 2.5-3 times) may be achieved by condensing the chemically
reacting gas on finely-undulating ribbed (finned) tubes (horizontal or slightly inclined). In the course of
condensation there is a marked intensification of the heat-transfer processes on account of the evolution
of the heat of the chemical reactions [12].
Experimental work has also been carried out on the heat transfer associated with boiling both inside
a tube and also inside a large vessel with a pressure of up to 90 abs. atm and thermal loadings of (1-4) ? 105
kcal/m2 ? h. With such loadings, boiling in a large vessel gave experimental heat-transfer coefficients of
15,000-30,000 kcal/m2 - h -deg. A characteristic of the boiling of N2O4 is the fairly substantial dissociation
of N204 2NO21 which occurs chiefly at high pressures. Owing to the heat of the chemical reactions, dis-
sociation is accompanied by an intensification of the passage of heat to the vapor bubble as boiling takes
place through the boundary film of liquid. In addition to this, on passing from the liquid into the vapor
state there is a sudden increase in the degree of dissociation of the N204 molecules. Corresponding to this
process the thermal effect of the chemical reaction is added to the heat of vaporization, and thus intensi-
fies the process of heat transfer while boiling [11-14].
Thermal Schemes of Heat Conversion in Nuclear
Power Stations Using N204
An advantage of the dissociating N2O4 system is the possibility of achieving a gas-turbine cycle, so
as to be able to use the temperature-varying molecular weight of the gas (specific work of the turbine and
compressor) in order to produce a substantial increase in the effective efficiency factor and the specific
power of the cycle [17]. Nitrogen tetroxide has a low heat of vaporization, 5.5 times lower than that of
water. This simplifies the scheme of heat regeneration in the gas-liquid cycle, since the amount of heat
in the gases leaving the turbine is sufficient not only for heating and evaporating the liquid but also for
superheating the gas and regenerator by 100-200?C. _
At supercritical pressures the use of a regenerator or regenerator and preheater ensures the re-
generative heating of the coolant until it passes into the gaseous state, so that a gas-cooled reactor may
be used in a condensation cycle with both sub- and supercritical parameters.
The chemical-dissociation reactions occurring on the high-pressure side involve a lower heat of
chemical reaction (149 kcal/kg) than those which occur on the low-pressure side in the course of recom-
bination (294 kcal/kg); hence in gas-liquid cycles based on N204 we have a clear possibility of achieving a
higher degree of regeneration of heat in the cycle than in the case of water or C02, and hence better ther-
modynamic indices. As a result of the increased heat-transfer coefficients these processes may be ef-
fected in compact heat exchangers.
When studying the mechanism of the chemical reactions and the kinetic constants of the N2O4 2NO2
2NO + 02 system it was established that, in any gas-dynamic calculations of the flow parameters of the
thermodynamic cycle, the turbines, and the heat-transfer processes taking place in the reactor and heat
exchangers, it was essential to allow for the time characteristics of the dissociation and recombination
processes. Estimates of the times of chemical relaxation based on existing experimental velocity con-
stants of the chemical reactions showed that in the temperature and pressure ranges of practical impor-
tance the first stage of the reaction (N204 2NO2) took place in an equilibrium manner (10-6-10-8 sec),
while in the second stage of the reaction (2NO2 2N0 + 02) the chemical relaxation time might vary be-
tween 10-3-10-4 and 0.1-1 sec, depending on the thermodynamic and geometrical parameters [7]. For in-
termediate regeneration pressures of 20-25 abs. atm, total regeneration efficiency is achieved, and the
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TABLE 1. Results of Radiation Tests on Micro-Fuel
Elements
Thickness of SiC coating, p
Burn-ups achieved'' '~
Limiting temperature, ?C
Number of particles tested
Number of particles ruptured under
irradiation
50
11.7
925
500
3
effect, of the kinetic characteristics of the chemical reactions is almost eliminated;, in the high- and low--
pressure turbines the flow of gas is almost equilibrium [7, 12].
In the N204 gas liquid cycle, the efficiency of the nuclear power' station in the temperature range '450-
550?C and the pressure range 80-240 abs, atm is 4-6 abs. % higher than that of the water-vapor cycle (7].
For systems- with intermediate regeneration we may find an additional increase in the thermodynamic effi-
ciency on account of the introduction of the regenerative take-off processes. However; for nuclear power-
station schemes involving- fast reactors the greatest interest will lie in arrangements involving intermediate
regeneration and one take-off (Fig. 1),.. and 'also a simple scheme with-, intermediate regeneration (Fig. 2).
The pressure of the gas-liquid N204'cycle after the turbine is equal to 1.5-2 abs.' atm, while the spe-
cific volume of the N204 is 34-40 times less than that of. water vapor at the same condensation temperature
(ps = 0.04 abs, atm). This, enables us to. increase the .power of a single stroke of the gas turbine very sub-
stantially, bringing it up to,500-1000 MW. In addition to this, when using N204 the, flow section of the gas
turbine may be made with a small number of steps, since the heat differentials in N204 turbines are 2.2-
2.5 times smaller than in water vapor [12]. Gas-dynamic calculations and design developments for an
N204 gas turbine with a power of 500 MW at 280?C. and 150 abs. atm revealed the possibility of making
single-shaft two-flow turbines with a total weight of ? 250-270 tons and a length of 16-18 m [6].
The results of experimental investigations carried out at 25-550?C under thermal'conditions at 1-150
abs. atm and radiation-thermal conditions 'at 10-60 abs. atm showed that the,N204system had thermal and
.radiation resistance sufficient to allow its practical use in?nuclear power production. The thermal/radia-
tion' resistance of N204 was studied:.experimentally in the -field of n, y radiation produced by, a thermal-neu-
tron reactor in a flow-type loop installation operating at 30 abs. atm and 550?C, and the radiolysis of. this
coolant in the n,,yradiation field of a fast reactor was 'calculated. lathe dissociating system N204 2N02
2N0 + 02, by.virtue of the specific characteristics of fast reactors, radiochemical effects arise from the
action of radiation at a high dose rate (1019 eV /cm3 - sec) after contact times of 10-2 in the active zone,
the temperature varying over the range 160-280 or 200-450 (500) ?C and the pressure over the range 150-170abs.
atm. The radiolysis products of NO2 are N20, N2, and 02: The `proportion of decomposing coolant in a fast
reactor is about 10-6 [111.
A,great deal of attention has been paid to verifying the chemical stability of an N204 gas-liquid cycle.
Experiments have been carried out in a test-bed with a closed N204 gas-liquid cycle at '504?C and 5 abs.
atm (3s6 0 kg of N204 in the circuit) at a power of 1070 kW with, a gas flow of 1400 kg/h for 372 h. In a con-
tinuouexperiment lasting 170 h, more than looo'cycles involving the return of the dissociating N204 cool-
ant along the circuit were achieved on the principle of the sequence liquid-evaporation-heating-dissocia-
tion-cooling-recombination-condensation-liquid, and so forth. In these experiments a high thermal stability
of the gas parameters was achieved, no irreversibility of the coolant being'apparent [12]. In experiments
carried out over the period 1966-1971 in a closed gas-liquid loop for 600 h, . using .80 kg of N204, with
parameters of 30-550?C and 10=60' ails. atm, . complete reversibility of the gas-liquid cycle was, also -
achieved. Altogether' the'test-bed operated for more than 1500 h under working conditions, including 1000 h
without any change'of coolant. Chemical analysis''of'samples taken off. revealed: no marked change in the,
com'positionof the coolant (impurities in the circuit amounted to no .more than 0.8%) [12]..'.
In a loop installation working for 600 h under. radiation-thermal conditions at 10-30 abs, atm and
500?C; the stability of the gas-liquid cycle was further confirmed [11]. ,
These experiments indicated the complete reversibility of the gas-liquid cycle and the practicable
applicability of N204 as a working substance and coolant for nuclear power installations.,.
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 Declassified and Approved For Release 2013/02/21 : CIA-RDP10-02196R000400040001-6
TABLE 2. Characteristics of the Thermodynamic Cycle of an N204
Atomic Power Station
Gas temperature at the reactor inlet/outlet, ?C
166/280
200/450
Gas pressure at the reactor inlet/outlet, abs, atm
169/154
169/154
Thermal power of the reactor, MW
4190
3253
Power of the gas turbines, MW
1200
1148
Flow of gas through the reactor, kg /sec
5930
3853
Efficiency of turbine installation,%
28.6
35.3
Efficiency of nuclear power station (gross),%
27
34.5
Power of the turbine drive pumps, MW
80.7
Efficiency of the high-pressure turbine, %
88
88
Efficiency of the low-pressure turbine, %
87
87
Efficiency of the turbine pumps, %
80
.80
Minimum head in the regenerator, ?C
12
15
Minimum pressure in the condenser, abs. atm
2,06
2.06
Minimum temperature, ?C
35
35
Pressure of N204 beyond the pump, abs. atm
173
173
Proportion of coolant take-off to the preheater,%
20
Proportion of gas take-off to turbopump drive,%
1.5
Gas pressure beyond the HPT, abs, arm
32
22
Gas temperature beyond the HPT, ?C
190
327.5
Temperature at the condenser inlet, ?C
65
65
Pressure at the condenser inlet, abs. atm
2.36
2.36
Corrosion Resistance of Construction Materials
in N204
Many years of corrosion tests applied to the construction materials (including tests under stress and
thermal loading) revealed no specific forms of corrosion. The construction materials employed were
mostly highly resistant to N204: For those with chromium contents of over 10-13% (Kh18N1OT and Kh16N15-
M3B steels, etc.) a loss of 0.001-0.005 mm/year was typical at pressures of 1-150 abs. atm and tem-
peratures of 25-700?C, while at lower temperatures good results were obtained for aluminum and titanium
alloys, high-chromium cast iron, graphite, silicized graphite, Teflon, and so on. When studying the
Kh18N10T and Kh16N15M3B steels used for the cans of the fuel elements, a high resistance to corrosion
was obtained at temperatures up to 750?C and pressures of 150 abs. atm under static conditions for tests
lasting 1000-12,500 h, and under flow conditions at temperatures up to 550?C and pressures of 50 abs. atm
for 4000-5000 h [12, 15, 161.
Using a PGZh-1loop installation and an ampoule system, the effect of the n, y radiation of a nuclear
reactor and dissociating N204 on the corrosion resistance of such stainless steels as Kh18N1OT, Kh16N15 -
M3B, etc. was studied at 20-50 abs. atm and 500-550?C. Under these conditions, no difference appeared
in the corrosion rate of irradiated samples as compared with those working under simple thermal condi-
tions [11].
The surface of the construction materials exhibits a. compact protective oxide film, firmly adhering
to the surface of the steel. The preliminary formation of an oxide film by using special solutions or heating
in an oxidizing atmosphere at 300-400?C greatly reduces the corrosion rate. Prolonged tests (up to
11,000 h) at 500 and 700?C and a pressure of 50 abs. atm showed that the protective properties of the film
were not only preserved but also enhanced. This leads to a further reduction in corrosion rate [12].
Tests were carried out on samples of Kh18N1OT steel under loads of 0.9 UO.2 at 500-550?C and 50
abs. atm for 1000 h and. also on Kh18N1OT bellows in a stressed state with a stress factor of 3.25 kg/mm2.
Metallographic analyses of welded and nonwelded heat-treated samples after tests in the stressed state re-
vealed no structural changes. No intercrystallite corrosion was observed; the mechanical characteristics
remained unaltered.
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 Declassified and Approved For Release 2013/02/21 : CIA-RDP10-02196R000400040001-6
TABLE 3. Thermophysicaland Gas-Dynamical Characteristicsof a 1000 MW (El.)
Gas-Cooled Fast Reactor Working with N204
Thermal power of reactor, MW
3250
3980
Flow of coolant, kg/sec
3850
5900
Gas temperature at reactor inlet/outlet, 'C
200/451-
166 /281
Gas pressure at reactor inlet /outlet, abs. atm
i6/154
169/154
Specific thermal stress of active zone, kW/liter
727
960
Volume of active zone, liters
4220
3840
Diameter of active zone (equivalent), m
2,32
2.168
Height of active zone, m
1.0
1.04
Flattening of active zone (D/H)
2.32
2.08
Diameter of fuel-element cans, mm
6.2 x 0,3
7.0 X 0.4
Number of fuel elements in cassette
397
397
Size of cassette under the "key" and can thickness, mm
145 X 2
161 x 2
Thickness of the side breeding zone SBZ, mm
300
330
Thickness of the end breeding zone EBZ, mm
400
400
Thermal loading of fuel-element surface, kcal/m2 ? h
average
1.93.106
2.90.106"
maximum
2,53.106
3.77.106
Maximum temperature of fuel-element cans (outer/inner), ?C
549/578
341/412,
The same allowing for superheating factor, 'C
625/657
-/453
Maximum fuel temperature (nominal /allowing for superheating), ?C
1134/1314
568/635
Mean gas velocity in cassette, m/sec
25.3
30
Mean heat-transfer coefficient in the most-stressed set of fuel elements,
22730
68000
kcal/m2. h- *C
Fuel. Compositions of' Fast N204 Reactors
As regards the one-circuit arrangement for converting the heat of nuclear power stations with fast
N204 reactors into energy, stringent demands are made upon the air-tightness of the fuel elements for a
burn-up. of up to 10 at. % in U02 and 5 at. % in metallic fuel in order to preserve access to the working
circuits of the nuclear power station.
For gas-cooled fast reactors with a high specific thermal stress. of 1000-1200 kW /liter and a- high
external pressure, fuel composites of the matrix type (which have higher thermal conductivities [19, 201)
are preferable to those in the form of pure U02 [10] or uranium carbide. At the present time fuel com-
posites of U02 + 30% Cr-Ni in fuel-element cans developed in the Federal German Republic are under-
going reactor tests in Belgium, using a BR-2 reactor operating at up to 5-6 at. % burn-up, with fuel-ele-
ment diameters of 5.5-7.5 mm, fuel-element can thicknesses of 0.4 mm, a linear thermal loading of the
fuel element equal to 500-600 W/cm, a fuel-element can temperature of 600-700?C, and -a fuel temperature
of 650-1300?C. No radiation swelling of the fuel'elements has been detected [17].
Good prospects for use in fast reactors and excellent compatibility with N204 are offered by a fuel
composite made of UC + SiC in the form of microparticles 0.6-2.5 mm in size. Tests were carried out in
[20] on fuel particles 850 ? 50 p in diameter made from (U0.75, Pu0.25) 02-x using the gel method; these are
being used in producing U02-Pu02 spheres for the VIPAK reactor (a sodium-cooled fast reactor). The
particles are coated with a porous buffer layer of pyrocarbon 45-55 p thick with a compact layer ofpyro-
carbon 10 + 5 ? thick on top of it, followed by a layer of SiC 45-90 p thick. Then the particles were roasted
and the fuel core sintered to a high density, forming fuel of the "peas in a pod" type.
Tests were also carried out on two forms of particles under- irradiation at 900-950?C in a medium of
CO2. We. see from Table 1 that the problem of obtaining fuel for fast reactors is being solved positively,
and even at the present time we have fuel yielding average burn-ups of 5-8 at. %.
On the basis of experience gained in the use'.of'fast reactors in the Soviet Union and elsewhere, we
may conclude that the most suitable material for the cans of fuel elements in the next decade will be Khl 6 -
N15M3B austenitic stainless steel, which exhibits a high corrosion resistance in N204 up to 150 abs. atm
and 700?C [8]. -
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 Declassified and Approved For Release 2013/02/21 : CIA-RDP1O-02196ROO0400040001-6
TABLE 4. Physical Characteristics of a Gas-Cooled 1000 MW (El.) Fast Reactor Working with
N204
Thermal power of reactor, MW
3250
3250
3980
3980
Gas temperature at reactor inlet/outlet, ?C
200/452
200/452
160/281
160/281
Gas pressure at reactor inlet/outlet, abs. arm
169/154
169/154
169/154
169/154
Specific thermal stress, kW /liter
1167
1167
960
960
Form of fuel
Matrix
Matrix
Metal
Metal
Critical mass, kg
1555
2040
1985
2960
Enrichment Qf fuel with respect to profiling zones,
9.6/13
14.5/20.8
5.4/7.7
8.1/11.6
Average enrichment of fuel,
11.5
17.5
6.5
9.8
Maximum burn-up, at. %
10
10
5
5
Plutonium conversion ratio
1.55
1.15
1.95
1.45
Doubling time for a single recharging in closed cycle
Te.c. = 1 year
72
5
3.1
Te.c. = 0;5 year
5
3
Off-loading of excess plutonium from the reactor in the con-
618
931
944
tinuous off-loading mode
Annual consumption of natural uranium per ton of excess
plutonium, tons
Nuclear power campaign (V = 0.8), years
0.81-
0.875
0.83
0.82
Time required to find fuel in side screens, years
0.97.
1.14
2.1
2.4
Characteristics of fission-neutron spectrum:
average number of neutrons per capture n
1.90
2.59
2.04
average ratio of fission to radiative capture cross section a
0.237
0291
0.134
0.213
In open cycle.
Choice of Parameters for the Equipment of a
Single-Circuit Nuclear Power Station with a
Fast N204 Gas-Cooled Reactor
The characteristic features governing the cost of electrical power in nuclear power stations with fast
reactors (capital component 60-70%) suggest that the optimum top temperature of the cycle will be con-
siderably lower than in thermal power stations. In view of this, calculations were carried out for the gas-
cooled reactors, regenerators, condensers, preheaters, and turbines of N204 nuclear power stations on a
thermophysical, gas-dynamical, and physical basis, with a view to optimizing the parameters, due allow-
ance being made for the kinetics of the chemical reactions. The remaining components of the net cost,
such as the technological water supply, the fittings, pipelines, auxiliary systems, and the cost of erection
and structural work were taken into account by analogy with data relating to nuclear power stations of other
types.
These calculations revealed the maximum gas pressure in front of the turbine (150-170 abs. atm), the
minimum temperature heads in the heat exchangers (15-30?C), and the bottom pressure of the cycle (1.9-
2.1 abs. atm) giving the best economic indices. These calculations also showed that, owing to certain
special characteristics of the properties of N204 within the hypercritical range of parameters (p > 103 abs.
atm), there were two economically-equivalent optimum regions as regards specific computed expenditure
forthe maximum temperatures of the gas beyond the reactor: 430-480 and 250-300?C.
The thermal system of the nuclear power station provides for two parallel-connected gas turbines of
2 x 500 MW for each 1000 MW (el.) reactor. The characteristics of the thermodynamic cycle for these
versions are given in Table 2 [111.
It is very hard to analyze the emergency conditions applicable to gas-cooled fast reactors and to en-
sure the reliable emergency shut-down cooling of a fast reactor in the case of a rupture in the pipeline of
the main circuit. The system of emergency shut-down cooling for a gas-cooled fast N204 reactor con-
stitutes a closed gas-liquid circuit connected to the reactor in parallel with the loops of the main circuit,
and consists of an evaporator-regenerator, a condenser, a pump, and a liquid store (30-50 m) containing
liquid N204. In the case of emergency involving a loss of pressure in the reactor vessel, superheated
liquid is emitted from the pressure store and ejected into the active zone (even passing out into the reactor
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 Declassified and Approved For Release 2013/02/21 : CIA-RDP10-02196R000400040001-6
Fig. 3. Basic thermal
scheme of the BRIG-50: 1)
First reactor; 2) HPT; 3)
mains water heater; 4)
water pump; 5) water heat-
er; 6) LPT; 7) electrogene-
rator; 8) condenser; 9)
pumps; 10) regenerative pre-
heater; 11) regenerator.
room itself). In order to neutralize the emergency ejections of cool-
ant, a water spray is provided, with subsequent neutralization by
means of an alkaline solution.
Thermophysical and Physical
Characteristics of Gas-Cooled
N204 Fast Reactors
Thermophysical calculations of a gas-cooled fast reactor were
carried out for an active zone corresponding to the dimensions of a
steel vessel of the VVER-1000 type with working pressures of 170
abs. atm, a maximum diameter of 4.2-4.3 m, and a height of 10-12 m.
Engineering design showed that vessels of this kind might house the
active zone of a fast reactor with dimensions of 2.2-2.4 m and a diam-
eter/height ratio of 1.8-2.2, with end and side breeding zones 400-500
mm thick, and also radiation and thermal screens for protecting the
reactor vessel from fast neutrons. The favorable thermophysical
properties of N204 facilitate the efficient take-off of heat from the
active zone of the gas-cooled fast reactor.
Thermophysical and gas-dynamical calculations of gas-cooled
fast reactors were undertaken on the basis of experimental data relat-
ing to heat-transfer coefficients for the two ranges 160-280 and 200-
450?C. The calculations showed that dissociating N204 gave an effi-
cient heat take-off from the active zone of a fast reactor with a mean
thermal stress of 1000-1200 kW /liter, a maximum fuel-element can
temperature (allowing for overheating factors) of 650-680?C, a gas temperature at the reactor outlet of
450-480?C, a pressure of 140-170 abs. atm, and a gas preheating of 230-270?C (Table 3).
As fuel for a fast reactor with parameters of this order a matrix-type fuel composition was con-
sidered. The maximum temperature of the matrix fuel equalled 1300-1400?C. The effective density of
the fuel was taken as 7 g/cm3. In calculating the doubling time, it was assumed that the time of the ex-
ternal fuel cycle was either 1 or 0.5 of a year; the maximum depth of burn-up was 10 at.%. In these cal-
culations it was found that T2 ='6-7.2 years (Tee = 1 year) or T2 = 4.5-5 years (Te.c. = 0.5 year); and the
conversion ratio CR = 1.55. Analysis of the results of the calculations shows that the reduction in the
doubling time relative to the sodium reactors is a consequence of the harder neutron spectrum and the much
greater energy stress in an N204 gas reactor. The gas reactor also has more favorable characteristics
from the point of view of safety (the reactivity effect arising from the removal of coolant is 2-3 times
smaller than in a sodium reactor).
Thermophysical and physical calculations of a gas-cooled fast reactor with maximum temperatures
of 200-300?C were also carried out for a fuel composite based on metallic uranium. The fuel composite
was taken in the form of a uranium alloy. Table 4 shows the results of the physical calculation for fast
N204 reactors.
Thermophysical calculations carried out on the basis of experimental data for thermal fluxes of (3-4)
? 106 kcal/m2 ?h and a temperature range of the gas in the reactor equal to 165-281?C at pressures of 150-
170 abs. atm indicated the practical possibility of reaching specific thermal stresses of 950-1200 kW /liter
for a maximum fuel-element can temperature of no greater than 500?C and a fuel temperature (allowing
for overheating) no higher than 635?C. Such thermophysical characteristics of fast reactors (using low-
alloy metallic fuel) yield a considerably harder neutron spectrum and a CR of 1.95, with T2 = 5-years, and
an annual off-loading of excess plutonium equal to 944 kg/year. The use of such a reactor in the guise of
a refabricator yields high conversion ratios, amounting to 1.45 and a T2 of 4 years on supplying 235U (10(7
enrichment). Working in the guise of reactor-refabricator with U5 + U8, the annual development of plu-
toniumamounts to 1700 kg /year.
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 Declassified and Approved For Release 2013/02/21 : CIA-RDP10-02196R000400040001-6
TABLE 5. Technical Characteristics of the BRIG-50 Nuclear Power Station
NBRIG
Low Tome
Coolant
N204
N204
Number of circuits in the nuclear power station mode
1
1
Number of circuits in the power-and-heat and power-or-heat
3
3
modes, including the external thermal newtorks
Number of loops of the nuclear power station from safety
considerations
Number of systems of emergency shut-down cooling
2
2
Electrical power (net), MW
50 (25x2)
50 (25x2)
Thermal power of the reactor, MW
227
168
Basic parameters of the coolant in front of the turbine;
temperature, ?C
320
450-
pressure, abs. atm
150
150
Power required for internal use, MW
6
5.2
Maximum heat production in the power-or-heat mode,
147
123.5
kcal/h
Parameters of the hot water of the external network
temperature at outlet from power-or-heat and power-and-heat stations,?C
150/130
150/130
temperature at inlet into power-or-heat and power-and-heat stations, ?C
70
70
Parameters of the water in the intermediate circuit
temperature (inlet/outlet), ?C
80/160
80/160
pressure, abs. atm
15
15
Net efficiency of the installation in the nuclear-power station mode
22/28.9
29.8/32.3
(tf = 150?C/130?C),
Net efficiency of the installation in the power-or-heat mode
tf = 150?C/130?C),%
Net efficiency of the installation in the power- and-heat mode
(tr =150?C/130?C),
Flow of cooling water in the condenser in the case of a cooling agent, tons/h
reservoir
1.28.104
1.06.104
cooling tower
2.12.104
1.41
Supply of cooling water in the case of a cooling agent, m3/sec
reservoir
0.028
0.024
cooling tower
0.075
0.05
Total period of service, years
30
30
Campaign of reactor, years
0.63
0.4
Maximum possible fast neutron flux operating as nuclear power station,
3.56.101s
4.5.1015
power-and-heat and -power-or-heat station, neutrons/cm2. sec
Maximum possible fast neutron flux with the active zone working as a re-
search reactor (fuel-element diameter 3 mm), neutrons/cm2?sec
Initial charging of 235U, kg
414
200
Enrichment of nuclear fuel with 235U, ?fo
34.7
75,6
Doubling time (Te,c = 0.5 year)
4.1
4.4
Technical Proposals for an Experimental-Industrial
Nuclear Power Station with a Fast-Neutron
Gas-Cooled Reactor Having an Electrical
Power of 50 MW
In developing ideas for an experimental-industrial nuclear power station with a BRIG-50 reactor (a
fast reactor based on N204), the following aims were envisaged: the achievement of nominal fast neutron
fluxes of (6.7-7.0) -1015 neutrons/cm2 ? sec and specific thermal stresses of 900-1200 kW/liter for gas
parameters of 130-150 abs. atm and 280-450?C, as characteristic of 1000-1200 MW nuclear power stations.
For such values of the thermal stress it is possible to achieve a short doubling time for the nuclear fuel
(4-5 years). It was also naturally wished to select the minimum thermal power of the experimental reactor
in order to reduce the cost of constructing the experimental-industrial nuclear power station. Further-
more it was extremely important to ensure the possibility of testing the fuel composites, fuel elements,
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 Declassified and Approved For Release 2013/02/21 : CIA-RDP10-02196R000400040001-6
TABLE 6. Characteristics of the Reactors of the Experimental-Industrial Installation BRIG-50
NBRIG-50 ( T
ow
VBRIG-50
test
producer
(research)
Thermal power of reactor, MW
227
227
168
242
Flow of coolant, kg/sec
295
295
246
238
Gas temperature at reactor inlet/outlet, ?C
171/321
171/321
208/451
175/451
Gas pressure at reactor inlet/outlet,abs. atm
169/154
169/154
169/154
169/154
Average energy stress of active zone, kW7liter
.
843.
1880
1300
2560
Equivalent diameter of active zone, m
0
654
0
528
0
586
0
492
Height of active zone, m
,
0,763
,
0,524
,
0,455
,
0,472
FlatteningQ
of active zone (D/N)
0;857
1,01
1,29
1,04
_
Volume of active zone, liters
256
114
7
122
8
89
8
Thickness of end screen, m
0
6
,
2 st.
6+0
0
,
6
0
,
0
4
Thickness of side screen, m
,
0
6
,
,
6+0,2 st.
0
,
0
6
,
0
4
Size of fuel-element array under the key, mm
,
39
,
39,5
.
39
,
39,5
Step of fuel-element array in a triangular lattice, mm
40
40,5
40
40
5
Number of cassettes in the active zone
242
154
195
,
134
Thickness of cassette can, mm
1
0
1
0
1
0
1
0
Number of fuel elements in a cassette
,
37
,
91
,
37
,
91
Diameter of fuel element, mm
5,0
3.0
5,0
3,0
Form of fuel
Metal
Metal
Matrix
Matrix
Effective density of fuel composite inthe core, g/cros
14,1
14,1
7
7
Total thickness of fuel-element can, mm
3
0
0,16
3
0
-0
16
"Step of fuel element (triangular lattice), mm
,
5,8
3,8
,
5,8
,
3,8
Maximum temperature of fuel-element can (external/internal), ?C
410/437
442/467
570/629
591/624
Maximum temperature of inner can,allowingfor superheating factors, ?C
Maximum temperature at the center ofthefueltablet, allowingfor
superheating factors,
Enrichment ol" fuel with the fissile isotope 235U
28,25
46,8
73
90
Maximum fuel burn-up, at. %
5
5
10
10
Reactor campaign,years
0,6
0,21
0,38
0,20
Coefficient of nonuniform ity of heat evolution Kr/Kz
1,37/1,2
1,2/1
1
37/1
1
2
1
15/1
1
Volumetric proportions of materials in the active zone,
fuel
25
65
,
46
21
,
,
27
8
,
,
coolant
,
33
16
,
40
48
,
39
5
metal structures
,
27
01
,
46
25
,
32
7
others
,
68
14
,
13
6
,
Volumetric proportions of materials in the side screen, % .
fuel 22eU
,
56
61
,
56
61
steel
,
-19
65
80
,
19
65
80
coolant
,
21
08
20
,
21
08
20
Charging of fuel per campaign, kg
,
316
215
,
219
177
Maximum flux, neutrons/cm2 ' sec
7,74.1015
7.1015
11
9.1015
9
14.1015
Maximum neutron flux integrated with respecttotime, neutrons/cm2
11,2.1022
,
6,26.1022
,
11,8.1022
8,8.1022
Maximum neutron flux with E ? 0,4 MeV, neutrons/cm2 ? sec .
3,56.1015
6,16. 1015
4,5.1015
6,3.101s
Maximum neutron flux with E >_ 0.4 MeV, integrated with respect to time,
1.1022
4
4;1022
5
4.1022
neutrons/cm2
Total conversion ratio
,
,
1,28
and construction materials developed in nuclear power stations of the BRGD-1000 type -under conditions in-
volving the use of an experimental fast-neutron reactor with a neutron flux characteristic of high-power
nuclear power stations.
These aims may be achieved in a nuclear reactor with a thermal power of 150-200 MW using the pro-
posed parameters of the coolant and high specific thermal stresses. The nominal neutron flux characteris-
tic of the BRGD-1000, (6-7) ? 1015 neutrons/cm2 ?sec, may be achieved in the experimental BRIG-50 reactor
by increasing the specific thermal stress to 1800-2500 kW/liter on converting it to the research mode of
operation (Tables 5 and 6).
The experimental-industrial nuclear power station with the BRIG-50 reactor is intended for solving
the principal technological and constructional problems involved in making a fast-neutron reactor of the
gas-cooled type using the dissociating coolant N204. It is intended to carry out the following work with the
experimental-industrial nuclear power station:
1) Mass resource tests on fuel elements under conditions corresponding to the working parameters
(with respect to fast-neutron flux and thermal and energy stress) of BRGD-1000 industrial reactors; the
experimental-industrial development of a fast gas-cooled BRIG-type reactor as the prototype of a reactor
/secondary-nuclear-fuel producer as recommended for power-or-heat and power-and-heat stations. For
these purposes an electrical power of the atomic power station of around 50 MW is assumed (Fig. 3);
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 Declassified and Approved For Release 2013/02/21 : CIA-RDP10-02196R000400040001-6
2) the experimental-industrial development of an efficient closed gas-liquid cycle, equipment for the
technological circuit, and a complex of technological problems associated with the use of N204 as a coolant
and working substance for high-powered nuclear power stations;
3) the development of standards for the main equipment of nuclear power stations and the carrying
out of the necessary material-science tests in order to verify the validity of the choice of construction
materials.
The design of a fast gas reactor should ensure the accommodation of two active zones in the reactor
vessel, corresponding to low-temperature (NBRIG) and high-temperature (VBRIG) versions of the nuclear
power station. The technological scheme, the equipment, and the composition of the nuclear power station
should have a single technological embodiment, allowing the practical use of the experimental-industrial
nuclear power station with NBRIG or VBRIG reactors, and also in the modes of combined and separate
production of heat and electrical power, i.e;, the nuclear-power-station, "ATETs, " and "AK" modes.
Nuclear and Chemical Safety of the BRIG-50 Installation
In view of the absence of experience in the use of nuclear power installations involving fast reactors
cooled by means of a dissociating coolant, a particularly meticulous approach must be made to the solution
of these problems.
Emergency situations to be foreseen, as far as possible prevented, and ameliorated as regards their
harmful consequences, include the unsealing or rupture of the circuit at any point in the technological sys-
tem, deenergizing of the whole nuclear-power installation, and also processes in the reactor or reactor
control and monitoring systems leading to a dangerous change in reactivity or radioactive contamination of
the circuit.
Under certain emergencies such as the stopping of the pumps as a result of power failure or jamming,
the use of N204 as coolant in the gas-liquid cycle may play a decisive part in preventing the melting of the
active zone. In this case there are two favorable circumstances: the large amount of coolant in the circuit
as compared with its rate of flow, and the considerable pressure differences in the circuit, which may
provide for an initial natural shut-down cooling of the active zone within 20-60 sec.
In order to ensure safe use of the BRIG-50 nuclear-power installation, the following measures of
radiation and chemical safety are proposed.
1. To use a two- or three-loop system of reactor cooling and two independent loops of emergency
shut-down cooling, based on different principles. Simple operation of the installation with a single loop
is not to be allowed.
2. To eliminate the possibility of rupture in the reactor vessel and the cooling pipes in front of the
cut-off valves by suitable constructional measures (by taking large reserves of strength in the design cal-
culations and ensuring quality control of the manufacture and erection). In the case of rupture of the ves-
sel, to ensure cooling of the active zone with liquid coolant.
3. To provide reserves for the pumps of the main circuit and the coolant circulating system, and to
provide inertial masses on the rotors for emergency running-down purposes.
4. To provide the BRIG-50 nuclear-power installation with autonomous emergency electrical supply
systems.
5. To enclose the whole equipment of the installation (including any parts which might lead to radia-
tion or chemical hazard in the event of damage or partial failure) in a hermetic concrete sheath; personnel
and equipment should only pass into this region through special locks.
1. A. P. Aleksandrov, Izvestiya, December 25 (1971).
2. G. B. Levental' et al., At. Energ., 32, No. 3, 187 (1972).
3. A. M. Petros+yants et al., At. Energ., 31, No.. 4, 315 (1971).
4. V. V. Orlov, ibid., 295.
5. V. V. Orlov, M. F. Troyanov, and V. B. Lytkin, At. Energ., 30, No. 2, 170 (1971).
6. A. K. Krasin et al., Fourth Geneva Conf., R-431 (1971).
7. V. B. Nesterenko, Physico-Technical Bases of the Use of Dissociating Gases as Coolants and Work-
ing Substances in Nuclear Power Stations [in Russian], Nauka i Tekhnik, Minsk (1971).
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8. A. K. Krasin and V. B. Nesterenko, Thermodynamic and Transfer Properties of Chemically React-
ing Gas Systems in Russian], Nauka i Tekhnika, Minsk Pt. I (1967); Pt. II (1971).
9. A. K. Krasin et at., Heat Transfer in Chemically Reacting Gas Coolants [in Russian], Nauka i Tekh-
nika, Minsk (1971).
10. A. P. Leipunskii et al., see [3], p. 383.
11. "Dissociating gases as coolants and working substances of power installations, " Trans. of the Third
All-Union Conference on Dissociating Gases, Nauka i Tekhnika, Minsk (1972).
12. "Dissociating gases as coolants and working substances of power installations, if Trans. of the Second
All-Union Conference on.Dissociating Gases, Nauka i Tekhnika, Minsk (1971).
13. L. I. Kolykhan, V. F Pulyaev, and V. T. Derov, Izv. Akad. Nauk Belorussian SSR, Ser. Fiz.-
Energ., Nauk, No. 2 (1971).
14. B. S. Petukhov, A. S. Komendantov, and S. A. Kovalev, Teplofiz. Vys. Temp., No. 6 (1969).
15. A. M. Sukhotin and L. N. Lantratova, Izv. Akad. Nauk Beloruss. SSR, Ser. Fiz.-Energ. Nauk,
No. 2 (1968).
16. A. M. Sukhotin and L. N. Lantratova, Handbook of Chemistry, Corrosion and Protection of Chemi-
cal Apparatus, Vol. 3 [in Russian], Khimiya, Leningrad (1970).
17. M. A. Bazhin et al., Optimization of the Parameters of Power Installations Using Dissociating Work-
ing Substances [in Russian], Nauka i Tekhnika, Minsk (1970).
18. J. Yellowflies et al., Atomnaya Tekhnika za Rubezhom, No. 3, 14 (1973).
19. H. Bumm et al., Proc. Inernat. Meeting on Fast Reactor Fuel and Fuel Elements, Karlsruhe, Sep-
tember (1970).
20. I. Sayer, The U-. K. Support of the Coated-Particle-Fuelled Gas-Cooled Fast Reactors,. Minsk (1972).
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E. P. Anan'ev and G. N. Kruzhilin UDC 614.876
The problem of radioactive safety in the operation of nuclear power stations is quite multilateral and
widespread. Nevertheless, it may be mentioned, bearing in mind the previous period, that its tenor has
been well-known to specialist physicists from the very beginning of the construction of nuclear installations.
Undoubtedly, it may be assumed also that at that time, just as now, it was important to preserve the active
zone from burn-out of the fuel elements in all possible situations at nuclear power stations. This problem
is imposed, first of all, on the automatic emergency protection system and signalization of the reactor,
and secondly on the active zone cooling system during normal operation of the reactor and during shutdown,
and also in the various possible emergency states of a nuclear power station.
However, in water-cooled reactors, there is a quite substantial radioactivity even under "normal"
operating conditions. It comprises the "oxygen" radioactivity of the water, the radioactivity of the circula-
tion loop corrosion products present in the water and the radioactivity of fission fragments which enter the
coolant flow from individual leaky fuel elements. The first two components are inherent in the very pro-
cess taking place in the reactor, i.e., they are technological. The latter component, i.e., the radioactivity
of the water, due to leakage from the fuel elements of fission fragments, is essentially of a probability
nature. Despite the great improvements achieved in the design development of fuel elements for power
reactors and in the technology for their mass production, nevertheless, during operation of nuclear power
stations some of the fuel elements become leaky, and the consequence of this is that highly radioactive
fission fragments start to discharge into the coolant flow. In general, this is not difficult to understand,
because on the one hand the fuel elements are quite "delicate" in their construction, and correspondingly
complex in relation to ensuring their reliability under operating conditions and, on the other hand, their
total number in power reactors is enormous. Thus, in the Novovoronezh nuclear power station there are
about 31 ? 103 fuel elements in the VVER-210 reactor, about 43 ?103 in the VVER-365 reactor and about 43.9
? 103 in the VVER-440 reactor.
It is well-known that in the core of a uranium metal fuel element, the water which is passing through
its leak-tight cladding interacts chemically with the uranium. As a result of this, uranium hydride is
formed in particular with a friable structure and with a considerably greater specific volume than uranium,
which leads to total destruction of the cladding and leaching-out of the core material. In consequence of
this, the radioactivity of the circulating water becomes extremely high even in the case of damage to only
a single fuel element. As a result of this, the inside surfaces of the circulatory loop also are strongly
contaminated and become highly radioactive; these contaminants are leached-out only poorly and maintain
a high radioactivity over a long period. Operation of power reactors under these conditions would be margi-
nally difficult. On account of this, the cores of power reactor fuel elements are made from UO2 and not
uranium metal, although in this case the fuel requirement is increased somewhat.
Uranium dioxide does not react with water and it is not leached-out. Because of this, a leaky fuel
element - if it appears under operating conditions - does not rupture in the flow of water. In this case,
mainly the fission products xenon and krypton pass through the crack of a leaky fuel element. They diffuse
preliminarily from the body of the fuel to the surface of the core, which amounts to 0.1 of their total amount.
Then they diffuse through the gas space to the crack which has formed; the quantity here can be estimated
at 0.01. Thus, the release of gaseous fission products from the fuel element can be estimated at 10-3 of
the total amount of fission products. As a result of this, it can be assumed that the radioactivity of these
Translated from Atomnaya Energiya, Vol. 37, No. 1, pp. 22-27, July, 1974. Original article sub-
mitted September 7, 1973.
?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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fission products is less, approximately by a factor 10-5, than the radioactivity which would enter the flow
in the case of rupture of a single fuel element of uranium metal. Thus, the practical value of converting
to the use of UO2 in reactors with water cooling is extremely significant. Research work resulting from
this conversion was carried out at one time in the I. V. Kurchatov Institute of Atomic Energy by V. V.
Goncharov and other coworkers, under the direction of Academician A. P. Aleksandrov.
The widespread practice of operating power reactors with UO2 fuel elements has shown that their
operating conditions are completely acceptable in relation to the radioactivity of the environment. At the
same time, it is obvious that certain improvements in this respect must be carried out. Obviously, the
most important of these improvements must be special technological measures for increasing the reliability
of the fuel elements. As shown in practice, decontamination of the flow of circulating water in ion-exchange
filters is also of great importance; here, the radioactive particles are trapped in ionic form and thus the
amount of corrosion products in the circuit is maintained at a low level, which leads to a significant effect
in relation to the radioactivity of both the cooling water and of the surfaces of the cooling circuit. As
applicable to a power generating unit with a boiling reactor, methods of handling the radioactive gases
from the turbine condenser and the conditions for operation and maintenance of the turbine also are im-
portant. There are problems associated with this, which are narrowly specialized in their tenor; they
are being solved in the proper manner with the assurance of the existing standards of radioactivity both at
nuclear power station sites and beyond their boundaries.
We note additionally that fuel elements of UO2 have a special future: if a fuel element becomes leaky
.in the water flow, its shape remains unchanged and therefore it is not amenable to visual detection. Defini-
tion of a cassette with leaky fuel elements is quite difficult in this case. In case of necessity, this is un-
dertaken after removing the cassette from the reactor; according to the presence in the fuel element of
specific radioactive isotopes [1]. This operation is complex and requires a long time to carry it out. Be-
cause of this, and also owing to the generally insignificant escape of fission products (mainly gaseous)
through leakage, as mentioned, a search for the damaged fuel elements as a rule is not undertaken in the
period between fuel rechargings into the reactor. According to current estimates, it is considered 'per-
missible" if the number of leaky fuel elements does not exceed 1% of the total number in the active zone.
However, according to operating data, the relative number of damaged fuel elements is considerably less.
Undoubtedly the damage probability will be reduced further by improvements in design and production
technology. Therefore, the special features of UO2 fuel elements mentioned are recalled here only as a
statement of fact and not for the purpose of any justification or protection whatsoever of the "admissibility"
.of damage to fuel element claddings occurring during operation of the reactors.
Thus, in the light of performance and accumulated experience, the first and principal barrier against
radioactivity in a nuclear plant has always been assumed to be the fuel element claddings which, in their
turn, are protected from dangerous burn-out by a system of automatic equipment and by the appropriate
systems for cooling the active zone in normal, shutdown and emergency situations.
Among the possible emergencies in nuclear power stations, the case of coolant leakage must be
counted as particularly valid. Such cases are inevitable during prolonged operation. In organic-fuelled
steam powered plants, very small leakages through looseness of the individual parts of the shut-off equip-
mentor its packing gland present no hazard and therefore they do not cause any special worry. But with
the appearance of a leak through breaks in the flange gaskets or through direct damage to one or another
small-diameter tube, a more or less urgent shutdown is necessary for repair, in order to avoid the pos-
sibility of this type of damage developing with increased leakage, leading to more dangerous consequences.
However, in these cases shutdowns do not usually take place immediately, but they are effected during cer-
tain hours in order to take into account the operating requirements of the users. Thus, on power. stations
feeding into the.grid system, shutdowns for maintenance in such cases are coordinated provisionally with a
grid system dispatcher and account is taken of the grid load, for the possibility of shifting beyond the peak
loading period. It may be mentioned for correctness that, during the operation of modern power generating
units with a capacity of 300 MW, such delays in a shutdown occur with a leakage through a damaged tube of
up to 20 t/h.
It is obvious that in nuclear power stations there is the same attitude toward similar leakages. For
the closed circulatory loop of a water-cooled/water-moderated reactor, only leakages are classified as
"small" which do not lead to a reduction of pressure in the circuit; in the contrary case, the emergency
protection system operates, with shutdown of the reactor. However, in this case of reactor shutdown,
safety is not completely guaranteed because when the pressure in the circuit is reduced to the saturation
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pressure corresponding to the temperature of the circulating water, steam generation will take, place in the
reactor vessel and, as a result of this, the danger will arise of steaming of the active zones with heating up
or even burn-out of the fuel elements. Therefore, when the pressure in the circuit is reduced to a speci-
fied limit, an emergency reactor cooling system is switched on automatically. For this, the required quan-
tity of water containing boric acid (in order to exclude nuclear reactions in the active zone [2]) is supplied
to the reactor vessel from special reservoirs by means of emergency electric pumps.
With the appearance of hot water leakage from the loop, steaming occurs and, at the same time,
radioactive contamination of the corresponding location occurs. In order to localize this hazard, the
places through which the conduits carrying the hot radioactive water pass are made hermetically tight.
Moreover, they are provided with sprinkler devices for spraying cold water for the purpose of condens-
ing any steam which is released by the leak.
The emergency cooling and hazard localization systems mentioned are designed to operate on the
appearance of a significant leakage of hot water from the reactor circuit, including damage or fracture of
any small-diameter pipe and, in our opinion can be regarded as the second radioactive safety barrier.
It should be emphasized that the efficiency or, more precisely, the effectiveness of this barrier is
determined, as in the case of the first barrier mentioned above, also by the reliability of the electric
power section, including the power supply to the electric motors of the emergency pumps. In order to
avoid interruption of the power supply to a number of important consumers, an emergency electric power
supply is provided for reactor shutdown in these circumstances, in the form of a large battery of accumu-
lators, calculated to operate for approximately one hour, and Diesel generators designed for continuous
operation over several hours or even several days. It is obvious that the importance of this plant is enor-
mous, just like the electrically powered safety control, as a whole, in the radioactive system in nuclear
power stations. The main specific difficultly consists in that the emergency equipment mentioned is
located in a deep emergency system. At the same time, in accordance with its purpose, it must always
be in complete readiness. Therefore, a systematic check of the readiness of this equipment, in particular,
is exceptionally important under normal operating conditions of the nuclear power station.
The concept discussed is completely logical and conclusive. It can be assumed that it has found wide
recognition among specialists and that it is used in practice. Nevertheless, in designing nuclear power
stations with pressure-vessel reactors in the USA, one further safety barrier has received widespread
recognition - protective envelopes (the so-called containment vessels), intended to retain the steam re-
leased from the entire bulk of hot water contained in the hot water loop. As reported in primary sources,
for this, one must bear in mind the case of total transverse rupture of the main duct of the circulatory loop,
with the discharge of vaporizing water from both ends of it.
The majority of experimental reactors, and also plutonium production reactors have been constructed
without containment vessels. Gas-cooled power reactors without containment vessels have been constructed
in England and France. Containment vessels appeared for the first time at the Shippingport (USA) reactor,
where three were installed - one above the reactor and two correspondingly above each of the two steam
generators. In connection with this, one of the Westinghouse technical directors, I. Simpson, said at the
time [3]: "It is obvious that containment vessels are expensive and will be phased out with time. We con-
sider it to be extremely improbable that these devices would ever function. But in order to double the
safety in this first commercial nuclear power station, we consider it necessary to construct these expen-
sive facilities."
Since that time, American firms have continued to construct and improve containment vessels. For
example, in the latest design of a containment vessel built by the firm of Westinghouse [4], ice condensers
are being installed (around the perimeter), intended for rapid condensation of steam escaping as a result of
bursting of the primary loop duct. The total weight of ice amounts to 1000 t. It is contained in the form of
fine pieces in wire baskets at a temperature of -10?C approximately, which is achieved by means of con-
tinuous blowing with cooled air and also by suitable thermal insulation structures. At the Sequoia (USA)
nuclear power station, the ice condenser for the 1125 MW unit with a water-cooled/water-moderated reac-
tor consists of 24 moduli, each having cross-sectional dimensions of 4 x 4 m approximately and a height
of 24 m. Thus, I. Simpson's forecast relating to the rejection of containment vessels has not come true in
American practice.
Let us consider now the question of containment intrinsically. As already mentioned, the advisability
of this problem is motivated by the assumption of total rupture of the primary duct of the circulatory loop.
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Because of this, we recall that in the Novovoronezh nuclear power station the primary circulatory duct of
the VVER-440 reactor has an external diameter of 560 mm and a wall thickness of 32 mm. For the VVER-
1000 project, the primary pipeline will operate at a pressure in the first loop of 150 atm, its internal diam-
eter will be 850 mm and the wall thickness 55 mm. The circulatory pipeline in the Sequoia nuclear power
station operates at a pressure of 175 atm and its diameter is 740 mm. Thus, it is obvious that the primary
ducts of power-generating reactors are related, in their dimensions, to a class of pressure vessels. The
reactor vessel is also related to this class, rupture of which is excluded according to present day concepts.
In view of this, on the basis of the existing formal logic, the obvious scientific intension would appear
to be that rupture of these ducts should also be assumed to be excluded.
But, of course, experience here is decisive. According to verbal information from the Deputy Head
of the Division of Boiler Inspection (Kotlonadzor), L. B. Segalov, during at least 13 years in steam boilers
there has been no first category emergency (according to after effects). Accidents from rupturing of drums
have occurred earlier, on fire-tube horizontal boilers and also on vertical boilers of the Shukhov type.
The cause of these accidents were oversights of the water level, the formation of a layer of scale and caus-
tic embrittlement in riveted seams. As a consequence, production of these boilers was discontinued in the
USSR and they were replaced by DKV low-capacity tubular steam boilers of the Biisk boiler factory, which
are safer in these respects.
The degree of reliability of operation in industry of tubular steam boilers can be defined in that of
50,000 twin-drum tubular DKV boilers, manufactured by the Biisk boiler factory since 1950, not a single
boiler has had an accident of first category nor has a single drum ruptured. The drums of these boilers
are rolled from a sheet of boiler steel, they have longitudinal and transverse welded seams and welded
spherical bottoms. The drum diameter is 1000 mm and the wall thickness is 13 mm. The boiler tubes,
with a diameter of 51 mm, are secured in openings of the drum with expansion, so that the calculated
attenuation factor (weakening factor) from the openings is 0.5. The boilers operate up to a pressure of
13 atm. The overall length of the 50,000 DKV boilers amounts to about 400,000 linear meters.
It is more convenient, according to wall thickness, to compare as containers the circulating ducts
of nuclear power stations with the drums of high-pressure steam boilers. In the USSR, these drums are
manufactured from carbon steel 22K at a pressure of 115 atm, with internal diameter 1300-1600 mm and
wall thickness 70-90 mm. The drums of boilers at a pressure of-155 atm with internal diameter 1300-1800
mm- and wall thickness 90-115 mm are made of 16 GNM steel. A large number of these drums have been
operating in our electric generating stations during approximately 20 years without accidents caused by rup-
ture of the welded seams or of the walls. Nevertheless, with drums of 166 GNM steel, considerable diffi-
culties were experienced at first (this steel is_a new one, developed by the Central Scientific Institute of
Technical Information for Machinery Manufacture (TsNI Hem . Longitudinal surface cracks appeared
on the inside wall of drums made from this steel, sometimes several mm in depth. In order to prevent
possible complications or accidents, these cracks which were revealed by inspection during installation of
the boiler, were removed by grinding with an abrasive wheel down to intact metal and without subsequent
filling. Further, in order to increase reliability, the wall thickness of these drums was increased by 13
mm. Moreover, this steel was produced for them by only one factory.
Accidents in steam boilers in industrial plants and in the Minenerge system of the USSR, with rare
exceptions belong to the second category according to consequences. Mostly they are expressed in over-
heating of the DKV boiler tubes because of oversight of the water level through noninspection by the operat-
ing personnel. There have also been accidents from damage to the welded seams in steam tubes. The
appearance of a through transverse crack in the vicinity of the seam is typical of this, as a consequence of
unsatisfactory heat treatment of the regions of the conduit adjoining a welded seam under assembly condi-
tions. These cracks are revealed by leakage and they sometimes reach 1 /3 of the circumference in size.
In addition, numerous cases are known of damage of a different nature to small diameter pipes under
operating conditions of steam power plants. After a long period of service, the thin-walled pulse tubes
which lead to the flowmeters and manometers, corrode. The elbows of the bypass and drainage pipes un-
dergo internal erosion due to the movement in them of the steam-water mixture at a high flow rate. Fatigue
cracks appear at the transverse seams of fine pipes because of normally imperceptible vibration. Damage
to these pipes occurs as a consequence of unsatisfactory compensation of their thermal expansions.
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Under operating conditions, damage is encountered which arises from factory defects: for example,
the use of fine tubes and accessories with imperceptible internal defects; imperceptible "undercuttings"
during working of the connecting pipes of accessories, which become concentrators of mechanical stresses;
low-grade bends of tubes in the cold state, which may lead to considerable thinning of the wall and, at the
same time, to a decrease of its elasticity; this is especially dangerous in the presence of bending stresses.
It is well-known, in particular, that as a result of the latter cause, cases have occurred in our "old" high-
pressure steam generators of fractures at the bends of unheated tubes with external diameter of 133 mm
and with a wall thickness of 10 mm, made from boiler steel 20.
For even larger ducts (such as the main circulation ducts of power generating nuclear reactors with
water cooling), the most probable cause of their damage with the appearance of cracks are only thermal
stresses, associated with their thermal compensation. From this point of view, there is a definite value
for the freedom of box-grouping of the pipes and which is characteristic for our nuclear power stations. In
the case of construction of a containment, constrained grouping of pipelines is inevitable, in consequence
of which their operating reliability, at least according to this assessment, is correspondingly reduced. At
the same time, constrained grouping of plant under containment and especially under structures which are
dependent on the conditions of their leak-tightness, considerably complicates the operation of nuclear
power stations. This concerns monitoring the state of the plant, which is carried out periodically in some
or other established order, including work due to maladjustments, and also during startup of the plant after
temporary shutdown or after repair, when personnel inspection by specialists is necessary. This relates
also to various maintenance operations. In carrying out the latter on radioactive plant, the most impor-
tant factor is the total expenditure of time, as the radiation dosage for maintenance personnel is approxi-
mately proportional to this time. It is well-known from experience, that in order to shorten the time of
maintenance, suitability of procedure and access to the plant are of first degree importance and, especially,
freedom for carrying out work on the disassembly and removal of one or other damaged unit, and also the
delivery and installation of a new unit. Because of this, it is relevant to assume that with respect to main-
tenance conditions for radioactive plant at nuclear power stations, it is essential to have convenient and
maximum possible freedom of grouping of the facilities.
Thus, the installation of a containment, together with an increase of radioactive safety of nuclear
power stations, involves a number of negative instants which cannot possibly be taken into account. Costs,
even on the design of the containment and its system, on the whole are considerable in absolute magnitude.
Therefore, the advantage of installing a protective envelope is not only not obvious but, to a certain extent,
it is even doubtful.
The practice of constructing protective envelopes in the USA is associated, first of all, as already
mentioned, with the history of the problem or, more precisely, with the start of construction of nuclear
power stations, when "double safety" also was important for social opinion in the USA. In connection with
this, reference should be made again to Simpson, who from the very beginning drew attention to the fact
that in essence protective envelopes for nuclear power stations are not necessary. But, up to the present
time, American specialists have emphasized invariably that protective envelopes in nuclear power stations
are to be constructed in case of: "the improbable fracture of the circulatory duct of the reactor" [5], or in
case of "hypothetical fracture of the reactor duct" [6]. Obviously, it may be supposed that the continued
construction of protective envelopes is also associated with the caution which naturally exists with the doc-
trine of safety on the one hand, and with the technology of heavy machine construction on the other hand.
In pressurized installations (especially thermal), relative to conditions of strength, an excessive in-
crease of pressure is the most dangerous. According to statistical data of Kotlonadzor, accidents from
this cause with fracture of one or other unit of the plant are still occurring. The majority of cases are due
to the fact that with controlled hydrostatic molding, wedging of safety valves is practised and after molding
in haste, it is forgotten to unwedge them.
Of course, such cases are possible only on subsidiary industrial plants with a correspondingly low
level of operation. As far as we know, such cases have never occurred in domestic thermal generating
stations. It is obvious, that nuclear power stations in relation to operating level, should be at least no less
than the State Regional Electric Power Stations. Nevertheless, as applied to nuclear power stations, this
situation is worthy of specific attention. It should be noted that the dangers associated with it can be elimi-
nated by the installation in the reactor circuit of two groups of valves - dump valves and emergency valves
- designed so that the latter operate at a somewhat higher pressure than the former. In this case, the
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limiting pressure for the dump valves is set at the normal level. In case of nonfunctioning for any reason
whatsoever in an emergency situation, the emergency valves will operate at a somewhat higher pressure.
We recall that in the circulatory loop of the reactor at the Sequoia nuclear power station, both these
groups of valves are present.
In relation to mechanical strength, the most important and decisive factors are the choice of materials,
the technology of manufacture of the plant components, and also the methods for their assessment. As ap-
plicable to nuclear reactors, the latter have been extended recently with the introduction of a computa-
tion on the limiting state, developed by N. P. Mel'nikov [7], which will provide an additional guarantee of
the operating reliability of the reactor circulatory loop.
Only a part of the problem being considered here has been touched on. Actually, the urgency remains
for further improvement to some or other degree of all systems as a whole, which will ensure in our nu-
clear power stations in emergency situations, the necessary cooling of the active zone and also will prevent
the escape of radioactivity beyond the bounds of the nuclear power station itself. In this respect, it can be
assumed that the system of box-grouping of reactor plant has been successfully and completely accepted in
the USSR. A logical extension of this may be an additional system for steam condensation in emergency
conditions by means of a water bubbler, located in an annex. It is important, that in the counterpoise to
the protective envelope of this type of system, it does not affect at all the grouping conditions of the main
ducts. Nevertheless, by choosing the dimensions of the water bubbler and taking account oftlie-`possibility
of rupture of the main duct, the "double safety" which has been mentioned is ensured with this type of sys-
tem.
Thus, the advisability of double safety is motivated by universal practice in the design of nuclear
power stations and is intended for their widespread construction. It should be noted that any discussions
whatsoever in this connection have almost no significance at the present time. But in spite of our confi-
dence in the fact, that with the present-day level of technology, the possibility of rupture of the main duct
is improbable, they, undoubtedly, remain valid.
LITERATURE CITED
1. F. Ya. Ovchinnikov et al, Operation of the Reactor Facilities of the Novovoronezh Nuclear Power
Station [in Russian], Atomizdat, Moscow (1972), p. 102.
2. F. Ya. Ovchinnikov et al, Operation of the Reactor Facilities of the Novovoronezh Nuclear Power
Station [in Russian], Atomizdat, Moscow (1972), p. 123.
3. I. Simpson, The Pressurized Water Reactor Forum, Mellon Institute, Pittsburg, December 2,
4.
5.
6.
1955, p. 29.
A. Tredale and N. Grimm, Nucl. Eng. Internat., 16,
185,
47 (1971).
S. Weems, Nucl. Eng. Internat., 15, 164, 47 (1970).
A. Tredale and N. Grimm, Nucl. Eng. Internat., 16, 185,
864 (1971).
7. N. P. Mel'nikov, Structural Shapes and Methods of Calculating Nuclear Reactors, [in Russian],
Atomizdat, Moscow (1973).
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 Declassified and Approved For Release 2013/02/21 : CIA-RDP10-02196R000400040001-6
LOW-TEMPERATURE SPECIFIC HEAT AND
THERMODYNAMIC FUNCTIONS OF
URANIUM BERYLLIDE
O. P. Samorukov, V. N. Kostryukov, UDC 669.2/85:669.822:536.63.7(083)
F. A. Kostylev, and V. A. Tumbakov
Of the fundamental thermodynamic constants for uranium beryllide UBe13 only the heat of formation
[1] and the entropy at standard temperature, found by the calculation method of [2], are known.
In the present paper we present experimental data on the specific heat of UBe13 in the 13.5-30314?K
range and the values, calculated based on them, of the thermodynamic functions in the condensed state in
the 0-300?K range.
Preparation and Identification of UBe13
To prepare UBe13 we used uranium in the form of fragments cleaned of the surface oxides by elec-
trolytic polishing in H3PO41 and electrolytic-beryllium powder. The impurity content in uranium and beryl-
lium and their assumed phase composition are represented in Table 1.
A mixture of beryllium (33.09 wt. %) and uranium (66.91 wt. %) in a crucible of BeO was heated in a
quartz ampoule in pure hydrogen (600 mm Hg) at 1300 f 50?C for 1.5 h. Then, the hydrogen was pumped
out at 600?C. The cooled product was ground in purified argon.
An x-ray investigation showed that the sample consists of a single phase - UBe13. A metallographic
investigation indicated that no visible impurities were present. The UBe13 sample was analyzed for uran-
ium and beryllium content by chemical methods [3, 4], and for oxygen content by the neutron-activation
method and vacuum melting in a graphite crucible with a platinum bath. We found 66.82% uranium, 32.42%
beryllium, and 0.35% oxygen. The errors in the determination of uranium, beryllium-, and oxygen equal
?0.2, 0.4, and 20%, respectively, where for oxygen this
TABLE 1. Content of Impurities in Uran- error refers to both methods of: analysis. The total con-
ium and Beryllium and Their Assumed Phase tent of the component in the sample (taking into account
Composition the impurities introduced with the original uranium and
content, wt. % beryllium) is 99.71%.
Impurity in in form Iin in form In the calculation of the phase composition of the
uranium of phase Iberylliuml of phase sample (on the basis of the composition of the mixture,
Carbon
6.10-2
UC
11.10-2
Be2C
Oxygen
6.10-3
U02
22.10-2
BeO
Nitrogen
5.10-5
UN
6.10-3
Be3N2
Silicon
3,6.10-2
U3Si
9.10-3
Si
Iron
1,4.10-2
We
1,4.10-2
Fe
Manganese
3.10-3
U01n
1.10-3
Mn
6.10-4
UsNi
-
Magnesium
Elim from the direct beam. In order to reflect the ultracold neutrons with E Elim a copper foil
(13) 50 u thick is placed between the copper part of the neutron guide (12) and the tube (14).
For the diffuse reflection of neutrons from the walls of the neutron guide in transmission there is no
major difference between the smooth bending of the tube indicated in [3, 4] and the direct rotation employed
here.
The whole system is evacuated through a vertical internally polished tube (15) 135,mm in diameter,
using a diffusion pump. In order to prevent oil vapor from falling on the copper surfaces and the conver-
ter, nitrogen and semiconducting traps are provided. ,
The working vacuum in the system is 5.10-4 mm Hg. The total length of the copper neutron guide is
a little greater than 6 m.
The ultracold neutrons from the converter (4) are transported along the neutron guide to the detector
system (18), within the jacket of which lie two FEU-13 photomultipliers with ZnS scintillators covered with
a thin layer of a lithium compound (-'0.05 mg/cm2 Li) enriched with the 6Li isotope [3]. The detectors are
in turn covered with a thin (10 p) copper slide. The pulses from the detectors'are fed through amplifiers
and discriminators to scalers or to a multichannel AI-100 pulse analyzer.
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 Declassified and Approved For Release 2013/02/21 : CIA-RDP10-02196R000400040001-6
neutrons/cm2 ?sec
the filling system is passed into the neutron guide.
Merasurements and Discussion
In order to calculate the 'expected yield of ultracold
neutrons and the induced activity in various parts of the
equipment, the distribution of neutron flux along the tan-
gential channel was measured by the method of radioactive
indicators 18]. For relative measurements we used copper
indicators (5 mm in diameter and 15 u thick). The absolute
J-
100
1 values of the neutron flux were obtained by measurements
100 0 100 200 T 100 with gold indicators 0.26 mg/cm2?thick at the ends of the
Length, cm . -... ,i I , tangential channel=and in one of.the,?vertical channels. The
thermal neutron flux distribution was determined from
Fig. 2. Thermal neutron flux distribution measurements with indicators;=.either,inside cadmium
along the open tangential reactor channel jackets or without. any, cover.
for two configurations of the reactor active
Figure 2 shows the thermal-neutron flux distribution
zone.
along the tangential channel. In,the center of the channel
the thermal-neutron flux equals (2.4 ? 0.3) ? 1012 and (4.8
?0.4) ? 10'2 neutrons /cml ? sec respectively for two typical
configurations of the active zone of the reactor. The mean square errors of measurement, allowing for
the normalization error, are indicated in the figure. The installation of a water moderator. 175 mm in
diameter and 300 mm'long in?ide'the channel;, contrary'lto expectation, -did not lead to any marked increase
in the thermal neutron'flux.
The energy spectrum of the neutrons, in the. tangential channel'was -notmeasured.
If we -consider'that'a Maxwelliari neutron energy distribution-is established in the converter, with a
value of EN - kT, ?tlien adcoiding-to' [4] the maximum-flux of ultracold?neutrons is
t7l.. ' (I)UCN? Q)p (Lliml' Vrrtcool (2)
8 \ IN 1 .aa I cheat ` I r, in ~.
where 4 is the thermal neutron flux, Elim is the limiting value of the energy, T is the effective tempera-
ture of the incident neutron spectrum, acool? is the scattering cross section involving the transfer. of energy
from the neutron to the converteraveraged over; the thermal spectrum, cheat is the scattering cross sec-
tion involving the transfer of energy from the converter to. the neutron, as is the neutron capture cross
section. For example, in the case of an aluminum converter at 400?K,- the flux of the ultracold neutrons,
r 1 . r,,i
1
200
100 200 300 ;?i p11, mm Hg
Fig. 3 Fig. 4
r . - ,
Fig.. 3. Dependence of the ultracold-neutron count rate (counts ?p'er 1(0'sec) '
,. .. , , ._.. J' .. - .,
on the temperature of a magnesium converter. The theoretical curve is
normalized to the experimental value of'I at roon- temperature.
Fig. 4. Dependence:of the ultracold-neutron. count rate (counts per 100 see),
onthe helium pressure in the neutron guide. The theoretical curve is nor-
malized to the experimental value of I for p = 0.
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 Declassified and Approved For Release 2013/02/21 : CIA-RDP10-02196R000400040001-6
TABLE 1. Ultracold-Neutron Count Rate close to the converter equals 5 neutrons /cm2 ? sec, which,
rate
-
; sec-1
Converter
tempera-
back- back--
tire. *K
ground
effect
I
ZrH 1.92
-
'
50?
1
20?1
H2O in aluminum ampoule
290
?
3,00,2
6,4?0,2
Aluminum ampoule
-
2,0?0,3
4,0?0,2
Magnesium
320
5,0?0,2
11,2?0,4
on allowing for the transmission of the neutron guide
(^-0.2), agrees satisfactorily with the observed value of
the ultracold-neutron flux at the outlet of the neutron
guide.
Ultracold neutrons were first obtained in the fore-
going apparatus and duly recorded in July 1971. By way
of converters, use was made of water flowing through an
aluminum container with a front wall 250 p thick (water
converter), zirconium hydride ZrH1.9, and magnesium.
Table 1 shows the ultracold-neutron count rate recorded
by a scintillation detector with a working area of 14 cm2 for a reactor power of 10 MW. In a four months
work with a water converter and an average reactor power of 2.5 MW, theultracold-neutron count rate re-
mained constant.
Figure 3 shows the dependence of the ultracold-neutron count rate on the temperature of the mag-
nesium converter. The experimental results agree with the calculated curve [9].
In order to estimate the character of the ultracold-neutron diffusion from the converter to the detec-
tor, we measured the relationship between the ultracold-neutron count rate and the helium pressure in the
neutron guide (Fig. 4). The resultant relationship I(P$e) shows that the intensity of the ultracold neutrons
falls by a factor of two if the helium pressure in the neutron guide equals 60 mm Hg. This confirms the
data of (4] and may indicate a considerable proportion of specular reflections from the tube walls.
A vacuum valve between sections 17 and 19 of the neutron guide with an open aperture 100 mm in
diameter for the ultracold neutrons reduced the ultracold-neutron count rate by a factor of two. This also
indicates the considerable proportion of specular reflections from the tube walls during the passage of the
ultracold neutrons through the neutron guide.
A gradual but significant fall in the intensity of ultracold-neutron recording was noted in [4] over
several months; in the opinion of the authors this was associated with the contamination of the surface of
the neutron guide as a result of radiation-induced corrosion. In the apparatus here described there was no
marked reduction in ultracold-neutron yield over six months (for an average reactor power of 2.5 MW).
The ultracold-neutron yield may be increased by cooling a number of converters to liquid-nitrogen
temperature, and also increasing the thermal-neutron flux close to the converter. The efficiency of the
apparatus as a whole may be greatly increased by using ultracold-neutron detectors with a better efficiency
and a greater working area and by improving the construction of the system. According to estimations this
should enable us to record as far as 103 ultracold neutrons per sec on the working surface of the detector.
The results of our measurements of the ultracold-neutron yields from several converters are now
being analyzed.
Thus the foregoing results confirm the conclusions of [4] as to the possibility of extracting ultracold
neutrons from the channels of a stationary reactor. The favorable characteristics of the arrangement
utilizing the open horizontal tangential channel of a VVR-K reactor enable us to solve a number of both
methodical and physical problems.
The authors wish to thank I. M. Frank for interest in the work and also all who assisted in its execu-
LITERATURE CITED -
1. Ya. B. Zel'dovich, Zh. Eksp. Teor. Fiz., 36, 1952 (1959).
2. V. V. Vladimirskii, Zh. Eksp. Teor. Fiz., 39, 1062 (1960).
3. V. P. Lushchikov et al., JINR Preprint, R3-4127, Dubna (1968); ZhETF, Pis. Red., 9, 20 (1969).
4. L. V. Groshev et al., JINR Preprint, R3-5392, Dubna (1970).
5. A. Steyerl, Phys. Letters, 29B, No. 1, 33 (1969).
6. A. V. Antonov et al., Trudy Fiz. Inst. Akad. Nauk SSSR, 57, 270 (1972).
7. A. V. Antonov, D. E. Vul', and M. V. Kazarnovskii, ZhETF, Pis. Red., 9, 307 (1969).
8. R. B. Novgorodtsev et al., in: Transactions of the Second Coordinating Congress on the Dosimetry
of Heavy Doses [in Russian], Izd. FAN, Tashkent (1966), p. 125. -
9. V. V. Golikov, V. I. Lushchikov, and F. L. Shapiro, JINR Preprint R3-6556, Dubna (1972).
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 Declassified and Approved For Release 2013/02/21 : CIA-RDP10-02196R000400040001-6
V. L. Serov, Yu. F. Orlov, UDC 621.384.6
Accelerated widely spaced dense charged-particle bunches can have various uses (both scientific and
applied). For example, impact countercollisions of electron and positron bunches allow. one to attain coun-
terreaction luminosities which cannot be attained by other methods (with comparable rf-power expenditures)
The process of accelerating dense bunches is accompanied by intense coherent energy radiation in a
resonance cavity at the oscillation natural modes [2-8); this reduces the acceleration efficiency and leads
to a need for compensating significant integral energy losses by increasing rf-generator energy.
The integral energy-loss magnitude from a real elongated bunch, like the energy obtained by it from
the field created by an external generator, depends on the number of particles and the bunch and resonator
geometries.
The present paper considers the problem of optimizing energy transfer from the generator to the
accelerated bunch, using as an example a system of independent resonators. We assume that the bunches
are rather widely spaced and the generators are turned on only for a certain time Tg before a?bunch appears
in a given resonator and are turned off immediately after its passage. The rf energy remaining in the
resonator after the bunch passes is scattered in its walls and coupling unit in the time interval when the',
generator is not operating. Thus, the bunches are not linked to each other through the resonator field, so
that the problem posed is solved for a single bunch.
Below, we optimize the quantity Tg and find the optimal acceleration regime for a linear. accelerator
and also for proton and electron ring accelerators.
In the time Tg, there accumulates in the resonator field energy Wr, which is then partially absorbed
by the bunch in time At (we assume that At > 1. In the time 6t with the generator turned off, it is easy to
achieve the necessary accelerating-field phase change (always less than 7r radians) so that the foIllowing
bunch will arrive at the right phase.
When the magnetic field increases linearly (H = const), the particle-energy increment per cycle is
(AE)cy = e/c(LORH = const, where L0 is the orbit length. In this case, all the equations in the preceding
section are satisfied with the difference that n now denotes the number of resonators in the orbit and E in
Eq. (2) must be replaced by (AE)cy.
We will consider a ring accelerator for opposed electron and positron bunches. In this case, due to
synchrotron radiation, the necessary value for energy transfer to a particle per cycle (AE)cy does not re-
main constant, but increases sharply toward the end of acceleration. Therefore, one cannot 'obtain an opti-
mum, constant efficiency value over the entire acceleration cycle. Taking this into account, we optimize
the r-power magnitude, averaged over the acceleration cycle. We will take E, LO, N, and a as given; the
number of resonators in the orbit n is subject to optimization. Energy losses to resonator excitation, on
the one hand, are proportional to their number, while, on the other hand, the rf-energy magnitude needed
to accelerate a small number of particles (when losses to coherent radiation in the resonators may be dis-
regarded) falls in inverse proportion to the number of resonators in the orbit. However, if the number of
resonators increases due to decreased magnetic length 2irR, then the rf-energy dispersion increases to
compensate for synchrotron radiation.
The generator energy expended during an acceleration cycle (for kmax particle revolutions) and opti-
mized with respect to Tg, is
kmax R max
W, z 1.22nin u Wr (k) = 0.61Cnin VI (k) dk ; (13)
k=1 0
here n1 is the number of bunches in the orbit (in the case considered, n1 = 2); C is the resonator equivalent
capacitance. Now, we make the following definitions: l 1
V r (k) = cos W [ ' n)cy + rzeo \ ? ' h1~k) (14)
eMM1
*A. A. Naumov, whom the authors thank, pointed out that in the method using widely spaced impact colli-
sions, which requires strong focusing of opposed particles, measurements are difficult at small scattering
angles, which is a shortcoming of this method.
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 Declassified and Approved For Release 2013/02/21 : CIA-RDP10-02196R000400040001-6
[AE (k) Icy= 30RAH (k) -[-0.88.107 E Rk) + AEother, (15)
where R is the magnetic-track radius of curvature, in km; AH is the field increment per cycle, in kG; E
and AE are expressed in gigaelectron-volts. The second term on the right describes synchrotron radia-
tion in one cycle. The quantity DEother includes all possible energy-loss mechanism, except for synchro-
tron and coherent radiation in the resonators.
When the bunch-acceleration cycle frequency is f, the average rf-system power is 31 = WEf. The
condition for the average-power minimum takes the form:
'max
a tt n VI (k) dk = 0
hmax
dk r Sez.+(AE)cyj l 8e2N,l _ (AE)cy L2 O(AE) OR
0 l nepa n neoa n OR y n ] - 0.
(17)
For the case when the magnet radius of curvature R = const is considered given, we can find the following
expression from Eq. (17):
.1_..dk (AE)'Cy
Lacc= hnopt = I a 0
max
dk (Nip)2
0
The corresponding minimal value for the average rf power is:
k max
7EOah
8e2
1.22/ f _np t (AE 8e2N ) dk.
4.8.10'- e211124nc o (M 1cos (P)2 n opty + neoa V
If the number of particles N, the bunch length hi, and the phase value q' do not change
cycle, then Eq. (19) takes on the form:
'max 'max
(1,2 S
F k) = kmax dk (AE)~y) J (AE)cydk;
0 0
( h h1
Q1=T _/
M2Mi h
Clearly, the minimal-power value is not directly dependent on the wavelength, but is a function of the
parameters h/A and hl /A, the assigned energy increment in the cycle, and the equilibrium phase. How-
ever, the bunch length cannot be assigned arbitrarily; it depends on the phase-stability-region magnitude,
which is 27rh1,Ati 3c'. The smallest value for the function '/cos2co occurs when co ,r 30-40?.
If the breakdown voltage (V/h)all can be considered given and independent of A, then we again obtain
A2 - N. Actually,
2N e (V) (AE)cy+ Sneoa nopt
Mall /all 1> Lacc (21)
If we take Nnopt/a from Eq. (18), then the numerator of this ratio will not depend on N or A. Consequently,
Lacc - h/V is independent of N and A, and we obtain from Eq. (18); A2 - N.
Equation (18) remains approximately true in the case when the assigned quantity is not R, but the or-
bit length Lo, if only the number of particles is large enough that Lacc 711, then
one should use the general equation, Eq. (17). In particular, if we wholly disregard coherent radiation
NfF(k)c I h . _h!
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 Declassified and Approved For Release 2013/02/21 : CIA-RDP10-02196R000400040001-6
in the resonators (and also magnetic-field variation over time), then when Lo = const we obtain the familiar
condition
hnopt= aR.
We will consider a synchrotron with opposed electron and positron bunches N- = N+ = 2.1013; E = 100
GeV; R = km, magnetic-field sinusoidal-oscillation (with magnetiziation) frequency fH = 200 Hz; Hmax
= 0.875 kG; Hej = 1.67 kG. In this case, at the end of acceleration, the losses to synchrotron radiation per
cycle are AEy = 4.4 GeV and the particle-energy increment at the end of acceleration is U = 1.6 GeV. The
relative energy fraction expanded on synchrotron radiation is rather small and equals Wy/E - 0.2, A /U
= 0.55. When A = 75 cm, h/A = 0.25, and h1/A = 0.25 (in principle, h1 can, with the help of damping, be
made almost constant over the acceleration cycle) from Eq. (18) we obtain n 8.103 and hn 1.5 km, i.e.,
hn ?aR = 6.3 km. The losses from a single particle to coherent radiation in all resonators are 4.14 GeV
/cycle; (Vr)max = 1.8 MeV; Vr/h = 96 kV /cm; the total peak power without generators is equal to 4.5 ?103
The average power is 35 = 2.2 -106 f W, where f is the bunch-accelerations frequency, f 0, bj, M.
Let, as a result of the calculation of the kinetics, there be known the dependence of the neutron multi-
plication constant on the burnup k(s), in particular, let it be given by the polynomial k(z). $ Then the con-
straint on the reactivity can be represented in the form
(k (z), zxj (z)) = (Si jz+k (z), x)> k (0 < j N),
where the operator z+ reduces the power of z by one power, and the zero power of z vanishes.
The constraints (4) and (5) are inconvenient because they depend on the phase variable x, and not on
the control u. Taking into account that each of them is given by a linear functional of the form (f, x), with
the help of the adjoint functions
V -z+j1(z)=f (z) (0
um-i-m (N-M+1-i
(8)
(9)
The relations (8) have the following sense: fresh fuel, loaded at moment i, has burnup m at moment i + m,
where the quantity of fuel drawn off at all moments i + m should not exceed the value u?. The sign of the
equality arises in connection with the fact that the maximum admissible burnup equals k.
The functions l/ (z) _ km+izm-j +1 correspond to the functionals in expression (5), and the
1 M-1>m)j-{
constraints on the reactivity take the form
kt-j+lui- k'-j+m+1 um> k, di,
i>j>i-L(i)+I m=1
i i GM-1,
L (i)
M-1 i> M.
The calculation of the minimum of the total fuel consumption
``N
PUi
as a function of the components of the vector u for linear constraints (3), (8)-(10), represents a problem in
linear programming [51. If in addition we require that ui = uiMS = 1/MS from some moment i >- NS, then
at the moment NS + MS (or earlier).the reactor will be found in a stationary regime. There exists some
standard method of bringing the reactor to a stationary regime through MS intervals: uo = 1; ui = ui = 1
/MS (1 s i 5 MS), requiring a single excess loading [in the discrete representation (MS-1)/MS] in com-
parison with an infinite stationary regime. Therefore, to calculate an optimal strategy, the topic of dis-
cussion is the economy of a certain part of this loading OP = Pstand-P < 1.
In Fig. 1 we, represent a typical example of an optimal strategy,* calculated based on a VORS pro-
gram (Fortran, BESM-6). The values of km for AT = 600 MW -days/ton were determined based on the
data of Table 1, the value of MS was assumed equal to five (this corresponds to a burnup of 3000 MW-days
MS
/ton in a stationary regime), M = 7, k = kstat = I km/MS, OP = 20%; the moment NS was not fixed.
1
In Table 2 we represent optimal values of OP for division of the burnup in the stationary regime with
equilibrium burnup of 3000 and 6000 MW *days /ton for eight intervals (MS = 8), where for the last column
we replace the dependence represented in Table 1 by the linear function k(s).
On the basis of the calculations performed we can draw the following conclusions. The form of the
function k(s) has a very weak effect on the magnitude of P. If the admissible burnup does not exceed its
maximum value in the stationary regime (M = MS), then the gain of OP proves to be zero. To obtain the
principal part of the gain it is sufficient to increase the admissible burnup by 10-15%c; its further increase
(up to ^'40%) leads to a rather weak variation of P. For the optimal output in the stationary regime we
require about two and a half runs (see Fig. 1). If we require a more rapid approach in the stationary re-
gime, i.e., if we must reduce NS, then OP at first will vary weakly, and then sharply fall to zero.
*The function k(s) represented in Table 1 is characteristic of the reactor of the A-1 Atomic Electric Power
Plant in Czechoslovakia [6]:-
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 Declassified and Approved For Release 2013/02/21 : CIA-RDP10-02196R000400040001-6
LITERATURE CITED
1. B. L. Ioffe and L. B. Okunt, At. Energ., No. 1, 80 (1956).
2. D. Tabak, IEEE Trans. Nucl. Sci., NS-15, 60 (1968).
3. T. Sauar, Nucl. Sci. and Engng., 46, 274 (1971).
4. A. Suzuki and N. Kiyose, ibid., 11.
5. S. I. Zukhovitskii and L. I. Avdeeva, Linear and Convex Programming [in Russian], Nauka, Moscow
(1967).
6. V. M. Abramov et al., At. Energ., 36, No..3, 163 (1974).
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 Declassified and Approved For Release 2013/02/21 : CIA-RDP10-02196R000400040001-6
Since 1968, a beam of protons with energies from 70 to 200 MeV has been used at the Institute of
Theoretical and Experimental Physics for biological studies and for radiation therapy of cancers. The
pulse duration of the extracted proton beam is 1.6 psec, the repetition rate 0.25 Hz, the intensity range
2 ? (108-1011) protons/pulse; the transverse dimension of the beam varies from 3 to 90 mm, and the energy
range is 70-200 MeV [1, 21.
The performance of biological studies required the creation of an operational system providing mea-
surement of intensity in practically single beam pulses (reciprocal duty factor, 25 ?105) under conditions of
strong electromagnetic interference produced by the proton extraction system at the very time of mea-
surement. The requirements imposed on the measurement system were determined not only by the tech-
nical conditions but also by the particular use of the beam. During radiation therapy, the measuring sys-
tem is a medical instrument upon the accuracy, operational simplicity, reliability, and flexibility of which
the patient's health depends.
The system must ensure measurements having an accuracy better than f3% over the entire range of
energies, intensities, and transverse beam dimensions without any readjustment of the input device (sen-
sor). The sensor must not have any effect on beam characteristics. The method for monitoring the effec-
tiveness of the system must be simple and ensure checking of the entire channel. The method for con-
structing the measuring system is determined mainly by the choice of the sensor which directly senses
the beam [3].
The so-called transmission sensors - electromagnetic and electrostatic - fill these requirements
most completely. The use of an electromagnetic sensor in the system developed (current transformer)
made it possible to create an interference-proof system for operational measurement of the intensity of
the biomedical proton beam using comparatively simple equipment.
Fig. 1. Electromagnetic sensor.
Translated from Atomnaya Energiya, Vol. 37, No. 1, pp.. 69-70, July, 1974. Original letter sub-
mitted July 2, 1973; revision submitted February 4, 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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 Declassified and Approved For Release 2013/02/21 : CIA-RDP10-02196R000400040001-6
Sensor I PAI IE1EiPit
Final r----=------71
amplifier \ ~ II II ~
"I I
Times
pulse
Gating
circuit
Atten-
tuator
Output
signal
Matching
circuit
Fig. 2. Block diagram of system for operational mea-
surement of intensity.
Tuner pulse Amplifier pulse The electromagnetic sensor (Fig. 1) consists of the
permalloy ring 1 (external and internal diameters 168 and
120 mm, width 52 mm) with a winding uniformly distributed
Delay Discrimin- IDiscrimin- over its circumference. The winding is made of PELShO-
circuit ator,l ator 11 0.14 conductor; the number of turns is 89 and the sensor
inductance is -14 mH. The sensor is inside the duralumi-
num housing 2. Openings in the housing covers correspond
to the internal diameter of the ring. In the internal cavity
shaping : H Coincidence of the ring, there is the metal bushing 3 one edge of which
circuit circuit i is directly adjacent to the cover 4. Between the second
edge and the cover there is the circular plastic washer 5,
which is 2.5 mm thick and creates a "slit" in the solid shield
Delay providing interaction between the beam and current trans-
circuit former. The emf induced in the transformer winding deter-
mines the voltage pulse at the sensor output. The amplitude
of the voltage is proportional to the intensity of the proton
beam. The operating principles of an electromagnetic
shaping coincidence sensor are described in detail in [3] and reasons for the
circuit circuit II
choice of design are presented.
To record the pulses from individual signals, it is
convenient to use the VCh-17 pulse digital voltmeter, which
Trip Control Delay
circuit is commercially produced. The minimum pulse amplitude
circ uit circuit
recorded by this instrument is 0.05 V which calls for the
Gate pulse use of special amplifiers (Fig. 2). The choice of electronic
Fig. 3. Block diagram of gate unit. units for the measuring system and their layout is mainly
determined by the presence of high-intensity pulsed inter-
ference produced by the beam extraction devices. To re-
duce these effects to a minimum, the electromagnetic
sensor is thoroughly shielded; the preamplifiers PAI (K = 7) and PAIL (K = 30), the delay circuits, and
the matching circuits are located directly in the sensor housing 7 (see Fig. 1). Connection to the final
amplifiers at the control panel and the power feed to the electronic circuits is accomplished by means of
a special cable through the 2RM14B4Sh1V1 connector 6 (see Fig. 1) attached to the sensor housing. The
signal is fed over a shielded RK-75-2-12 cable which accompanies the power conductors and is enclosed
in a common external shield. - The signal is fed to the VCh-17 by the amplification stages AIII (K = 7)
and AIV (K = 12) of the final amplifier through the gating circuit. Attenuators are introduced into the
preamplifier and final amplifier in order to expand the dynamic range of the amplification channel. The
preamplifier attenuators have two fixed values (1 : 1 and 1 :10) and the final amplifiers attenuators have
four fixed values (1 :1, 1 : 2, 1 :5, and 1 :10). The gating circuit (Fig. 3) opens the recording channel only
at the time of a useful signal. For effective operation of the gating circuit, a circuit to delay the signal by
8 psec is introduced into the preamplifier.
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The introduction in the gate unit, of circuits to discriminate against amplifier pulse height. (two dis-
criminators with thresholds Uthresh I = 0.2 V and Uthresh II = 0.7 V) and the selection of this pulse in time,
made it possible to identify a useful signal from the amplifier at the output of the gating circuit and to en-
sure effective selection of it. These measures completely protected the measurement system from inter-
ference, which made it possible to automate control of their irradiation procedures - to carry out the ac-
cumulation of a total dose with automatic beam shutoff in accordance with a previous setting, to perform
programmed control (with respect to dose) of patient rotation, etc.
To check the dynamic range of the measuring system and to evaluate the accuracy of the attenuators
and the stability of the entire system, a VCh-7 pulse digital voltmeter was used having a basic error A
= ?0.5% Umeas ? 1 ml of discharge.
In studies, the beam was simulated by a current pulse from a G5-15 oscillator transmitted along a
conductor passing through the sensor opening. Measurements were made at a pulse repetition rate of 1
kHz. The tests show the error in measurements of proton beam intensity over the range 2 ? (108-1011)pro-
tons /pulse, does not exceed f 3%.
In order to determine the accuracy of direct measurement of proton beam intensity, a comparative
calibration of the measuring system was made using the induced activity in 12C for a broad range of beam
intensities. The test results indicated agreement within the limits of accuracy of the method of induced
activity (?6%).
LITERATURE CITED
1. S. I. Blokhin et al., in: Dosimetry and Radiation Processes in Dosimetric Systems [in Russian],
Fan, Tashkent (1972), pp. 71-75.
2. V. S. Khoroshkov et al., Meditsinskaya Radiologiya, No. 4, 56 (1969).
3. V. G. Brovchenko et al., Electronic Devices in Electrostatic Accelerators [in Russian], Atomizdat,
Moscow (1968).
753
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A.
S.
Belousov, E.'I. Malinovskii,
S.
V.
Rusakov, S. P. Kruglov,
and
V.
D. Savel'ev
The radial distribution of absorbed energy and the transmission curve in lead were measured for a
beam of 31-GeV electrons from the Serpukhov synchrotron [1]. Measurement of the radial distribution of
absorbed energy, E(r, t), was made with LiF thermoluminescent detectors (TLD) placed within a lead ab-
sorber. The absorber was made of lead disks 200 mm in diameter. Since the TLD are small (diameter
3.5 mm and thickness 1.2 mm), their introduction into the absorber satisfies the requirements of the Bragg
-Cray principle [2] and consequently permits a determination of the energy losses in the surrounding me-
dium from the energy losses in the TLD. In addition, shower distortion associated with transition effects
at the LiF - Pb boundary is negligibly small according to [31 for our detectors having a thickness 0.6.10-2
X0.
Figure 1 shows the radial distribution E (r, t) of absorbed energy for various absorber thicknesses.
The measurement error is 3o%c or less.
The transmission curve 11(t) can be obtained from the relation
E (r, t) r dr
11(t)
1 E (r, t) r dr dt
0 0
10 15 r, X0 0 5 10 15 20
Fig. 1 Fig. 2
Fig. 1. Radial distribution of absorbed energy for various values of t: 1) 3.4 X0; 2)
6.2 .X0; 3) 13.7 X0; 4) 16.4 X0; 5) 19.6 X0; 6) 25.5X0 .
Fig. 2. Transmission curve: 0) experimental data from this work; A) theoretical
data from [4]; ----) exponential function with an attenuation factor ?min = 0.47 cm 1.
Translated from Atomnaya Energiya, Vol. 37, No. 1, pp. 71-72, July, 1974. Original letter sub-
mitted July 9, 1973.
?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 and, 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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r, Xo A numerical method was used for the integration.; the result-
- .. 1.,.11..u11.?ivu a.ut vc is D11VWi1 111 r%. G. 111 Lne caiCUlaLlons, a
radiation lengthX0 = 0.562 cm was used. Beyond its maximum, the
experimental transmission curve is given by an exponential function
with an attenuation factor p = 0.48 ? 0.01 cm-1, the value of which
is in good agreement with the value for the minimum attenuation
coefficient in lead ?min = 0.47 cm-1.
Results of an analytic calculation of cascade curves have been
4 - ~~~~",-
:- ( given [4], i.e., the total number of shower electrons as a function
of lead absorber thickness, -5e(t), for primary electrons having
2 - I energies from 3.108 to 1014 eV. The calculation took into account
the Compton effect, the effect of incomplete screening on the nair-
Fig. 3. Isoenergy curves: 1) 30%;
2) 40%; 3) 50k; 4) 60%; 5)70!c; 8)
80%; 7) 85%; 8) 90%.
/ ~e (t) Eo
rip lt) = Eo 7
and is shown in Fig. 2. A value Fo = 7.6 MeV, the same as that used in [4], was used in the calculations.
One of the reasons for the difference between the theoretical transmission curve and the experimental curve
may be the fact that in the calculations the ionization losses were assumed to be Eo = 7.6 MeV while the
energy lost by electrons over a path X0 exceeds the critical energy for small thicknesses.
To determine the, efficiency of a total absorption detector, it is necessary to know the value of the
absorbed energy
r t
AE S 1 ` E(r, t)rdrdt
E0 (?
S J E(r, t)rdrdt
0 0
for cylindrical volumes of given dimensions. Isoenergy curves are given in Fig. 3 which were obtained
from experimental results.
In conclusion, the authors thank M. Ya. Borkovskii for valuable discussions.
LITERATURE CITED
1. S. S. Gershtein et al., IFVE Preprint 72-93, Serpukhov (1972).
2. F. Spiers in: Radiation Dosimetry, G. Hine and G. Brownell, editors [Russian translation], Izd.
Inostr. Lit., Moscow (1958).
3. K. Pinkau, Phys. Rev., 139, No. 68, 1548 (1965).
4. Z. Buja, Acta Phys. Polon., 24, No. 3, 381 (1963).
photons and electrons with energies