JPRS ID: 10603 USSR REPORT ENGINEERING AND EQUIPMENT

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• Content Type: REPORTS

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JPRS L/'106Q3
21 JuNE 1982
U SS Re or#
p
- ENGINEERING AND EQUIPMENT
(FOUO 4/82)
Fg~$ FOREIGN BROADCAST INFORMATIO~I SERO/ICE
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JPRS L/10603
21 June 1982
USSR REPORT
ENGINEERING AND EQUIPMENT
(FOUO 4/82)
CONTENTS
AERUNAUTICAL AND SPAG'E
Rocket System Gas Generators 1
NUG'LEAR ENERGY
FB.St. Hydrogen Atom In~ectors
Use of Burning ~bsorbers in F{BMK 15
Choice of Extractors and F"i.lters for Plants Reprocessing
Spent F1ze1 Elements of Atomic Power Stations 22
Ra,diation Damage to Steel in Water-Moderated, Water-Cooled
Reactor Vessels 32
Radiation Safety and Protection of Nuclear Power Plants ............e 37
NON-NUCLEAR ENERGY ,
Industrial Magnetohydroc~ynamic Equipment and Processes 53
INDUSTRIAL TECHNOLOGY
Strength and Reliability of Technical Devices 55
TURBINE AND ENGINE DESIGN
Optimal Last Stage Design of High-Power Steam Turbines 64
- a- LIII - USSR - 21.F S&T FOUO]
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N A~T~ GATI ON AN D GUI DAN GE S YSTEMS
Correcting Inertial Guidance Sys~ems by IIaing Combined
Subsidiary Positional and Velocity Information 67
FLUID l~CHANICS
Supersonic Flow Pertur'bations With In~ection of Mass and Heat .....e.. 76
TESTING AND MAT~RIALS
Investigating Effi ciency of Sliding Bearings in Helium Environment 78
Adaptive Measuring Ins~,ruments 86
�
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AERONAUTICAL AND SPACE
UDC 629.7.064.2
ROCKET SYSTEM GAS GENERATORS
_ Moscow GAZOGENERATORY RAKETNYKH SISTEM in Russian 1981 (signed to press
14 Apr 81) pp 2-4, 152
[Annotation, prefac~ and table of contents from book "Gas Generators for
Rocket Systems", by A1'bert Alekseyevich Shishkov�an3 Boris Vasil'yevich
Rumyantsevi Izciatel'stvo "Mashinostroyeniye", 1183 copies, 152 pages]
[TextJ The book gives a systematic presentation of principal designs, charac-
teristics and peculiarities of. working processes in gas generators using chem-
ical fuel (liquid, solid and mixed) that are sources of power and gas ~ets
on flightcraft and in ground-based rocket systems. Methods of experimental
. development of gas generators are briefly covered.
`The book is intended for engineers and designers in ttie field of rocketry.
Pref ace
Gas generators are extensively used in rocketry. Their major components are
many ways similar to those of the main rocket engines; however, the working
processes in gas generators have important distinguishing features that must
be taker into considerati~n in design and development.
In connection with the development of rocketry and expansion of areas of appli-
cation of gas generators in recent years, a number of patents and journal
articles have been published with results of research on gas-generator devices
[Ref. 12]. Condensed information on gas generators is given in handbooks
on the principles of rocket engine design [Ref. 2, 3]. However, as~a whole,
the published materials on gas generators are disconnected, fr2.gmentary and
procedurally nonuniform.
In this book the authors attempt to syste.matize the designs and particulara
, of workir.g processas of gas generators using different fuels based on major
pri.nciples of the theory of rocket engines.
_ The book has five chapters. Chapter 1 presents the basic characteristics
of gas-generator devices and the fuel compositions used in them, giving indi-
vidual attention to methods ~f model, laboratory and stand tests.
1
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Chapter 2 deals with pecullarities o~: gas flows in gas generator, ~~s line
and exhaust nozzles, and methods of calculating gasdynamic characteristics
of gas generators, and also take~ up problems af filtration of gas generation
r~oducts and gas-ther.nodynamic proce4ses in the devices that are used.
- A survey of materials on the peculiarities of devices and principles of calcu-
lation of one-component and two-component liquid gas-generators, as well as
gas generators based on fluidized powder fuel is contained in chapter 3.
Chapter 4 is devoted to designs, methods of intraballistic calculation and
different possible methods of regulating solid-fuel gas generators (particu-
larly with end-burned charges), including gas generators that can be repeatedly
_ energized. A solution is given for the problem of transient processes with
variation of certain parameters.
Tt~e last chapter deals with problems of development of a variety of combined
gas generators using solid (with separate components), quasi-hybrid and hybrid
~ fuels, vapor-gas genPratoxs, and gas generators for straight-flow rocket en-
gines and turborocket engines; engineering methods are given for calculating
the ma~or characteristics of some gas generators.
Chapters 1, 4 and 5 were written by both authors jointly, chapter 2-- by
_ A. A. Shishkov and chapter 3-- by B. V. Rumyantsev.
The authors are sincerely grateful to the reviewer, Doctor of Technical Sci-
~ences A. P. Tishin, for constructive recommendations that have improved all
parts of the manuscript, and to Candidate of Technical Sciences M. Ye.
. Yevgen'yev for useful advice in solving the problems in various sections.
We would be grateful to readers for comments and suggestions, which should
be sent to the pub~ishers: l~ashinostroyeniye, i07076, Moscow, Stromynskiy
- per. 4.
Contents page
Preface 3
~hapter 1: General Information on Gas Generators ~
1.1. Fields of application and principal characteris~ics of gas generators 5
1.2. Gas-generator fuels 13
~ 1.3. Gas-generator tests 28
Chapter 2: Gas-Thermodynamics of Gas Generators 37
2.1. Equations of the gas generator 37
2.2. Characteristics of the gas channel 42
~ 49
. 2.3. Filter~ng ~ombustion products
- 2.4. Gas-thermodynamic processes in devices fed by gas generation products 54
Chapter 3. Liquid Gas Generators 64
3.1. One-component gas generator 64
- 3.2, Two-component gas generator 6~
- 3.3. Use of fluidized fuels in gas generators 75
Chapter 4: Solid-Fuel Gas Generators 78
4.1. Desigr~ of solid-fuel gas generators 78
4.2. Intraballistic calculation of the gas generator 84
2
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4.3. Methods of regulating solid-fuel gas generators 92
4.4. Transient processes in the solid-fuel gas generator 103
Chapter 5. Combined Gas Generator Designs 111
5.1. Solid-fuel gas generator with separate components 111
5.2. Controllable gas generator using quasi-hybrid fuel 113
5.3. Gas generator using hybrid fuel lZ~
5.4. Vagor-gas generators 133
~ 5.5. Gas generators for straight-flow rocket engines and turborocket
engines 141
_ References 149
COPYRIGHT: Izdatel'stvo "Mashinostroyeniye",~1981
6610
CSO: 1861/192
- 3
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NUC'LEAR ENERGY
UDC 621.039.616
FAST HYDROGEN ATOM INJECTOR6
Mosaow INZI~KTORY BYSTRYKH ATQMQV VODORODA in R~assian 1981 pp 2-11, 18, 152
tExcerpts from the book "Fast Hyclrogen Atom Injectors", by N. N. Semashko,
- A. N. Vladimirov, V. V. Kuzn~:tso~~, V. M. Kulygin and A. A. Panasenkov,
Energoizdat, 168 pages]
[Excerpts~ Annotation
Equipment for.producing fast hydragen (deuterium) atams is described for.
_ heating plasma to thermonuclea~: tf~anperatures in closed and open magnetic sys-
tems. Powerful injectors for ~thex~nonuclear research, demonstration and reac-
- tor installatians are consider~ed. Methods of pr~~ucing flows of low-energy
atoms based on charge exchan~~.: of positlve ions ~:ad flows of inedium-energy
ator.is based on conversion af negative ions are described. Z'he book is i:z-
tended for scientific worker~s and engineers working in the field of controlled
thremonuclear fusion, accele~rator t~chnology and space technology. It may be
useful to students and grad~:~ate students of ghysical engineering specialtiss.
Contents Page
Chapter 1. Inj~:ctian of Fast Atoms--A Method af Creating a 5
Thermonuclear Plasma . . . . . . . . . . . . . . . . . . . . 6
1.1. Types of therm~~nuclear installations with injection 8
1.2. Requirements an inj~2ction systems . . . . . . . . . . . . . . 9
1.3. Design principles of injection systems. . . . . . . . . . . .
11
Chapter 2. Structure of. the Inj ector. . . . � � � � � � ' ' ' ' ~ ~ ~ ~ 11
2.1. Injector wit'n production of fast atoms from positive ions 15
2.2. Production c~f fast atoms by canversion of negative ions 18
2.3. Injector su�port systems . . . . . . . . . . . . . . . . . .
. . . . . . .
Chapter 3. Ion Sourcfas . . . . . . . . . . . . . . . . . 30
- 3.1. Basic concepts . . . . . . . . . . . . . . . . . . . . . . 30
3.2. Extractior~ and formxti~~~n of ion beam . . . , . . . . . . . . 34
3.3. Plasma em'ltter of posiLive ions . . . . . . . . . . . . . . . 54
. . . . . �
3.4. Types of ion sources. . . . . . . . . . . . . . . 67
3. 5. Negative ion sources . . . . . . . . . . . . . . . e � � � � � 88
_ � 4
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F~
Chapter 4. Ion-Atom Channel of Injector . . . � � � � � � � � � � � . 94
_ 4.1. Divergence of ion beam with uncompensated charge 94
4.2. Ion beam transpcrt . . . . � � � � � � � � � � � � � � � � . 96
4.3. Ion beam charge exchang~ . . . � � � � � � � � � � � � � � . 103
- 4.4. C~lculating transit of neutral beam through injector
channe 1 . . . . . . . . . . . . . . . . . . . . . . . . . . 111
4.5. Beam separation. Ion energy recovery . . � � � . . . . . . 114
.~.6. Fxeacceleration of negative ions . . . � � � . � . . . . . . 120
4.7. Reionization losses in atomic beam . � � � � � � � . . . . . 124
Chapter 5. Injector Based on Dir~ct Charge Exchange of Positive Ions . 128
5.1. Initial prerequisites . . . � � � � � � � � � � � � � � � � 128
5.2. Structural diag~am of injector . . . . . . . . . . . . . . . 131
5.3. Coanponents of ion-atom channel . . . � � � � � � � � . . . . 133
_ 5.4. Cryogenic vacuum system . . � � � � � � � � � � � � � . 135
5.5. Injector vacuum preparation system . . . . . . . . . . . . . 137
5.6. Electric power supply system . . . . . . . . . . . . . . . . 137
Chapter 6. Injector Based on Negative Ions . . . . . . . . . . . . . . 142
~ 6.1. Selection of injector parameters . . . . . . . . . . . . . . 142
6.2. S~ru~tural diagrar.~ of injector . . . , . . . . . . . . . . . 144
6.3. Components of ion-atom channel . . . . . . . . . . . . . . . 148 .
6.4. Pumping sy~tem . . . . � . . . . . . . . . . . . . . . . . . 158
- 6.5. Electric powe~~ supply system . � � � � � � � � � � � � � � � 159
Bibliography . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 162
5
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CHAPTER 1
INJECTION OF FAST ATOMS, METHOD OF CREATING THERMONUCLEAR PLASMA
Injection of fast hydroqen (deuterium) atoms into magnetic systems is now re-
garded as one of the main methods of creatinq a thermonuclear plasma. This
method was born during the early 1960s when the Phoenix jl], Ogra-2 I2] and
2X [3]--open traps designed to store a plasma formed upon ionizati.on of fast
atoms enterinq an installation--appeared. The initial captures made by the
Lorentz io nization and an exponential increase of density then occurred due
to ionization upon collisions of the beam of atoms with particles of the
formed plasma. �
The injectors of that time were comparatively low-power; their rapid develop-
ment was larqely stimulated by interest in additional heating of plasma in
tokamaks. Installations of this type, developed at IAE [Institute of Atomic
Energyl imeni I. V. Kurchatov and that demonstrated good capabilities in pro-
ducing a dense hot plasma, spread rapidly throughout the laboratories of the
world. The corresponding intensive research led to conclusions on the inade-
quacy of Joule heating, ordinary for tokamaks, and the need for additional
input of enerqy, for example, by high-frequency heating or injection of power-
ful beams of~fast hydrogen atoms. The result~ of efforts ta develop injectors
meeting the needs advanced by tol~amaks are presented in Figure 1.1, which il-
lustrates the sharp increase of flows of atoms injected into installations
throughout the decade from 1969 through 1979.
These results were reflected in plasma parameters achieved in tokamaks with
injectors--a record value of 6(ratio of plasma pressure to magnetic fi.eld
pressure) in the T-11 [4] and an absolute ion temperature record in t'~e PLT
[5]. The main thinq is that the classical mechanism of energy trai:sfer from
a beam trapped in toksmak to a plasma is confirmed and it is shown that its
temperature increases in proportion to the output of the injected beam; no
additional losses relatad to the presence of a group of fast ions occur.
The appearance of powerful injection systems again aroused flagging interest
in open traps with injection since the hop~ful results obtained in the 2XIIB
[6] permit one to talk about a new quality of an old cambinatian.
~Thus, experiments of th~e last few years confirmed the practical importance of
the injection method to solve the problem af controlled thermonuclear fusion;
injectors of fast atoms will apparently be one of the main parts of the demon-
stration thermonuclear reactor of the near future.
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_ 3KD.
A 2X(19 ' (15
?O!B
~ ~ (15-20Ka6)
(f
101 (4a~36)
2xl! T-11 T-!!
(?OK38) (15K36) (15K3B)
f0~ --._-1
ORMAK
C[eo (.~OK38)
- 25?c3B)
10�
t3
Ozpa�3
(15K38)
10~~ --'~Photnix 1/ _
(10K38)
- j ozPa-z
(75K38)
1965r. 1970r, 1975r. 19dOr.
Figure 1.1. Growth af Flows of Atoms Injected Into Installations
Key: 3. Ogra
I. Equivalent amperes
2. keV
- 1.1. Types of Thermonuclear ~nstallations With Injection
As already noted above, fast atoms are injected into both open and closed mag-
netic systems. It is assumed that both methods may lead to the goal-~develop-
ment af a thermonuclear reactor, initially a demonstration reactor and then a
commercial reactor that is economically advantageous and capable of competing
with other energy sourc~s. Tokamaks are presently the most advanced represen-
tatives of closed systems toward this qo~l.
Accordinq to modern concepts, at least one of three varieties of a thermonu-
clear reactor based on a tokamak.with injection can be achieved: 1) a reac-
tor ~aith ignition, 2) a reactor with constant heating and 3) a two-component
reactor.
The first version is the most preferable. In this case the injectors should
be switched on only during the initial operating cycle. By using them, the
plasma temperature is brought up to the required temperature for "ignition"
of an intensive thermonuclear reaction, after which the injectors are switched
off, while "combustion" is self-sustained due to the energy of a-partic].es
formed as a result of the reaction it~elf. It is clear that the a-particles
must not leave the system without having releasec~ their energy to the plasma
and tr.at the rate of energy loss from the plasma to the reactor wall not ex-
ceed some maximum value to realize this attractive possibility.
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If heat transfer is so high that the reaction is not seZf-sustaining, the
reactor will be able to aperdte only provided there is continuous makeup of
the departing energy due to continuous injection of high-energy particles in-
- to the plasma, i.e., we have a system with continuous heating in this case.
It should be noted that practical realization of a thermonuclear reactor can
' be an int~rmediate version of these tw~o vari~ties: a powerful heating injec~
tion system at the beginning of the operatinq cycle brings the reactor "almost
to ignition," after which the required energy balance is maintained by a cam-
paratively weak control ir.jec~or. This scheme is convenient from the view-
point of monitoring the operating mode, since it permits control of reactor ~
"criticality" by changing the output of the injected flow.
Finally, the third possible variety--a two-component tok~mak reactor j7)--is
_ a system in which energy is praduced in reactions occurring upon collisions of
- high-energy injected particle~ with particles of a comparatively cold plasma
that serves as the target bombarded by a beam impinging on it rather than in
a thermonuclear reaction, i.e., not in the combination of nuclear reac~ions
in a plasma heated to a high temperature. A tw~o-component toka;mak, being a
powerful neutron generator, can be used specifically not only in the form of
a gurely plasma-beam system but also as a hybrid reactor in whose jacket ad-
ditional energy is produced as a result of neutron irradiation of fissionable
materials with simultaneous breeding of a nuclear fuel--plutonium.
If we can talk about a plasma heating method alternative to injection in toka-
maks--introduction of a flow of high-frequency electromagnetic radiation, then
the practically interesting idea of using an open trap without injectors
apparently does not exist at the given moment. Moreover, as already noted,
it i~ the modern level of developments in the field of injection systems that
made possible the appearance of such very promising proposals as systems ~~rith
reversal of the magnetic field an3 ambipolar traps.
The idea of reversing the magnetic field inside an open trap has existed since
the time of the astron [8] in which it was suggested that this reversal be ac-
complished by creating a conducting layer of fast electrons injected into the
trap. Realization of this old idea on a new base--due to the dimagnetism of
- a plasma created in a trap with a powerful injector--now appears quite real-
- istic. The results of experiments on the 2XIIB [6] indicate that total repul-
sion of the magnetic field from the center of a plasma has now been fully
achieved in this installation. If a magnetic field reversed to the external
field can be produced in the center of the plasma, then s qualitatively new
field configuration is achieved--the magnetic system is closed due to the di-
- magnetic current in the plasma itself' (Figure 1�2)�
An ~?mbipolar trap was suggested comparatively recently at Novosibirsk by G. I.
Dimov et al [9] and independently by Fowler and Logan and is being a.ctively
analyzed in both our country and abroad. It consists of three parts--two
small traps into which are injected high-energy particles (1 MeV) are jained
to an ordinary open trap, while medium-energy particles (1G keV) are ~njected
into the main trap. By corresponding selection of the injection currt:~ts, one
can achieve a plasma density in the outer traps nk higher t:han in the cenzral
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~ ~
~
" . I I � -~-I-
i~- ~ / i i
o~~ i
~ ~
- Figure 1.2. Reversal of Magnetic Field in a Reflected Trap and
Formation af a Closed Configuration
- trap n~ (Figure 1.3) and in this case the ambipolar electric field forms a
potential barrier in the axial direction to ions located in the central trap.
8M
Bo
~PK ~Po
nK
no
Fiqure 1.3. Distribution of Magnetic Field, Potential and Density
of Plasma Along Axis of Ambipolar Trap
1.2. Requirements on Injection Systems
The requirements placed on an injectian system during design are determined by
the type, designation and scales of the installation. Let us enumerate the
main characteristics primarily stipulated by these r~quirements: 1) the out-
put of the flow of atoms injected into the installation, 2) the total injected
flow of atoms, 3) the enerqy of the beam particles, 4) the injection pulse
length, 5) the geametric characteristics of the beam of atams, 6) the accom-
panying flow af cold gas, 7) the energy efficiency of the injection system and
8) the operating reliability and service li�e of the system (including its ef-
ficiency under canditions of neutron and gamma-radiation).
It seems at first glance that one could limit oneself only to two of the first
three beam characteristics. However, depending on the typE of installatinn
- for which the injector is being designed, pair combinations different for them
emerge to the forefront. Thus, the specific parameter in a heated toka,mak
(with ignitian or with constant heating~ is the oui~~t of the beam injected
into the installation. The particle anergy must then be selected so as to
provide the particles with sufficiently deep penetrai:ion into the plasma and
capture m~re or less uniform through the pinch cross-section. It is natural
that in this case the total ~njected flow of atoms is a deviation. If the in-
_ jector is designed for a two-component tokamak, t;hen one must first select an
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energy s~ as to achieve the maximum yield of nuclear reactions with estab-
lished energy distribution of trapped fast particles. The design intznsity
of the reaction with selected injection energy permits one to de~ermine the
total injected current. Thus, beam intensity is now a derivative.
The remaining parameters can remain unchanged with formulation of the initial
requirements on the injector.
Table 1:1. Parameters of Injection Systems of Some Existing Installations and
Those Under Construction
Y ~ $4) ~ o Y 7) ~~8 s ~9 (~10) .
A (
Y e
t ~ =(2 = (3 : gt6 R~ ~ ~ �a
YcranoeKa " R = (5 ~ o . u
~m a ~ ~C a' ~
o{o im 5 om s= o Y S' S�
< a m~ ~ F a: a R a a
PI.T 4 40 4-(G) O,G-O,~J I GO 0,3 3,G 20
D�III 10 50 IG 0,6-0,9 1 60 1,0 4,0 15X30
PDX 4-10 90 4-6 0,6-0,9 I GO 0,3 3,G 20
Tt~7'R 20 120 4-(6) 5 3 70 0,5 8,5 9~X7U
T-IOM 4,5 40/802-(4) 2,5 3 35 I,5 6 20X80
Key :
1. Installation 6. Number of sources in injector
2. Injection pawer Pp, MW 7. Ion current fran source I+, A
3. Ene rgy of atoms E~ , keV 8. Pulse length , s ~
4. Number of injectors 9. Leng~h of injector channel,
5. Injector pawer, MW meters
10. Dimensions of inlet aperture, cm
Injection systems capable of injecting flaas of hydrogen or deuterium atoans
with power of 10-20 taW with particle energy of 40-120 keV and injection pulse
length up to several seconds are required for research installations of the
next few yeaxs. Some parameters of these systems in iustallations already
operating and those being desiqned are presented in Table 1.1 [10]. The de-
vices of the next generation, whicY: will actually be demonstration reactors,
will be designed for outputs of injected fluxes on the order of hundreds of
megawatts. The efficiency of the injection systezns will be of determining
. significance for these installations.
1.3. Design Principles of Injection Systems
The physical prinicple of producinq energetic beams of atoms, used in modern
injectors, includes acceleration of ions with subsequent conversion of them to
atoms. This principle can be realized by two schemes--through positive or
negative ions. Specific selection of one or another scheme is determined by
the desire to produce the maximum energy efficiency of the injector. Z'he fact
is th at it is easier to generate positive hydrogen ions than negative ions,
but the charge exchange cross section of positive ions on t:he target, which
10
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determines the coefficient of converting them to atoms, decreases strongly
with an increase of energy (Fiqure 1.4). At the same time, the "stripping"
cross-section of negative ions that converts them to atoms is weakly dependent
~ an energy up to very large values (Fiqure 1.5). Therefore, it is much more
advantageous'with respect to energy to use positive ions to produce cbcnpara-
tively law-enerqy fluxes of atoaas and negative ions if high-energy atanic
fluxes are required. The dependence of the enerqy efficiency of injeators
designed according to the two schemes discussed above are presented iz, Figure
1.6.
Go,
cn=
r6
10'�
/0 ~e
p t0o 100
E, keV
~'igure 1.4. Dependence of Resonance Charge Exchange Cross-Section
of Fast Hydrogen Ions to Atoms on Energy
z o:
~v , ,
z
3
0
r^� N=
I
N
J IO
b
~O
~ ~ H
z
10
` !
=a.
Il
0 S 25 5o E,ke V
Figure 1.5. Dependence of "Stripping" Cross-Section of Fast-Negative
Hydroqen Ions on Hydrogen, Nitrogen and Oxyqen Molecules:
1--quasi-classical approximatian; II--B arn approxima�
tion; 1-3--experi.mental data from [11]
11
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Technic~l embodiment of one or another physical scheme is always related both
, t~ funda~mental limits imposed by the laws of ~n~,t�ure anc3 to real techno-
logical aapabilities. For example, the restrictions of the typical dimen-
sions of the ion-optical system (IOS) of the ion source--the most important
component of the injector--are directly related to the maximum achievable pre-
cision of manufacturing its parts on metal-cutting lathes, to the mechanical
strength of the mat~rial and to the capabilities of heat dissipation. Thus,
the size of the accelerating gap is detesmined by its high-voltage strength,
which is very strongly dependent on the Suality of machining the electrode
surfaces. These restrictions in turn limit the ion current density from the
source at which the beam can be formed by the ion-optical system in a satis-
factory manner. An ion-emission surface must be developed, the number and
dimensions of the inr ut apertures in the installations must be increased and
_ so on to produce large flu~ees of fast atoms.
K n. B ~ ~
0,8 ,
3
0,6 -
0,4
0,2 ~ Z
0 100 f00 600 B00 1000 f,
keV ~
Figure 1.6. Depend~ance of Injector Effii;iency on Energy E~ for
DeuteY~ium: 1--injector based on direct charge ex-
change of positive io~ns; 2--the same injector with
recovery (85 percent) of ion energy; 3--injector
using negative ions with double charge exchange on
a sodium target
Key :
1. Efficiency 2. keV
~lnalysis of similar relationships and also the requirements of operating reli-
ability and operating convenience lead to fonnulation of the follawing struc-
tural principle of designing injection systems for thermonuclear installa-
tions: the injection system should consist of several inject~;.on zones, each
of which combines several injectors having self-contained vacuun: systems;
several ion sources with self-contained electTic pawer supply circuits and
a working gas supply are installed in each injector. This situation is actu-
- ally the modular principle that perntits independent optimization of the paran?-
eters of separate companents and that increases the operating reliability of
the entire system as a whole, since a failed or worn-out component can be re-
paired or replaced without interrupting the functioning of all the remaining
camponen ts.
The sodium charge exchange target (Figure 6.7) consists of a steam generator
1 with supersonic nozzle.
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1? J 4 S 6 7 e 9!0 l? 4 ln J 4
, . '
4
~ I I
~ I~
~ ~
, �
I I
' _l6 ' ~ I i
~ ' ~ 1J
15 ~
/4 �
Figure 2.8. Experimental MIN Injector: 1--positive ion source; 2--
insulator; 3--ion beam; 4--vacuum chamber; 5--evacuation
pipe with slide valve; 6--vacuum chan~ber of sodium target;
7--steam generator; 8--supersonic jet of sodium vapors;
9--condenser; 10-~-insulator; 11--preacceleration system;
12--diagnostic detector; 13--vacuum pumps; 14--platform
urider ground potential; 15--platform under potential of
100 kV; 16--vacuian slid~e valves
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~
~ ~
~ ili
I I .
' ~
1 M ~
ti~. .
2
i �
i
~ ~1~.,` �
I - -
. , . . 'I' -I,
I , 1
~ ~ ~y
.I
I .
~5
' I .
6' I
, i
I
Figure 6.7. Design of Continuous C1large Exchange Tar.get: 1--steam
generator; 2--separator; 3--windaas for beam travel;
4--interme diate condenser; 5--condenser; 6--electro-
maqnetic pump
COPYRIGHT: Energoizdat, 1981
- 6521
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uDC 62i.o39.516.22
USE OF BURNING ABSORBERS IN RBMK
Moscow ATOMIVAYA ENERGIYA in Russian Vol 51, No 5, Nov 81
- (manuscript received 13 Feb 81) pp 304-307
[Article by V. K. Vikulov, S. A. Dorshakov, V. I. Ilyukhin and Yu. I. Mityayev]
[Text] The intensive development of nuclear power is making even sma11 improve-
me nts in power reactor characteristics significant. One such improvement in the
RBNffC [1] f~zel cycle is enhancing uranium enrichment. This enables a significant
increase in its burnup, which in turn noticeably lowers the need for natural
- uranium aizd reduces the fuel component of electric power production cost. How-
ever, it is not always possible to enhance uranium enrichment. In the contin-
uous fuel recharging mode, and when bringing the xeactor to this mode, the
fuel channels of the core have a different power output (burnup), from zero
(fresh) to that of the spent ones. Therefore, the content of fi.ssile isotopes
in these channels is unequal, causing an additional power variation factor due
to overload [2]. EnhanQing the enrichment raises the overload factor, which
can lead to unevenness of power distribution that is unacceptable in terms of
- the cooling conditions. Such a constraint is especially significant for
re actors with high specific f1ie1 power, such as the RBMK-1500 and the RBNdCP.
One possible method of overcoming such a limitation is to use a self-shielded
burning absorber inserted as an absorbing rod in the fresh ~1Ze1 assembly during
loading into the reactor. Implementing this method is aided by the fact that
the design of individual RBNIIC fuel assemblies provides a centrel hole 6.5 mm [3]
in diameter.
Usually, burning absorbers [4-7]are used to lower the reactivity, compensated
by mobile control elements, where its burnup reserve is large. In such a case,
the reactor operates in the continuaus recharging mode, and the compensating
capacity of the regulation elements is known to be sufficient. The capacity
for local balancing of the power distribution is limited, however, since each
of them affects a rather large region of th.e neutron field that includes both
the freshly loaded and burned up f1ie1 channels, and the effect of the control
elements on fuel channel power is insufficiently differenti~.ted. Burning
absorbers are thus used in this case only for local balancing of the power
, distribution.
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The authors were faced with two tasks: to determine the density of the absoroing
material'and diameter of the absorbing rods inserted in the center of the fuel
assembly which will provide maximum reduction in the power of the loaded fuel
channels with the least total nEUtron absorption in the absorber during the
time of operation; and, given the existing power engineering constraints on
the maximum fuel channel power, to find tho~e conditions of channel utiliza-
tion allowing enhancement of enrichment and burnup of uranium, resulting in
improved economic indices of the fuel cycle.
The theoretical approach to solving the first problem, that of choosing the
optimal burning absorber witi~ the best coordination of its burnup with that
of the fuel, is the same as that given in works [6,7]; however, the method of
solution is different. It is based on using the VOR [2] computer program,
enabling calculation of the spatial distribution of the thermal neutrons of
the fuel channel cell in a P3 approximation with part of the calculation grid
in the region of the abs~rbing rods (such that the distance between the calcu-
lation nodes is much less than the neutron path length). It thus becomes
possible to directly calculate the "~arning" [7] of the rods using a small
time step, with simultaneous burnup of the uranium, and correspondingly allow
_ for the change in the neutron-physical characteristics of the fuel channel
~
. cells with the run . .
The solution of the second problem, requiring a heterogenous design of the
reactor, used a two-dimensional diff~sion grid HODIHER program [8], whose
two-group cell macrosections are calculated in the first part.
The experiments showed that when absorbing rods 5.6 mm in diameter are
inserted in the central flzel assembly hole, the power of the fresh fuel
channels declines by 10-15~. In both experiments and ca'_culations the
absorber was gadolinium oxide GIYZOg, having two strongly absorbing isotopes
155Gd and 157~~ If an identi~.al uranium and e,bsorber burnup rate is
assumed, then the reactivi.ty and power of the fresh fuel channels and the
absorbing rods will remain consistent until the absorber burns up. The power
distribution variation factor in the reactor will ~hus drop by the same
10-15~. In practice, it is impossible to ma.intain an identical uranium and
absorber burnup rate during the entire burning time of the absorbing rods.
- However, it follows from the results ox'raork [7] and this research, that the
~ discrepancy in the burnup rates can be reduced to a minimum given a certain
gadolinium concentration. This is canfirmed by graphs of Fig. l, illustrating
the varying effect of the self-shielded burning absorbers on t he RBMK fuel
channel reactivity i,n the process of simultaneous burning of uranium and
gadolium. ih~anium enrichment in the fuel channels is 2.6~. Given a low
* The authors first used such a calculation method in 1966, in designing the
- first loading of the reactor of the second stage of the BAES power plant.
Experiments during the physical start-up confirmed the basic calculation
results.
~6
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absorber concentration, the burnup of ~he rods i~ faster than that of the ura-
nium, and a fast growth in reactivity, and thus pawer (curve 2), is observed.
On the other hand, given a hig~? concentration of gadolinium, the rods burn too
slowly (curve and although the maximum ft~el channel power Nm~ declines (it
is defined by the initial point~, the reactivit,y loss is too great, which
un,~ustifiably reduces the uranium burnup. There is obviously an absorber con~
centration at which ~he reactivity loss will be minimal (curve 3) given the same
_ action on Nmax� ~lculations show that for F{BN~{ fla.el channels the opti.mum
absorber concentration corresponds to a GdZ03 density of 0.~+-0.5 g/cm3, and is
almost independent of the rod diameter if it changes in the range of 3-12 mm.
The initica..l. influence of the rods on the flzel channel reactivity and power is
basically determined by their diameter, and increases with it.
Under conditions of continuous recharging of the fuel channels and burning ab-
sorbers, the diameter of the rods with an optimal absorber concentration unam-
biguously determines such mutually connected characteristics as Nm~, rod burnup
time t~ and uranium burnup time P. The greater the diameter, the..lower Nmax and
P and the greater t~. Spatial calculations of the reactor have been performed
to establish the relation between such characteristics. Fig. 2 gives curves for
determining the ~ossibility of ldwerin~ Nmax and the relative underburning of
uranium when using rods of different diameter in fuel channels with 2.6~ enrich-
ment. The calculations were performed according to an iteration scheme worked
out by the authors. Assuming a continuous recharging mode, it allows for
_ underburning of the spent uranium given the use of absorbing rods. The power of
the fuel channels containing the rods is assumed equal to Nmax~ and does not
change until their complete burnup. This enables determination of the fuel
- channel run, t~, and thus the proportion of the fuel channels containing the
absorbing rods; i.e., formulation of the reactor composite lattice whose calcu-
lation is used to find Nmax�
The calculations done with allowance for the different diameter of the rods and
the enrichment of the recharged uranium showed the possibility of a maximum ~
increase in its enrichment, given a predetermined constraint on the maximum fuel
channel power depending on the diameter. It is more convenient to use the
power distribution variation factor KE or overload factor Kov than Nm~ as the
constraint. It can be seen from Fig. 3 that despite a certain loss of uranium
burnup due to the additional neutron absorption resulting from the use of rods,
its enrichment and burnup increase noticeably. Naturally, less uranium burns
up in this case than in the absence of the constraint (curves 1, 2).
These results describe only the technical aspect of enhancing uranium enrich-
ment and burnup, and are still insufficient for a clear conclusion on the
economic advisability of using absorbing rods. Data are needed on the eco-
nomic and natural indices of the fuel cycle. To provide them, the calcula-
tions of the relative change in fuel components of electric power production
cost and ad,justed costs for producing electric power, and tbe uranium f~.iel
17
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a~ate required for annual make-up of th~~ reactor, have been done as a flzriction
= of the rod diameter. The calculation results show (Fig. that in the RBMK-
1500 and RBN~tP it is advisab~e to use absortirig rods up to 10 mm in diameter.
The greatest technical and economic effect is achieved using rods 1+-6 mm in
diam.eter, which is very important from the p.ractical standpoint, since it
- enables their installation into currently manufactured fuel assemblies without
any design modifications. The effect is greater the more rigid the constraint
- on power distribution variation. For exam.ple, at KoV = 1.27, the fuel compo=
nents of power production cost and ad~usted costs for electric power production
can be reduced by 7-8 and around 2~, respectively; 2-2.5~ less natural uranium
is consumed.
FY.nally, it is interesting tc~ point out certain results of a calculated study
of the effect of absorbing rods on the space-energy distribution of thermal
neutrons in the fuel channel cell. This effect, studied using the NEKTAR [9~
progra.m, turns out to be negligible (cf. table), which makes it possible to
simplify the rod burnup calculations and preparation of the two group constants.
In particular, it turns out to be possible to use the single-rate VOR progra.m,
to whose sections are added the necessary thermalization corrections. Deforma-
tion of the neutron distribution due to the use of rods can cause a relative
' increase in the power of the fuel elements of the outer row of the fuel assem-
bly; however at a 5-6 mm diameter it does not exceed 2~.
The results given are also valid for the use of other self-shielded thermal
neutron absorbers, for example those containing 113Cd or 1~9Sm.
I
~
~
_ ^
' I.
i
P MWt. day/kg
/1 .f P,M�v~ C;r;rt ~
Fig. l. Dependence of the neutron breeding factor on an infinite lattice in
an RBMK fuel channel K~ on uranium burnup P at a Gd203 density of 0(1),
0.2 (2), ~.45 (3) and 1 g/cm3 (4).
18
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.
Ke K, r, %
~4 ~
9B
l,J /
96
1,? -
p Z y 6 B d,MM
Fig. 2. Effect of the absorbing rod diameter on uranium burnup (1) and
pow~er d~~ ":ribution variation factor
~ JM~a .
30 I - ~ / 16
~ - /
/
~ . ~ ? / 1?
~
a3 " / /
~ ~ ~6 - B
~
_ py J 4
~ i
a' pp ~L ~ L-L_ 0
l,o z, z~ 4 ; c.~, io
FYg. 3. Dependence of spent uranium burnup on i$s enrichment x in the absence
(1) and presence (2) of a constrai.nt on Nm~, ensured by the corresponding
diameter of the absorbing rods (3). Kov = 1�27 and 1.3, respectively
0 2 a 6 e e,MM ~
~ r~
z
d G, 2 ---a
_ \ -
~ ~
-2. - \
, \ ~ \
_y _ 2
-6
-B -
nr~. ~ : _ _ . A3�~b
Fig. 4. Effect of the absorbing rod diameter on the relative cha.nge in fuel
- cycle indices: a- natural uranium consumption G; b- Puel component of electric
power prodt~ction cost CT (1) and ad~usted costs ZT (2). KoV = 1.27 and 1.3,
respectively
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TABLE
Distribution of thermal neutron fluxes F* (relative units) and neutron gas
temperature Tn~g (K) in a fuel channel cell as a function of absorbing rod
diameter (density of Cd203 0.~+5 g/cm3).
~ 7'oania b~BWU~ C'1'~~yGii ~uuiana d~ 1'p~upuT
d,
HM
I ~x.r. ~J' I Tu.r Q' I 7u.r d' /n.~.
0 I o,5181 848 ( O,G68I 7G2 I 0,8cw I 811 I l,Q:t~J I 8~i7
IJ SJO Slil
~ I U,43A I$SG I U,G5'l l 7G7 I U,800 81'Z 1,U~i1 $~i8
U 58G 8G'1.
~ IU,47l1 RI;I) IO,6~i41 768 I0,70JI 813 (9,11/,2 &4J
U,58U SG3
f0.1 U,4~i~Jl 8G2 I O,G33I 770 I0,7~JG I 81~i I i,0/i3 8/i!)
* ~t
a. Fuel elements b. Water ~ c. Channel pipe d. Graphite
* The average values by volume of the indicated elements are given Por
fluxes normalized to the average value by cell volume.
The data in the numerator are for the 12 outer fuel elements; those in
the denominator, for the 6 inner fuel elements.
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REFERENCES
1. Yemel'yanov, I. Ya. et al., ATO:~II~IAYA ENERGIYA, Vol 46, No 3, 1979, p 139.
2. Dollezhal', N. A., Yemel'yanov, I. Ya., "Kanal'nyy yadernyy energeticheskiy
reaktor" [Channel Power Reactor], Moscow, Atomizdat, 1980.
3. Bulkin, Yu. M. et al., in: "Opyt ekspluatatsii i puti dal'neyshego
razvitiya atomn~y energetiki" [Operational Experience and Paths of Future
Developmen~ in Nuclear PowerJ, Obninsk, izd. FEI, 1974, Vol 2, p 28.
4. Radkovskiy, A., in: "Trudy vtoroy Zhenevskoy konfarentsii. Izbrannyye
doklady inostranykh uchenykh" [Transactions of Second Geneva Conference.
Selected Papers of Foreign Scientists], Vol 3, Moscow, Atomizdat, 1959,
p 717 .
" 5. Volkov, V. S. et al., ATOMNAYA ENERGIYA, Vol 11, No 2, 1961, p 109.
6. Toshinskiy, G. I., Kalashnikov, A. G., in: "Teoriya i metody rascheta
yadernykh reaktorov" (Theory and Methods of Calculating Nuclear Reactors],
_ Moscow, Gosatomizdat, 1962, p 118.
7. Orlov, V. V. et al., "Tret'ya Zhenevskaya konferentsiya, 1964, doklady
SSSR" [Third Geneva Conference, 1964, Soviet Papers], No 354.
8. Gorodkov, S. S., "Instructions on Using the HODIHER Program for Hetero-
geneous Reactor Design", Institute of Atomic Energy Preprint No 2578,,
Moscow, 1975.
9. Vikulov, V. K. et al., "Voprosy atomnoy nauki i tekhniki. Seriya Yadernaya
~izika nizkikh i srednikh energiy, reaktornaya fizika" [Probl^ms of Nuclear
Science and Engineering. Series on Nuclear Physics at Low and Inte:rmediate
Energies, Reactor Physics~, No 5, 1977, p 57.
COPYRIGHT: Energoizdat, "Atomnaya energiya", 1981
9875
CSO: 8144/1002 .
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. irnC 621.039.59
CIiOICE OF EXTRACTORS AND FILTERS FOR PLANTS REPROCESSING SPENT FUEL
ELEMENTS OF ATOMI C POWER STATIONS
Moscow ATOMNAYA ENERGIYA in Russian Vol 51, No 5, Nov 81
(manuscript received 8 Dec 80) pp 317-320
[Article by A. M. Nudel' , A. P. Roshchin and V. V. Dolgov]
[Text] The chemica,l conditions of processing spent nuclear power plant
effluents have specific equipment requirements, such as nuclear safety, remote
control servicing and maintenance, possibility of decontamination, air-
tightness, a minimum of radioactive wastes, etc. This complicates the design
of the equipment and increases its cost and operationa~. expenditures. Con-
sequently, equipment design significantly influences the effectiveness of the
radio-chemical plants being examined here. In principle, the optlmal construc-
tion or type of equipment for given conditions may be selected on the basis of
technical and economic calculations. However, the necessity to take into con-
sideration complex and contradictory requirements, and the absence of reliable
mathematical models of the process makes it necessary to introduce varying and
- often sub~ective hypotheses that lower the reliability of such calculations.
Even with adequate information on the process as a whole, the reliability of
the comparative evaluation of equipment may be improved through application of
empirical methodology. For example, a system of criteria for selecting
extractors was suggested [1, 2]: value intervals for each criterion selected
in advance are evaluated by a speciPic number of points and the sum of points
received by each compared piece of equipment determines its advantages.
The basic inadequacy of present methodologies is that their authors attempt to
derive an absolute evaluation of equipment that does not depend on what it is
being compared with; consequently, after a single comparison, the criteria
tables themselves become unnecessary since the evaluation remains unchanged.
In addition, it is impossible to suggest a system that would permit an intui-
tive determination of the selected interval values of the criteria and the num-
ber of points they were assigned. One more substantial inadequacy is that it
is impossible to determine the expediency of utilizing different types of
equipment in a single teclinological chain. For example, the ability of an
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extractor to handle unclarified solutions is evaluated by the highest mark 5,
while inability to do so receives no mark at all [2]. However an arrangement
including a filter and high-efficiency extractor may turn out to be more
advantageous than a less effective extractor without solution clarification,,
~ It~s also impossible to compare versions in which different types of equipment
- are used in the extraction and reextraction stage.
The following principles serve as the basis of the suggested methodology:
equipment for a specific technological process is compared; the comparison of
equipment assures that required technological parameters will be met and their
operational characteristics learned and verified. Also taken into considera-
tion is its connection to related equipment in the technological chain. The
final evaluation of the equipment is not absolute but its superiority or
+ inferiority is determined only in relation to other equipment being compared.
The methodology has been developed to compare extr.a.ctors and filters--the basic
tools utilized in modern radio-chemical reprocessing of atomic power plant fuel
elements. Other equipment ma~r be compared similar~y.
Basic criteria influencing effectiveness (Tables 1, 2) were selected in advance.
Selection was based on known and generally accepted evaluations recommended
for technical/economic calculations. Each compared piece of equipment :;.s char-
acterized by an absolute value on the basis of any criterion but its link with
related equipment in the tachnological chain is characterized by criteria that
take into consideration the value of auxilliary equipment for transmitting the
effluents and operating expenditures. Criteria are ranged by their value and
rated by coefficient r. Evaluation of the equipment's we,ter phase B according
to any criterion is determined by taking into consideration absolute values of
_ the latter with regard to the instruments being compared
~i~~-_ -
~~~r Ni - ~'t,
mB.X
where NP.- is the value of the criterion for the instrument being examined;
N1 - is the highest criterion among the instruments being campared; p- the
n~imber of instrument being examined; i- is the ordinal criterion number.
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Table 1
Filter Evaluation Critexia
huurc~mrl I ~i I Rt � tu-
~riterion
1 G(~eRusui c~cr~pocr~, ~)~u:tirr~~on~mui � 3,~~G 1,$'~
2 ~luc:io ~j~u:iirrpou, ~icofi~o;~~n+oc unf~ ~i,S7 3,Gt
- OUCCdIC9L'i(IUI tronpcpianuuii pn6o'ri,~
~rllcneic~ui ~~ni:u~r~inu;iuNn c ~~~111u-
Itnlt IIj10I1~ilUYlllTl':ILI(UCTLlO
3 06~,ent cGpocn~~x pacT~sopon n nyn~,- ~i,35 3,97
ni.t n~~o~ii,iTnro orinxa, uia~~ainsic-
�i~~x tt~ iicPcl~aGortcy o~�~nnou "'r
4 Ilonrrxttor,�ri, il~ii~n,rpytun~cii tteporo- !'i,:i7 3,~J3
i,n;l~cu, �rpc6yinn~cii a~~iem~t, n p~c-
~c~TC r?~~ ~ Ti,ia ~t T~~Gu~rt,< i ncnn�or~?reni~niax p;~c:i�uc~- 5,'~!1 ~i, l
rnlt s+s
6(~fi~~ca~ cfi~,ocinax plcTnopnu n i~y~n,- (i,i~J ~i,8
nw oai;U:;i, uwl~lnacati,tx u~ uepc-
~,~Gorx~~ nepoll nox,i~uia:?itncii
7 1'r~nom~~r ~a�r~~~n~~ na :iucnnynmruuuu 7,3~i 5,~i~i
ocaton~torn ii UCIIQ~tOCil7'CJILU~~ro
~~Cin~>}�Itooannn
8 CTOn~iocri~ i~~~~:~i.�rpa (cyni~~upnaii 7,4G 5,53
cron>inc~�i. ~~a+:ii.rpon nrnrncriiin)
9 13iacn�r~i ii~uii:iuol(crnr,irni~ix na~ienU!- J,~7 7,:~2
n n ii
ip 1IJrou1a;it. :iaun,r.icaii~ix IljtAl~31t0J4C.T- 5~,~)li 7,38
nenn~,~~ n~,~,icnteunii
11 CTOII~IOG7'L ncno~wrarr.ni~u~rc~ uGory- ~ I,O 8, tS
;lona~mn (n:ir,nci,t ,R~ui unrotri~ntn
par,ruo~~nu n c~~cucnanii uny'rpu
' or:tencu nti)
; 12 :{:ITI78TL~ n,~ i~o;~rornutc,y p~~cruopon t t,;i3 3,~i
,t c~i~:ii~�r~�~u~uino (~J~noicr.yntntTia,
ucuo~mra�rc:ii.uiae ~~uini~rpyiun(ne
nr,ntecTna) �s'
13 It~:~.i~~icci~n~~ noa7l)'s.i, cfip,~c~ru~~c~~orn 11,74 8~7
14 ~~a or:tc:ic~inn na u~~ucrxy
C�ruis�~c~ri, cnri�c~iia lUIf(nA 11,7t1 $,7/i
_ ~ 15 C�r~n,inr�ri, c>ucocTU~ii ;inti~ip;rrypi,t, tt,~il $,7(i
ttc~~ixi,ll,,�~c,ii ;~n,t n~~p>iant,nu;i p~~-
~ioTia cj,u:ii~~r~~oii
~ 16 Cr~zt>tnr.Ti, xnurciincpon ~i n(~ucnu- 1~~~~~~i S),31
co6nannii n:~j~ ;luc'ram~uomiuii aa-
>icni,r ;ic~j~cicrni~s yannn n ;~~��r,uicii
~
� ll~~n p:~c~ir.ro n�niocurr~u.nnlf utlours~~ u~~urr.~~un ~~cnuni,-
:r;~rcn u~~nuamia. fifj~~8T11~11 r.rr~ a~ir,mm~�rnoaiy :oiavrnum; *:k nn
1~r~ ~picn.T~~MB; *a+ ua f ei nCxu~~nufl cycnrn:um.
Key follows on
_ next page
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Key to Table 1
i
1. Average filtration rate*
- 2. Number of filters necessary to assure filtration department's continuous
handling of assigned production
3. Volume of effluent solutions and washed sediment pulp provided to waste
processing .
4. F`iltering partition surface requiring replacement calculated for eight
thousand hours of operation per filter
5. Volume of auxilliary solutions*~*
6. Volume of effluent solutions and pulp sediment provided to reproces~ing '
prior to localization~*
7. Annual expenditures for operational and au3cil.liary equipment
8. Filter cost (total cost of filters for department)
9. Height of production premises
10. Tota1 area used by production premises
11. Cost of auxilliary equipment (pumps to transPer solutions and suspensions
inside the department)
12. Expenditures for preparation of filtration solutions (flocculants,
auxilliary filtering substances)~*~
13. Amount of air discharged for cleaning by department
14. Cost of the control and measurement instrumentation and automation system.
15. Cost of storage equipment necessary for normal filter operation
16. Cost of containers and equipment for remote substitution of defective
_ units and parts
* When calculating the relative evaluation of the criterior,, a value inverse
to its absolute value is utilized
Per 1 m3 of filtrate
***For 1 m of initial suspension
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- Table 2
Extr.actor Evalu&tion Criteria
1(pt+rrpntl ( ~i ( ~tf'
1 1'ec~~pc paGoTi,~ ~im~ps~Ta nu riunc- 5,7fi ` 2.~i~J
�ruii.n~.~ro pc?u~n�rn �
2 Itjx;~in n~~efi,rn;unui ~i~:~a o:utnapuru li,OA '?,(i:j
3 Aiex,wn~iccr.nii yuoc opruim~~echoii fi,t'l 2,fi5
~~r,i:na npn uu~~nn,~!ii.nnii upouauo-
!tn reni~iioc~�n
4 'liicno nuncip~rrnu, ne~Gso~~uai~ix Il~ur 7,7~1 :i,3i
- ~:i~~au~wii upnn:iou~~u�raii~nocrn
5 Ilpano~i~~cnTeni~nuc~ri~ ~10GOTW ~nnn- 8,21 3,55
p1r;i no a~rn~cTr�u npn ne~xpa6oT- .
xo p~~cTnupon, collep~r.auu~~ Tuo1}
ny~o ~~~~y �
6 A[mina~s~ni,uo rirnycTUni;?it p~ianocr~_ 3,43 3,(i5
nnnTi~uc~rn i~~a:~
7 t3pc~iri utitxolta ainian:~ra ~u? cT,titn~~- l8 3,!)7
i ~ Iiil(~IlLlll POSKI~~1 1fOCJ10 I)f�T81f001C11
Ilil ~1 11 60nC0 �
g ~Incnu Tettyiqne n un~nouo-n~~eAY- D,~i�"~ 4, l
ripc)u~rc~u,nw~ Pc~~nnrou nnnapaTa
(.?tumpcirou) n pacaoTO n:i 8 Twc.
p;?f ot ia
9 ~~JIIITI`1lI~lI0CT1~ ~~l`~II)I17'il ilI1I171~7:ITi1 11 ~~7J ~i,'!'~
- cnyv,ic ;iuapniitrnro nw~u~~:~ ~~a
c~~p~~~i oru ucn~~ninrx y:uiui+
10 H~uwvecTnu ~~pr:itni~t~~cxoii ~jr,iaia i~ 1q,''G ~i,4~i
:~uno~~.rru (.~iniap,rr:~x)
11 li~uiii~wc�rm~ cnc~r~~ru u~~a~lY~a, uc~r,rv- ~11,37 ~i,!1'l
n~uuu~cr~~ n;i nqllcTlty upn pa6orn .
a~nui~~ira (anu;~p;~'ron)
12 111iuniaiuni~an nonyci�n~icic urnninumio 11,38 ~i.')3
noTnicnu arr~inn~iccu~iii u uu~~iioii
i)r,ia
~ 13 llliuiu>iani~nu Aun~~cTiir~oo oruau~ein~e 11,~J8 5,19
noror.on uo~~nni~ n npr~nii~~ccicc~ii
~1~;~:,
' 14 :S:IT~187'bi (r~~ltom,~i;) n;~ :~xcnnvaiT:u~mo 1'L,2'~ 5,'l~l
r~cnmuii,i e n ItCIlu~lpphTCJIbi11dC
~IIII:I~GIT~~II, GIIC'~Ctlbl ri11Tt1>t7~T~NiOC-
t:nTr~ )'ttp;ituiciuin n ~u:r~'nn~,nu;uinf~
1 S (1liuina~a:it~n~i nec~Gsn;Uiai,in np~~nauull- I:,',4~i CI,Gr
c�ruciinan iinnnl~~;1~~ Jl~ui ~~;~,~nien~e-
- utui ami~par~i (:nniap~rr~~i~)
16 Ilinirn~~a~i~~n:isi m~tcn�r;i n(~on:~nuJlcr- 1:t,/,li ~i,8:i
pril~UdS IIUAt0111~'~init 11~~f1 P~~~~~~1t~C-
mui auuapura (~it~n:ip~rrnu)
17 C�ron~~oc~�i~ aini~~~,a7,i (:iini;i~~a�run) 1:4,S~i G,(1
18 CTnii+tncri~ nr,nua~nra�nvii~uurn ~~Gupy- 1~i,33 li,'~
noo;unu~, 1(OIIT;U(7'll(1\'Iflil(OP(1 c pn-
lUiuauTnumae?ii ~~,ic'rnnpsi~ut (tiynr~-
19 c,~rnpw, n;~coci,i n~r. n.)
CTUnau~cri~ cnc�rrai~,i au7�uariTn~icacur~~ 1''i,23 G,S!1
pory~iupou~iniu? n ynpnnnonnJr p~- ~
fioTUii ?niu~p,rrn (nimaparon)
20 C�r~~n~~ocTi~ iton�miiucpon n 1lPYi�~~~ i(i,G3 7.'.'.
~~~?~~~~~~~~~,r�~~~~~~~~~ ,,,~1~ ,u~~T~~~~~u,~>~~-
~ior~~ ~~om~~uic~~ u ;~c�onrusua aiiua-
~~a7�a, ~ro ~~anon u ucno+ioraTenu-
21 trorn ~6op~~ltonnimn
Key follows P~r,~o:~ :~ncttrpo:~nc~~rim rr~ii pnGoTe 1(i,Gf, 7,'l1
aun~~pnT~ (annapar~~u)
on next page
~ 1lpu PAC~tcrc ntnocurrni~nnll nUe~~i~~~ 1;pnrepnn Ncnna~.-
- ~ycrcn urmi~iunn, c~bp,irnnn cr~ nacom~rrnu~iy aunvenmu.
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Key to Table 2
l. Operative life of equipment prior to overhaul
2. Time that phases remain in equipment
3. Mechanical entrainment of organic phase at nominal capacity
4. Number of devices n~ecessary for assigned production scale
5. Length of equipment operation prior to stripping when reprocessing
- solutions with solid ph:.~,se content.
6. Minimally permissible difference in phase density*
7. Time for equipment to reach steac~y operation after shutdown of 12 hours
or more.
8. Number of routine and preventive maintenance service calls of equipment
(devices), calculated for eight thousand hours of operation.
9. I?uration of equipment repair in the instance of accidental failure of its
basic units.
10. Organic phase quantity in the equipment (devices).
11. Amount of condensed air released for cleaning when equipment is in
operation.
12. Minimum permissible ratio of organic and water phase flows.
13. Minimum permissible ratio of water and organic phase flows.
14. Annual expenditures for operating basic and auxilliary equipment, and
automatic operation and control systems.
15. Minimum production area necessary to set up equipment (devices).
- 16. Minimum necessary .height of production premises to set up equipment
(devices)
17. Cost of equipment (devices)
_ ~When calculating the relative criterion evaluation, a value inverse to its
absolute value is utilized.
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Key to Tsble 1 (continued)
- 18. Cost of auxilliaxy equipment coming in contact with radioactive solutions
(pulsators, pumps, etc).
19. Cost of automatic aperation and control system of the equipment (devices).
20. Cost of cont ainers and other devices for remote assembly and disassembly
of the equipment, its units and auxilliary equipment. ~
21. Use of electric energy when the equipment is in operation.
Values of Bpi are totaled utilizing all criteria. Since criteria are formulated
in a manner where the least value corresponds to the best piece of equipment,
and the more important the criterion, the lower the value of the coefficient,
it follows that equipment receiving the least sum of evaluations is given
preference over all the others. For example, two pieces of equipment are com-
pared where the absolute criterion value with significance coePficient 0.2 is
equal to 10 and 8. Then Bli = 8: 10 � 0.2 = 0.16, and B2i = 10 : 10 � 0.2 =
0.2 (to facilitate calculation, each evaluation based on a11 of the criteria
may be multiplied by a constant greater than unity, which will not affect the
results). ~
The suggested methodology is more flexible than those presently known, as it
permits the rel~stive advantages c� a�~'i^_es being compared to be evaluated. In
addition, criteria intervals and their evaluations are not included here since
they would be impossible to substantiate.
Obviously, comparative evaluation depends first of all on correct determination
� of the significance coefficient ri, which is impcssible to determine by strict
~ calculation. For this reason, a professional pol?. of experts working in this
area was conducted. Experience, intuition, and understanding of the essence
of the problem permitted them to evaluate the significance of individual
criteria, and they were able to decrease uncertainty and sub~ectivity by
assessing the evaluations with known methods. The experts were given ques-
tionnaires containing the criteria shown in Tables 1 and 2. The arrangement
of criteria in the questionnaire was arbitr.ary. Each expert was required to
assign a number to each criterion, beginning with the most significant (in
his view). It was also permitted to assign the same number to any number of
criteria. N~~nbers did noi; have to be consecutive, however they were not to
exceed the ~:otal number of criteria. It was also permitted to delete those
criteria which the expert considered unnecessary and to add those which in his
opinion had been omitted fram consideration. In addition, Y:e had to evaluate
his own competence according to a scale: am completely familiar with planning
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issues of the divisions and with development and operation of the equipment
(v = 3, the "weight" of the expert); am well ecquainted with planning issues
of divisions and development of equipment (v = 2); am acquainted with the
operation of the equipment (v = 2); am acquainted in general with planning and
operation issues (v = 1). In response to the questionnaire, 35 were received
concerning filters and 26 concerning extractors. Professional positions held
by the experts ranged from engineers to directors, with more than 10 years of
on-the-~ob experience. :They were employed by design organizations, scientific
research institutes, and plants. The ma~ority of experts rated their own c,~m-
petence at v = 2.
The questionnaires were handled in the following manner ~3, k]: criteria were
set up in order of signifi cance, in relation to the number of ranges deter-
mined by the expert (with the most signiPicant listed first). When some of the
criteria received the same evaluation, they were then marked with identical
standardized ranges determined by the eguation
k "i"(k ~ I ) I . . . ~ (!c ~ ~
- - I ,
where k, (k + 1),...,~k + b) are positions occuped by criteria with an identical
evaluation. Standardized ranges were utilized in flarther calculations.
After ranging ~ach expert's criter~a, the r, mean t otal .range was determined for
i
criteria provided by a11 of the experts:
, ~
. ~,t _ ?,lt~'1~~ vl,
i=i ~ =i
where m- the t~tal number of experts; 1.2,..., m- the expert's number;
i= 1.2,..., n- the crite rion number. The unanimity of the experts was eval-
uated by utilizing the concordance coefficient w, determined by Kendall's
equation S
n~ m ~
~ Z
- ~2 ( ~ v1) ~~t~-n~-~ T1
)=1
n m
where S - (r1- r) { v~}z;
i=> ;et
n m
- "l 1
r-- r i ln v~;
, t=f i-t
nt m
1 1 ~
~ ~u ~vj~~~-t~~~
i - 1 i-= t
t~ - the number of identical ranges, given by the ~-th exper~t.
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'I'h~: c~,ni'1 0.99; for extractors w= 0.287, confidence level p> 0.99. The
unanimity of the experts was not very high, however it cannot be considered
insignificant. Overall, the arrangement of criteria by their significance
should be acknowledge as successful and logical. Thus, Por filters, the
criterion taking into account the equipment's intensity of operation we.s rated
first; criteria chaxacterizing a single filter's maximal potential performance,
and one of the more important indicators, the amount of waste, followed next.
For extractors, the criterion determining its operational life stood first.
This reflected the tendency presently in existence to utilize the mos` labor-
~ intensive extractors.in the radio-chemiCal industry, those equipped with a
~ mechanical energy feed (centrally powered, mixer-settlers with mechanical mix-
ing, and so forth) for which duration of operational life is precisely one of
the basic indices. The criterion characterizing the duration of the equip-
ment's time phase, i.e. high performance, came next. The arrangement of other
criteria is also non-contradictory in both cases.
Of ma,~or interest is the comparison of expert'opinions from other countries.
Experts from the Atomic Energy Commission of France agreed to respond to the
questionnaire. The responses were processed in the same manner. The unani.mity
of French experts was higher than that of our experts (W = 0.793 for filters,
w= 0.59 for extractors). Characteristically, the differences in extractor
evaluations was higher here as well. Apparently, this is due to the greater
complexity of the extraction process. The distribution of the criteria pro-
vided by the French experts approximates that shown in Tables 1 and 2, even
though there are differences determined by the differing approaches to problems
in other countries. The degree of correspondence in the responses of the
experts from both countries can be conveniently evaluated by Speaxman's rank
correlation p :
3~
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, where x, y are the number of ranks assigned to each crit~rion by Soviet and
French specialists, respectively.
For filters, p= 0.64; for extractors, p= 0.82. This shows a rather high
unity of views.
Naturally, the given criteria systems and their evaluations reflect current
views and approaches to the problem. As science and technology develops, i.e.
as new information appears, conceptions may change; this will decrease the
level of uncertainty in evaluations based on new polls of the experts.
Thus, the suggested methodology does not require designation of an unsubstan-
tiated number of points for evaluating equipment under ca~parison; it also
permits the advisability of utilizing equipment of various types and design
to be evaluated with minimum uncertainty, as we11 as the advisability of their .
application within the framework of plants reprocessing atomic power station
fuel elements.
REFERENCES
1. Kasatkin, A. G., ed., "Zhidkostnaya ekstraktsiya (teoriya i praktika)"
[Liquid Extraction (Theory and Practice)], collection of papers, Moscow,
Khimizdat, 1958.
2. Karpache~~a, S. M., ATOMNAYA ENERGIYA, Vol 47, No 5, 1979, p 324.
3. Beshelev, S. D., Gurvich, F. G., "Matematiko-stazisticheskiye metody
ekspertnykh otsenok" [Mathematical-Statistical Methods of Expert Evalua-
ations], Moscow, Statistika, 1974.
4. Khan, G., Shapiro, S., "Statisticheskiye modeli v inzhenernykh zadachakh"
[Statistical Models in Engineering Problems], Moscow, Mir, 1969.
COPYRIGHT: Energoizdat, "Atomnaya energiya", 1981
- 9875
CSO: 8144/1002
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UDC 621.b39.531
RADIATION DAMAGE TO STEEL IN WATER-1~DERATED, WATER-COOLED REACTOR VESSELS
Ptoscow RADIATSIONNYOYE POVREZHDENIYE STALI RORPUSOV VODO-VODYANYKH REARTOROV
in Russian 1981 (signed to press 27 May 81) pp 2-7
[Annotation, table of contents, and ~ntroduction from book "R~diation Datnage to
Steel in Water-Moderated, Water-Cooled Reactor Vessels", by Nikalay Nikolayevich
Alekseyenko, Amir Dzhabrailovich Amayev, Igor' Vasil'yevich Gorynin and
Vladimir Aleksandrovich Nikolayev, edited by Igor' Vasil'yevich Gorynin,
corresponding member of USSR Academy of Sciences, Energoizdat, 1250 copies,
192 pages]
[Text] The results of resear~~h of a number of scientifi c and engineering prob lems
connected with serviceability of the nessels of water-moderated, water-cooled power
reactors at nuclear power plants are generalized and analyzed. A study is made of
the requirements imposed on materials used ia water-moderated, water-cooled power
reactor vessels, material selection principles and also the properties of used and
proposed materials. The greater part of the paper is devoted to the behavior of
vessel materials under neutron radiation conditions, in particular, in cantact with
the .reactor coolant. The basic laws of variation of strength, ductility and tough-
ness of steel and the metal of welded 3oints sub~ected to neutron radiation are
described as a function of heat treatment, structure, alloying, nature and
amount of impurities and a number of other factors. Information is also given on the
specific behavio'r of hydrogen in irradiated materials.
The book is designed for scientific workers and engineers dealing with the problems
of building nuclear pawer plaa~ts and also for inetructors, postgraduates and atudents
in the corresponding specialties at the institutions of higher learning.
There are 27 tables, 94 illustrations and 321 references.
32
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- Table of Contents
Introduction 5
Chapter 1. Materials Used in Building Water-Moderated, Water-Cooled
Reactor Vessels ~
1.1. Operating Conditions of WER [Water-Moderated, Water-Cooled
Pvwer Reactorj Vessels and Requirements Imposed on Them 8
1.2. Soviet Reactor Vessel Materials 11
1.2.1. Base Metal 11
1.2.2. Welding Materials 1~
1.3. Foreign Reactor Vessel Materials 17
Chapter 2. Influence of Neutron Irradiation on Strength, Ductility and
Toughness of Vessel Materials 20
2.1. General Concepts of Radiation Damage to Steel 20
2.1.1. Variation of Strength and Ductility 22
2.1.2. Variation of BrittlE~ Temperature 29
2.1.3. Influence of Irradiation on Work of Ductile Failure 32
2.1.4. Properties IInder Repeated Static Load 34
2.2. Basic Laws of Variation of Properties of Steel Sub~ected to
Neutron Irradiation 37
2.2.1. Influence of Radiation Flux aad~ Temperature 37
2.2.2. Influence of Neutron Flux Density 4~
2.2.3. Influence of Stresses During Irradiation 44
2.2.4. Recovery of Properties During Annealing 45
2.3. Role of Structure and Heat Treatment of Steel 50
2.3.1. Influence of Austenization and Subsequent Cooling
~ Conditions 52
2.3.2. Influence of Tempering Conditioas 56
2.3.3. Preliminary Aging 60
2.4. Influence of Chend cal Compoaition 66
2.4.1. Influence of Impurities 66
2.4.2. Influence of Alloying Elements 75
- 2.5. Materials of the WER-1000 Water-Moderated, Water-Cooled
~ Power Reactor Vessels 82
2.5.1. Base Metal 82
- 2.~.2. Metal of Welded Joints 84
2.6. C~~mparis an of Radiation Embrittlement of Industrially Produced
Vessel Steels ~ 85
Chapter 3. Mechanisms of Radiation Damage 92
3.1. Cold Shortness of Body-Centered Cubic Metals. Role of Neutron
Irradiation 92
3.2. Mechanisms Controlling Radiation Hardening 95
3.2.1. Radiation Defects Responsible for Hardening of Iron 98
3.2.2. Influ~ence of Interstitial Ir~purities 103
3.2.3. Influence of Substitution Elements 106
3.2.4. Influence of Coherent Segregations 113
3.3. Influence of Impurities on Surface Energy 116
3.3.1. Intergranular Segregation 116
3.3.2. Intracrystalline Segregation lI8
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Chapter 4. Hydrogen Embrittlement of Vessel Material under Irradiation
Conditio~s 124
4.1. Reactor Vessel ftydrogen Absorption 124
4.1.1. Sources of Hydrogen in Vessel Material 124
4.1.2. Absorption and Release of Hydrogen by Irradiated
= Steel 12~
4.1.3. Estimate of Hydrogen Aecumulation in Vessel Material 136
4.2. Influence of Neutron Irradiation and Hydrogen on Mechanical
Properties imder Short-Term Tension 143
403. Influence of Neutron Irradiation and Hydrogen on Cold
Shortness, Static Bending and Eccentric Tension Properties 148
4.4. Recovery of Properties of Hydrogen Charged Steel During Annealing 151
4.5. Low Cycle Fatigue of Steel After Neutron Irradiation and
Hydrogen Absorption 153
4.6. Delayed Fracture of Irradiated Steel imder Effect of Hydrogen 157
Chapter 5. Problems of Operatit~g Reltability of Reactor Vessel.
Estimating the Resistance to Brittle Failure 170
5.1. Brittle Failure Inhibition Criterion 170
5.2. Brittle Failure Initi~tion Criterion 173
5.3. Vessel Radiation Life 176
Bibliography T79
Alphabetic Sub~ect Index 189
~
Introduction
~aenty-five years have gone by since starting up the First Nuclear Power Plant in the
USSR. Whereas at the beginning of the 1950's nuclear power engineering had taken the
first steps in the direction of serving peaceful purposes, in the past two and a
half decades it has been able to win strong positions in the fuel and energy system
of the country and it has become an independent branch of the national economy.
The 26th CPSU Congress stated the goal of generating 1550 to 1600 billion kilawatt-
hours of electric power in 1985, including 220 to 225 billion kilowatt-hours at the
nuclear pawer plants and 230 to 235 billion kilowatt-hours at the hydroelectric power
plants. The increase in electric pawer production in the European part of the USSR
will take place primarily at the nuclear and hydroelectric power plants. New
capacity in the amount of 24 to 25 million kilowatts will be put into production at
the nuclear power plants [1-3].
A significant part of the program for development of nuclear power engineering in
the Soviet Union and the majority of other industrially developed countries is based
on the use of two-circuit water-moderated, water-cooled power reactors (WER) .
Reactors of this type are among the best assimilated. They are distinguished by
great compactness, high use coefficient of the pawer, comparatively low cost,
reliability and simplicity in operation and maintenance. With respect to specific
_ power intensity of the core and economy, the WER are at the.present time some of
the best [4, 5].
34
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Witti respect to chronological attributes and technical indices the WER can be pro-
visionally divided into three generat~ons [6-8]. The first generation reactors
(W~R-210, WER-70, WER-365) confirmed in practice the correctness of the scientific,
technical and design development solutions used as the basis for then. They promoted
the accumu~ation of experience in the industrial operation of WER nuclear power
plants and offered the possibility of training high ly qualified servicE personnel.
The secon d gener ation is made up of the WER-440 reactors used as the basis for the
Iarge series of nuclear power plants already in operation or being built in the
Soviet Union, German Democratic Republic, Finland and in a number of other countries.
Economic calculations and operating experience indicate that the technical and
economic indices of power plants equipped with these reactors are in no way inferior
(~nd in a number of cases are even superior) to the corresponding indices of classi-
cal thermal electric power plants operating on imported organic fuel.. The next
qualitative step in the development of WER (third generation) is manifested in the
WER-500 and the WER-1000 rea ctors with a unit electric capacity of 500 and 1000
megawatts respectively.
One of the most responsible elements of the structural design of WER to a s~.gnifi-
cant degree determining the unit power and operating safety is the reactor vessel.
Direct monitoring of the change in the mechani cal properties of the vessel material
- of an operating reactor and necessary repairs impose great technical difficulties.
High manufacturing quality, which can be insured on ly under the conditions of a
specialized plant, can serve as a definite gua.rantee of the serviceability of a
reactor vessel. In the Sovlet Union, the oossibility of plant manufacture of a
vessel subsequently transported by rail, has formed the basis for all structural
designs since the very beginning of the development of WER. This has led to the
necessity for solving a lsrge number of problems in the areas of material science,
metallurgy and welding technology.
First of all, it was necessary to develop andindustrially assimilate high quality,
low-alloy steel which has high metallurgical adaptability to manufacture, weldability
in large thicknesses and mechanical properties that insure operating reliability of
the vessel for no less than 40 years. The high volume of welding involved in manu-
facturing a reactor vessel has led to the necessity for creating new welding
equipment, the development of welding materials:, welding conditions and technology
guaranteeing the mechanical properties of the welds and high efficiency of the weld-
ing operations. High corrosion resistance of"the vessel material has been insured
by anticorrosive surfacing of the inside surface. In addition, the possibility of
operating WER vessels without anticorrosive protection has been proven. The prob-
lems of quality control of the base metal and welds by highly effective x-ray and
gaimna techniques, ultrasonic, color and luminescent flaw detection have also been
solved.
In contrast to ordinary high-pressure vessels, the reactor vessE~ls are subjected to
- powerful neutron and gatmna radiation causing significant changes in properties :of
the metal. The most unf avorab le of these changes is the loss o_f ductility and
increase in inclination of the steel to brittle failures. Accordingly, the problem
of radiation strength of reactor vessel material.from the point of view of pro-
- longed serviceability has acquired primary signifi cance. The first person to point
to the necessity for studying radiation strength of reactor matc~rials in the Soviet
Union was I. V. Kurchatov. By his initiative a number of scient:ific research
~5
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complexes we~e set up which are equipped with specialized materials science reactors
and "hot" chambers offering the possibility of studying the properties of materials
in intense ionizing radiation fields. In the final analysis, the broad development
of such research has led to the creation of a new field of science solid-state
radiation physics, a component part of which is radiation materials technology.
The general problems of the influence of reactor irradiation on the properties of
metals are fully covered in the book by S.T. Konobeyevskiy [9]. Various theoretical
and engineering aspects of the influence of neutron irradiation on the behavior of
materials designed for making WER vessels have been discussed in the reports by
a number of international and all-union conferences and numerous periodicals. How-
ever, the increasing scale of the construction of nuclear power plants equipped with
- WER has put the question of the necessity for systematization and generalization of
scattered information and the investigation of the problems of r.adiation strength of
reactor vessels as a whole on the agenda. The effort to solve this problem has been
undertaken in the book now brought to the attention of the specialists.
In conclusion, the authors consider it their duty to express their deep appreciation
to P. A. Platonov, G. P. Karzov, Yu. I. Zvezdin and V. Rybin for useful dis-
cussion of indivldual sections of the book, and also to L. M. Lebedev, V. I. Badanin,
E. P. Usatov, A. A. Kuznetsov and A. M. Morozov who participated directly in the
experiments.
- COPYRIGHT: Energoizdat, 1981
10845
CSO: 1861/196
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UDC 621.039.58
RADIATION SAFETY AND PROTECTION OF NUCLEAR POWER PLANTS
Moscow RADIATSIONNAYA BEZOPASNOST' I ZASHCIiITA AES in Russian 1981 pp 2, 64-78
[Annotation and article "Tritium at Ni.iclear Electric Power Plants" by V. V. Badyayev,
Yu. A. Yegorov, V. P. Sklayrov and G. F. Stegachev from book "Radiation Safety and
Protection of Nuclear Power Plants", edited by Professor Yu. A. Yegorov, Atomizdat,
256 pages]
[Text] The fifth edition includes part of the papers discu5sed at the Second
All-Union Scientific Conference on Protection of Nuclear Engineering Facilities from
lonizing Radiation (Moscow, MIFI [Moscow Engineering Physics Institute],
December 1978) and the Branch Conference on Exchange of Experience in Radiometric
and Dosimetric Research at Nuclear Power Plants (Moscow, NIKIET [expansion unknawn],
May 1979). Reports by American specialists heard at the Soviet-American Symposium
on the Problem of Fas t Neutron Reactor Shielding (Obninsk, FEI [expansion unknawn],
November 1978) were also included. The remaining tnaterials will be published in
the sixth edition. .
The results of studying the radiation conditions at nuclear power plants and the
problems of predicting activity are investigated. Beginning with this edition, a
new section has been initia~ed where the environmental protection proble.ms of operating
nuclear. power plants are discussfld. Ma.terials on passage of ~radiation through
reactor shielding and methods of calculating shielding are presented.
The book is designed for specialists in radiation safety and protection of nuclear
power plants.
There are 59 tables, 70 illustrations and 317 references.
37
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Tritium at Nuclear Power Plants, by V. V. Badyayev, Yu. A. Yegorov, V. P. Sklyarov,
G. F. Stegachev
Tritium, which is a superheavy radioactive hydro~en nuclide, plays an important role
iri the problems of radiation safety of. nuclear power plants. Tritium has a halflife
of 12.26 years, it decays with the em~ssion of S-particles with maximum energy of
18.6 kiloelectron volts (average energy 5.8 kiloelectron volts). Tritium belongs in
the category of so-called global radionuclides for it has a quite long halflife, it
can enter into the composition of organic and inorganic compounds and when it gets
into the environment it can spread to significant distances.
Tritium exLsts primarily either in gaseous form~or in the form of tritium water, but
it can also enter into the composition of various organic compotmds. The toxicity
of tritium as a radionuclide depends to a great extent on the form in which it is
found in the environment and in which it enters into the organism of man: in gaseous
form (HT, T2) or in the form of tritium water (HTO, T20). Gaseous tritium modifies
the skin and mucous membrane of the respiratory tracts of man by its R-radiation
(external irradiation) and causes damage to the internal organs of man, being found
in the chemically bound state in these organs. Upon enterin g the environrnent in the
form of tritium water, tritium can enter th2 organism of man both through the
respiratory tracts (trititun water vapor) and through the skin and gastrointestinal
tract (with food and water). It is noted that HTO penetrates the organism through
the entire skin surface of a human just as efficiently as throu~;h the lungs: up to
10 microcuries/minute with a tritium concentration in the air of 1 micr~curie/liter
[I]. When tritium water enters the gastrointestinal tract with food, it is quickly
- assimilated by the oxganism, and a few minutes after ingestion it can be detected in
the venous blood. It is also noted j2] that 24 hours after tritium enters the
organism, the greatest concentration of it is detected in the blood, liver and
sma11 intestine, and after 5 days the tritium concentration in all of the internal
organs is almost identical. According to the most recent findings, the process of
elimination of tritium from the human organism is described by a curve with three
exponential segments. At first, the tritium is eliminated quite intensely (effective
halflife ~10 days). This is connected with the water exchange of the organism.
Later on, the process of tritium elimination decelerates. It has been established
that the elimination rate of organically botmd tritium depends on the type of organ
and biological tissue: the effective halflife for the majority of organs varies
from 10 to 100 days, but there are organs for which the effective halflife exceeds
5 years.
According to the reco~nendations of the International Commission on Radiological
Protection, it is cansidered that on irradiation of the organism of man with
- S-particles with an energy to 0.1 megaelectron volts, their energy is absorbed by the
skin, that is, only the skin is irradiated. Considering that the skin of the entire
body of man belongs to group III of critical or~ans (tissues), the ma~dmtun permissi-
ble dosage of irradiation for which is 30 rem per year, the NRB-76 established the
following mean annual permissible concentrations ~MAPC): 2�10-6 curies/liter in the
air of the work place (for pers:onnel) and 6.6�10- curies/liter in atmospheric air
(for individuals of the population). For tritium in the form of T20 or HTO, the
MAPC are 4.8�10-9 curies/liter in the air of the ;vork place, 1.6�10-10 curies/liter
in atmospheric air and 3.2�10'6 curies/liter in water. Such a J_arge difference in
3$
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values of the MAPC for different forms of e~d.stence of tritium arises from the
difference in radiotoxtcity of these forms.
Usually tritium of natural and artificial origin is distinguished. Natural tritium
- is formed in the u~per layers of the atmosphere as a result of the interaction of
cosmic radiation with 14N and 160 nuclei. Natural trittum exists primarily in the
form of HT or HTO. The natural tritium content in atmospheric air is 1 atom per 1014
hydrogen atom s, and in water 1 tritium atom occurs for 1018 hydrogen ato~, which
corresponds to a specific concentration af 3.2�10'12 curies/liter. The total
natural tritium content on the earth is 20 to 80 million curies.
The primary source of artificial tritium until recently was thermonuclear weapons
testing (to 10 million curies per megaton of thermonuclear blast), as a result of
which the tritium concentration in entities of the environment increased signifi-
cantly: the total tritium concentration on the earth in 1973 was 2900+700 million
- curies, which is appreciably higher than the natural level j3]. Thus, in the 1960's
the tritium concentration in rainwater in the northern hemisphere was on the average
1.6�10-9 curies/liter [4]. Detailed studies of the tritium pollution of the water
of the Baltic Sea in 1972 [5] demonstrated that on the average the tritium concen-
tration in the sea was (1.98~+0.73)�10-10 curies/liter. Curtailment of thermonuclear
weapons testing is naturally accompanied by a decrease in the tritium concentration
in the environment. However, in connection with the development of nuclear power
engineering, nuclear power plants and plants for processing irradiated nuclear fuel
are becoming contributors of tritiitm to the envl.ronment. Processing of uranium fuel
to generate 1 megawatt of pawer is accompanied by the release of 19 curies/gram of
tritium; the same figure for plutonium fuel is 36 curies/gram [3].
An analysis performed in reference [6] shaws that in 1980 the tritium production at
nuclear power plants and plants for processing nuclear fuel will b~e 6.7 megacuries,
and 34 megacuries of it will be accumulated for all years of operation of nuclear
power plants. The same values for the year 2000 will increase to 140 and 720 mega-
curies, respectively. Inasmuch as the greater part of this tritium is discharged
into the environment, it can become a radionuclide significantly influencing the
radiation situation, especially if we consider the possibility of interference of the
discharge of nuclear power plants which in the future will be located comparatively
close to each other. All of this indicates the necessity for careful study of
sources and means of entry of tritium into the environment, accumulation and migra-
~ion of it, the forms of its existence in the environment, possibly the development
of ineans of localizing tritium in order to remove it from the nuclear power plant
waste and other radioactive waste [7].
In nuclear power plant reactors tritium is formed in the following processes:
Directly on fission of the fuel nuclei as a product of ternary fission;
As a result of capture of n~eutrons by deuterium nuclei in the coolant--water in the
form D20;
On capture of neutrons by boron or lithium nuclei in the coolant--water (horon
control, adjustment of the water regime);
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As a result of boron nucleus reactions in the control rods;
During capture of fast neutrons by construction materials.
The contribution of each of these processes to the total amoimt of tritiiaa f ormed in
a nuclear power plan.t is determined by the type of reactor, its parameters, operating
conditions and materials used in the reactor.
One of the primary sources of tritium is formation of it in the fuel elements during
ternary fission; the triton pield per fission of 235U [siJ is on the average
8.7�10'3%, that is, 1.23�10-Z curies of tritium is formed per megawatt of thermal
power of the reactor in the fuel elements per day. The fission process as a source
of tritium can be characterized by the following example: by 19ti0 an accumulation of
about 3�106 curies of tritium is expected in the fuel elements of the nuclear power
plants of Great Britain, and by the year 2000, about tenfold more [9].
As a result of diffusion through the fuel jacket and also through cracks and micro-
cracks in the jackets, tritium can get from the fuel to the coolant. It is natural
that under other equal conditions the tritium leakage out of the jackets depends on
the amount accumulated in the fuel element. Estimates made on the basis of experi-
mental data [10] indicate that the tritium leakage rate from fuel jackets with a
microcrack surface are~ of L of the entire jacket surface for fuel in the fo rm of
sintered uraniinn dio~dde is characterized by a coefficient of 0.86 per day. Accord-
ing to other data [llJ, the tritium leakage from fuel jackets made of corrosion
resistant steel is 1%, and from zirconium alloyed ~ackets, 0.1% of the total amount
of tritium under the jacket. Considering these data, the amotmt of tritium getting
from the fuel elements into the coolant can be defined by the ratio [12]
QT=1.32�10-16W, curies/sec, where W is the thermal power of the reactor, watts
(zirconium fuel ~ackets).
Table 1. Thermal Neutron Reactions with Tritium Formation
Content of target Thermal neutron
Rea ction nuclide in a natural reaction cross
mixture, % section, Q
;ti c,~. ;r? o,ois io--~
21 fe i, P) ~ FI I, 3� 10-� 5,�~ . I 0~
~I.i (n, a) ~F{ 9,3G�1~-
~I_i (n. rt' ,'~He) j H 92 ~�19 p~p`~�
Io~~rt. ~a)'~H ~q ~ p ~~n,
*Cross sections are averaged with respect to fission neutron spectrtnn.
The probability of tritium formation as a result of other reactions is characterized
by the data presented in Tables 1 and 2. It is obvious that the basic reactions of
tritium formation in nuclear reactors are the deuter~n and l~B nucleus (n, Y)
reactions. The D(n, Y)T reaction leads to the formation of tritium directly in the
coolant, and the amount of triti~un formed by this reaction in as1 ordinary water-
cooled reactor can be defined by the formula
1+0
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QT = 27.6nHV~� 10-19 curies/sec,
where nH is the nuclear density of hydrogen in the coolant, cm 3~ V is the coolant;
volume in the reactor core, cm3; ~ is the average thermal neutron flwc density with
respect to the coolant, sec'1, cm 2.
The 10B nucleus reactions, as has been noted, are possible during boron control--in
the coolant and in the rods of the control and safety rod system c?f the reactor.
Table 2. Fast Neutron Reactions raith Tritium Formation
Neutron Target Average
Reaction energy, nuclide can- value of the
Mev tent in reaction
natural cross section,
mixture % Q
ia~ ~,t isC 5-10 99,G3 3�10-.~
a[,~ (n, u~ ~F-( 0,01-10 7,52 ' 1,0
'Li (n, n', ~Hc C,-14 92,48 4� 10-1
lu[3 {ri, a) T 1-9 19,8 7,5�
a_b ~n f~ :ioK I�3,7 9~,018 Z. ~p-s
ao~~ ~n t~ ssh 14 , 7 96 ,~7 0, l� 10-s
saj:~ ~~i s~~~itt 14.G 5,84 7,~� lU-e
'~l:c (rt t) a=,a ~In l~t,~ 5,b4 6� lU-+
The contribution of the various tritium formation reactions to the total amoimt of
tritium in the coolant was estimated in reference j12] in the example of the
Rlieinsberg nuclear pawer plant for its operating period during boron control testing
in 1970:
Yield from fuel elements 0.5
Yield from the rods of the control
and safety rod system 0.15
Formation in coolant as a result of
the following reactions:
D~n~ Y)T 0.18
1 B~n~ 2a~T 5.23
lOg(n~ n?~ a)T lp-4
It is obvious that more than 85% of tritium is formed as a result of the fast neutron
lOg~n~ 2a)T reaction. The authors of reference [12] note that this does not agree
with the conclusions, for example, of reference j13], in which it was demonstrated
that the primary reaction of tritium formation under such conditions must be the
interaction of thermal neutrons with lOg.
The results of the approximate calculation j4] of the specific activity of tritium
in the coolant of the Kol'skaya Nuclear Power Plant also demonstrated that in
WER-440 water-moderated, water-cooled power reactors with boron control (boric
acid concentration 3.4 g/kg) th~ primary reaction of tritiwn formation is the
- ~ 41
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10~~n~ 2rx)T reaction: during the operating time of the reactor of 7000 effective
hours, the specific tritium activity in the coolant of the primary circuit as a
result of this reaction is approxtmately 10- 3 curies/liter, at the same time as the
specific activity will be about 5�10-6 curies/liter as a result of deuteriian activa-
tion and double activation of protium.
For nuclear power plants with WER-50 reactors [12], the estimates shaw that in
one r~m, the coolant picks up as much as 260 curies of tritium: 5.23 curies as a
result of ternary fission, 3.22 curies as a result of protium activation and
252 curies as a result of the 10B reaction (during boron control).
The presented calculated values agree qualitatively with the.data on the actual
tritium content in the coolant of a number of operating nuclear power plants with
pressurized water reactors ga~hered, for example, in reference j14] (curies/year):
Indian Point-1 (W=163 megawatts), 500; Yankee (W=183 megawatts), 1300; Haddon Rock
(W=462 megawatts), 1733. Al1 of these nuclear pawer plants have b oron control and
fuel elements in corrosion-resistant steel 3ackets. In reference [15] it is also
pointed out that the specific tritium activity in the water of the primary ~ircuit
of nuclear power plants w~ith PWR falls within the limits of (0.1 to 5)�10-3 curies/
liter as a result of introduction of boron into the coolant. The tritium activity
in the water of the primary circuit of the Novovoronezh Nuclear Power Plant [16]*
is 2.2�10'6 curies/liter for block I, 21�10-6 for block II, 72�10-6 for block III
and 95�10-6 curies/liter for block IV. The maximum recorded tritium concentrations
in these blocks in the period from 1973 to 1975 was 3, 10, 520 and 340�10'6 curies/
liter, respectively [17]. At the Kol'skaya Nuclear Power Plant, 3 months.after
beginning operation, the tritium concentration in the coolant was 31�10-6 curies/liter
~ [17].
It is natural that for nuclear power planrs with reactors of other types, the amount
of tritium formed as a result of one reaction or another and, consequently, its
- concentration in the coolant will be different. Thus, for nuclear pawer plants with
heavy water reactors the primary tritium formation reaction is deuteriwn activation.
Estimates show that in heavy water reactors up to 30 curies/year of tritium are
formed per megawatt of electric power. At nuclear power plants with boiling rea.ctors,
the tritium formation as a result of ternary fission and as a result of reactions
on the boron nuclei in the control rods is significant, but the primary source of
tritium is deuteriinn activation. Thus, calculation of the amount of tritium formed
by the D(n, Y)T reaction at the Dresden-1 Nuclear Power Plant demonstrated [1] that
the rate of trititun formation is 6�10'8 curies/sec, and experim~ntal information on
the accumulation of tritium in the coolant gives a rate equal to approxl.mately
5�10-8 curies/second, that is, it is obvious that the triti~ is forn~ed primarily
as a result of deuterium activation. Estimates made for nuclear pawer plants with
RBMK-1000 reactor also demonstrated that in the multiple forced circulation circuit
tritium is formed primarily as a result of the D(n, ~y)T reaction, its specific
*See also the survey: N. G. Gusev, "Provision for Radi~tion Safety at Nuclear Power
Plants," ATOMNAYA ENERGIYA [Nuclear Power], Vol 41, DTo 4, 1976, p 254.
42
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� activity in the multiple forced circulation circuit is low, and it amounts to
about 4�10~~ curies/liter.
The actually measured tritium concentrations in the coolant of nuclear power plan ts
with boiling reactors are characterized by the following values:
Blocks I and II of Beloyarskaya Nuclear Power Plant 20�10-6 and 28�10'6 curies/liter
[16], maximum recorded concentration for block I 29�10-6 curies/liter [17];
For b lock I of Leningrad Nuclear Power Plant 0.48�10'6 curies/liter [17];
At the nuclear power plant wj.th RBMK-1000 reactor (on the average) during the initial
operating period at stable power level, in the water of the multiple forced circula-
tion circuit and in the saturated steam condensate (1 to 2)�10-~ curies/kg;
At a nuclear power plant with VK-SO reactor the maximum recorded concentration is
4�10'6 curies/liter [17].
The amount of tritium produced by nuclear pawer plants with boiling reactors in a
year is characterized by the followi.ng data [14]: Dresden-1 (W=200 megawatts),
2.9 curies; Humboldt Bay (52 megawatts) and Big Rock Point (W=50 megawatts),
20 curies each.
- At n uclear power plants with channel boiling reactors, in addition to the multiple
force d circulation circuit, tritium is also formed in the control and safety rod
system circuit and in the gas circuit, that is, in the gas blown through the graphite
stacking of the reactor. In the control and safety rod system circuit at a nuclear
power plant wirh RBMK-1000 reactor, about 2.5 curies of trititmn are formed per year
as a result of reactions on the boron nuclei (under the assumption that the tritium
leakage from the jackets of the rods of the control and safety rod system is 0.1%)
and about 1 curie as a result of deuterium activation. The tritium noncentration in
the co ntrol and safety rod system loop in the absence of leaks is ~5�10'6 curies/liter.
Tritium is formed in the gas circuit as a result of reactions on helium nuclei. At
the nuclear power plants with RBMK-1000, up to 800 curies of tritium are formed in
the circuit, and its average concentration in the gas w~;ll be =6�10'~ curies/liter.
The results of ineasurements of the tritium concentrations give the following values:
in the control and safety rod system circuit (2 to 6)�10-~ curies/liter; in the gas
circuit, ~10'9 curies/liter of tritium in the form of tritium water vapor.
In general, it is impossible to note good convergence of the calculated and measured
triti um concentrations in the various circuits of nuclear power plants. At the same
time it is necessar}~ to note defined scattering of the experimental data on concen-
trations. All of this is connected, on the one hand, with the fact that it is
dif.fi cult to consider all the factors influencing the tritium concentration in the
calculations, for example, the state of the core, the actual magnitude of the leaks:
the water chemical regime of the coolant obviously influencing the chemical farms
of e xistence of tritium, the amount of impurities in the scavenging gas, and su
on.
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On the otheX hand, the majority of experimental data are the results of ineasuring
the tritium concentration in oxidized form, that is, in the form of tritium water.
The discovery of the influence of various operating conditions of nuclear power
plants and the sta~e of its components on tritium concentration in the p rocess
circuits is one of the proble~ which must be solved for reliable prediction of the
influx of tritium inta the enviranment and the development of devi~es for localiza-
tion of tritium if the necessity for such arises. It is naw possible to state that
at nuclear power plants with boiling channel RBMK-1000 reactors the tritium concen-
tration is below the MAPC of tritium for open bodies of water, and at the nuclear
power plants with water-moderated, water-cooled power reactors it is higher than
the MAPC by 10 to 100 times and no more.
The tritium forma.tion in the coolant circuit gives rise to its presence in the gas
and liquid effluents of nuclear power plants, and also, primarily as a result of
coolant leaks, in the air of the work places of the nuclear power plants. Thus, the
activity of the liquid discharge of many nuclear power plants with PWR is 85 to 99.9%
determined by tritium, and in 1973 it was from 16 to 4000 curies/liter [18]. The
maximum tritium activity in the liquid discharge of the Novovoronezh Nuclear Power
Plant (in 1973 to 1975) and the Kol'skaya Nuclear Puwer Plant were below the MAPC
of tritium in open bodies of water and equal to 10'~ to 10-6 and up to 2.5�10-6
curies/liter, respectively [17]. The tritium discharge with liquid waste from
nuclear power plants with boiling reactors is with respect to activity also a large
- portion--from 9 to 90%-- and at the Vermo.nt Yankee Nuclear Po-,aer Plant, 99.9%, but
with respect to absolute magnitude, the activity of the discharged tritium is
appreciably less: from 4 to 50 curies less (for 1973), for example, (at the Vermon.t
Yankee Nuclear Power Plant, 0.2 curies/year) [la]. It is natural that the tritium
concentration in the liquid waste of nuclear power plants with boiling reactors is
also lower: maximum tritium concentrations in the waste water of the Beloyarsk
and Leningrad Nuclear Power Plants do not exceed 0.26 to 0.27 curie/liter [17], in
the industrial sewage of Leningrad Nuclear Power Plant (6 to 7)�10-9 curies/liter,
that is, significantly below the MAPC for water. This conclusion agrees with the
- conclusion drawn in reference [19] that the tritium concentration at the liquid
waste discharge point into the Kaporskaya Bay on the Gulf of Fi.nland is more than
an order below the permissible concentrations in potable water.
The tritium discharge with llquid waste from nuclear power plants in the majority of
countries has not been standardized. In the United States before 1971 it was per-
missible to discharge liquid waste with tritium concentration of 3�10-6 curies/liter,
b ut from the materials of the Third Geneva Confe rence on Peaceful Uses of Atomic
~ Pawer it is known that in recent years this permissible discharge has been decreased:
- the tritium concentration in the waste before reservoir dilution must not exceed
5�10-9 curies/liter, that is, approximately no more than 25 times the average tritium
concentra~ion in the water of the Baltic Sea.
In the gas effluents of nuclear pawer plants the tritium content is much less than
in the liquid effluents, which is naturally explained by the inclination of tritium
toward oxidation. Thus, the activity of tritiimm discharged in 1973 by nuclear
power plants with boiling reactors falls within the limits from fractions of a curie
to 75 curies [18J, and in percentage ratio this is 10-3 to 10'4% of the total
discharged activity. In reference j20] a theoretica.'~ estimate was made of the power
44
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of the tritium disctiarge for nuclear power plants wi[h Wt:K-440 reactor. '1'eniriry
fission of 235U, activation of deuterium and basic reactions on 10B nuclei were
considered as the means of formation of the trititun. It was established that in the
absence of unorganized primary circuit coolant leaks, the primary source of tritium
at nuclear power plants is the water purification ponds, in the water of which the
tritium concentration can reach 1.8�10-4 curies/liter. Tritium gets into the dis-
- charge from the ponds with water vapor, and the pawer of the discharge can be about
130 curies/y.ear. In the case of tmorganized leaks of 0.2 tons/hour as a result of
partial evaporation the discharge power increases to 160 curies/year.
The presence of tritium in the air of the work places is prompting some specialists
to consider constant monitoring of the tritium content in the organizing of workers
at nuclear power plants ne cessary, 3ust as is done, for example, for iodine radio-
nuclides. Actually, at the nt:clear center in Karlsruhe, 7.2% of the personnel have
from 0.1 to 1 times the p ermissible amount of tritium in ~the organisms, and 60% of
the workers are tritium carriers; at radioisotope production facilities, the internal
irradiation by tritium is also on the ma~dmum permissible level [21].
At the Kozloduy Nuclear Power Plant (People`s Repub lic of Bulgaria), during the
initial period of operation, the tritium concentration in the air oP various rooms-
was (1.3 to 3.4)�10-13 curie/liter j22]; according to the estimates of reference [20],
under the above-described conditions in the facilities of nuclear power plants with
WER-440 tritiim? will be detected, but its concentration will not exceed the MAPC
for work places established by the NRB-76. The measurements made at the
Novovoronezh Nuclear Power Plant demonstrated that the maximum recorded tritium con-
centrations in the work places of the nuclear power plant fall within the limits of
(64 to 180)�10'9 curies/li ter. At nuclear power p lants with boiling channel reactors,
- these values are lower: at the Beloyarsk Nuclear Power Plant, at the background
- level, and at the Leningrad Nuclear Power Plant, to 10-9 curies/liter [17].
As is obvious, the data presented here on the actual liquid and gas discharge of
tritium by the nuclear power plants do not permit establishment of any laws. On the
contrary, the significant scattering of the information about the effluents even for
nuclear power plants with like reactors attracts attention. This is caused not only
by defined differences in the process flaw chart~s of the nuc~ear powex T~lants and not
only by insufficiently complete study of the sources of trititun at the nuclear power
plants, as was noted above, b ut also by insufficient study of the ways that tritium
gets into the waste facilities, the laws of its accumulation in the coolant. It is
natural that the absence of such information complicates even m~re the prediction of
the accumulation of tritium in the biome as a result of operation of the nuclear
= power plants and obtaining detailed information about the interrelation of the
process flow chart of a nuclear power plant, its operating conditions with the
formation and delivery of tritium to the waste units--the primary element of the
triti~un problem at nuclear power plants.
The experimental study of the behavior of tritium in the process circuits of nuclear
power plants and in the environment presupposes the taking of samples from the media
in the circuits and the objects of the experimental environment and meas urement of
the tritium concentration in them. It is possib le .to note the following peculiarities
of this experimental work:
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1) The investigated media can be different (water, steam, gas, air), and tritium
can be found in the same medium in different chemical forms (for exa~le, tritium
water and gaseous tritium in the water of the multiple forced circulation circuit,
in atmospheric air);
2) The activity of the samples, as is obvious from what has baen stated above, ia
low, but the~measurement results must be obtained with good accuracy inasmuch as
they are needed for forecasting purposes;
3) The energy of the tritium S-particles is quite low; therefore for reliable
recording of them, detectors and equipment with low intrinsic noise level and low
natural background are needed.
However, at the present time reliab le methods of taking samples of various media,
concentration of them (if necessary) and measurement of the activity, which will
permit us to obtain qualitative information about tritium at the nuclear power plants,
have been developed and are in use in practice [10, 23-25].
.r ` /S
~8~ h?isl.~ \ ~ O � J �~I - [
~ ~ O~~ ~ I I
~ O~Or~, ..~~C. I' 1 .
~b~ ~7vo6,z ~
~
p ~ r--
~ i
_L_--r-
_I``.. _ i..
- - E
~ Figure 1. Diagram of a device for taking tritium gas samples from air:
_ 1-- mixing chamber; 2-- thermocouple; 3-- furnace; 4-- tube with
pelletized copper oxide; 5-- quartz wool; 6-- Dewar flask with
liquid nitrogen
Key: ~
a. Nitrogen
b. Gas sample
If it is necessary to determine the tritium water content in water, the sample is
taken directly, that is, a vesse~ is filled to the appropriate volwme, mechanical
impurities and other radioactive nuclides are removed, for example, using an ion-
exchange resin, and the sample is stored in a sealed container until analysis
in order to exclude evaporation, inasmuch as evaporation can distort the results
as a result of isotopic effects. When determining the tritium gas concentration in
water, the sample is taken sealed. Gaseous tritium is extracted from the sample
by bubbling. A sample of atmospheric moisture is collected either by freezing out
~ 46
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uFSin~; cooling units or by pumping air through water absorbers, for exam{~le, through
concentrated sulfuric acid or silica gel [24]. On 2 kg of silica gel it is possible
to collect up to 250 ml of water. The water is isolated from the silica gel by a
closed current of dry air.
The gas samples for determining the concentration of gaseous forms of tritium are
taken with the help of a compressor in bottles or elastic shells. However, if a
sample must be stored for a prolonged period of time, this method of sampling can
lead to errors as a result of uncontrolled tritiimm loss by diffusion through the
walls of the sampling vessel. It is more reliable to take the sample by final oxida-
tion of volatile forms of tritiimm in a special device through which gas or air is
pumped. Final oxidation of tritium is ca.rried out in a nitrogen current on
pelletized copper oxide (Figure 1) at a temperature of 600�C. 7."he vapor in the formed
water is frozen out in a U-tube. By selecting the oxidation conditions (catalyst,
temperature, and so on) properly, it is possit~le to obtain infoi-mation about the
triti.um content in various chemical forms.
1
/
_ ~
_ - - 2
-
_ - _ - - `
i - . -
_
~
, Figure 2. Diagram of a hydrolyzer for concenrrating tritium samples:
1-- cooler; 2-- electrodes; 3--- water samp.~e s ubject to concentration;
4-- vessel for cooling the electrolysis cell.
If it is known in advance that tririum activity in the saniple is small, the sample
is concentra~ed. The difference in hydrolytic decomposition rate of the molecules
of ordinary, heavy and tritium waters is used for concentration. Concentration is
carried out in special devices called hydroly2ers (see Figure 2), and if necessary
the process is repeated several times. Usually eightfold to tenfold enrichment of
the sample with tritium water molecules is achieved in one cycle.
For determining the tritium activity in the sample, the scintillation or ionization
method is used. For implementation af the ionization method, etther proportional
internal filling meters (sometimes flow meters) or ionization cliambers (for
measuring high trititan activities in the sample) are used.
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In the scintillation method usua.lly liquid scintillators based on dioxane are used,
for it is possible to add a sample to diaxane in a volume of up to 20% of the volume
of the dioxane without having significant negative effects on its conversion
efficiency. Various scintillating additives and spectrum displacers are used; ready-
made ZhS-7 ~nd ZhS-8 (dioxane+a-methylnaphthaline+BPO) liquid scintillators are
suitable for recording tritium radiation. The conversian effectiveness of liquid
scintillators becomes noticeably worse on "poisoning" of them ~~ith oxygen and
various impurities, especially impurities containir~ sulfur and chlorine. Therefore
before introduction into the scintillator, the sample is purified of its imp urities
on ion exchange resin or by distillation, and after introduction of it into the
scintillator, the mixture ready for measurements is purged with argon to remove the
oxygen.
Th~ activity of the sample is measured on imits with l, 2 or 3 photomultipliers.
The advantage of units with one photomultiplier is simplicity. HoweveY, in such imits
it is necessary to use a photomultiplier with high sensitivity of the photocathode
and low energy equivale~it of the intrinsic noise. Sometimes the photocathode is
cooled for this purpose. On the USS-1 unit without cooling of the photocathode of
the photomultiplier, a sensitivity of -10-~ curies/liter of tritium is acPPeoved by
applying a scintillator based on dioxane with naphthaline (100 g/liter),
(8 grams/liter) and POPOP ~~�2 grams/liter). On units with :~everal photomultipliers
noise pulses can be avoided to a significant degree, and it is possible to determine
activity on the level of 10-10 curies/liter. The known stan.dard foreign-made devices
(TRICARB, SL-20, and so on) have a sensitivity of ~10-9 curies/liter.
Devices for preparing a sample with measuring by the ionization method and propor-
tional internal filling counters with active shielding made of gas discharge coim�t-
ers are described in reference j23]; they fully satisfy the requirements of the
problems solved when investigating the formation and transport of tritium at nuclear
power plants. Such devices T-1 and T-2, which are distinguiahed by small structural
changes from those investigated in [23}, are described be~low. These devices are
used in operations performed at the nuclear power plants , and they are distinguished
by their sensitivity: the T-1 is a device with relatively low sensitivity, to 10'8
curie/liter.
In the T-1 device, an internal filling counter ~ith a volume of 600 cm3 (Figure 3)
with cathode in the form of a corrosion-resistant steel eylinder and ~ith anode
made of, gold-plated tungsten wire 20 microns in dian?eter, is used. In contrast to
the counter described in reference [23], the flange seal of its ends has been
replaced by a seal using a nut, which has made it possible to reduce the overall
dimensions of the counter and to use mat for its protection and lead shielding from
the UMF device. In the massive organic glass end covers, devices are motmted for
putting tension on the filament--enode--an intake connection for the investigated
gas, a socket for connecting the electronic unit and thin-walled windaw of lavsan
film (50 microns thick) for admitting 55Fe x-radiation into the counter
_ *The autiiors express their appreciation to L. I. Gedeonov, V. I. Blinov,
V, P. Tishkov and their colleaguea for consultation and assistance in ad3usting the
units.
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�5.9 kev) used to select the opexating conditions of the coimter. The mat is
as~embled from the MS-6 type gas discharge cotmtera, and it is included anti-
coincident with the main cotmter. The lead..shielding is 50 mm tliick. The system for
filling the cotmter with the wor-king gas mixture is analogous to that described in
reference [23]. The ~vorking gas mixture is prepared from propane (90%) and the
investigated sample of hydrogen with triti~ (I0~); during the measurements a
pressure of 1 to 1.2 atm~spheres is maintained in the counter volume.
/ > .
~
~
~ ~ t~~ ,
� . :
: `t .
. ; ' n' / 5
~i ;
I ~
. I .
~
- .
: .
~
~ 6 '
; ~S
~ , ~
' )
. ~ .S%~
,;~~ij %
_ : _ % ' 2
~ ~
Figure 3. Structural design of a gas discharge internal filling covnter:
1-- cotmting gas intake caanection; 2-- pressure nuts; 3-- counter
housing; 4-- end type organic glass inaert; 5-- rubber gaskets;
6 lavsan film
The T-2 device coneists of an internal filling counter with a volwme of 4500 cin3
(diameter 120 ffin, length 650 mm) of analogous atructural design placed in the mat
of gas discharge counters type SI-6G. The counting module is placed in the cavity
(300X300X900 in lead shielding 150 ~ thick with sliding door. Just as in
the T-1 devlce, the known j23] system for preparing the sample and filling the
working volume of the counter with it is used (see Figure 4).
For meas urements an the T-1 devl.ce the samples are prepared in advance by decomposi-
tion of water by calcium oxide and zinc; on the T-2 device the sample is prepared
either in the counter filling system by decomposition of water by ~nagnesium [22]
in a special tube furnace or in advance.
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~ ~8~ 1~
6y.~:rc~ueH ~ iJ I I
^ ~I 1F I
~r~cnc!r ~,L, ~ t~ I
~b) ~
- ~
~ ~ J_' _1
~
.
~
.V 1-: - ~ i;i r-; _-i �
t ~ I~~ \y~ ~ ~ -
i ; -1- ~ ,
~ ' i 2G
- 1 ~ 6,
~~~5c~~;.~03 K 6cKyy;~- ~
uc,t~y r:acecy CaH~a
- ~ i
- ~ I I I
- 17e?c,�~ N~ZO ~ 21 ~
~d, . r-_--
Figure 4. Diagram of a stand for preparing low-active tritium samples
_ for measurements:
~ 1-13 vacuum valves; 14 vacuum meter; 15 vessel for mi~d.ng
butadiene and hydrogen; 16 vessel for storing excess $as;
17 H2 absorption; 18 HZ generator; 19 cotmter; 20 catalyet;
21 liquid nitrogen
Key :
a. Butadiene
b. Propane
c. Hydrogen for a vacuum pump
d. H20 feed
The electronic modules of both devices are the same, they have a total channel gain
of ~1000 with stability of no worse than 1%, deviation fram linea�rity of the ampli-
tude characteristics in the tritium S-particle energy range does not exceed +1%.
BIBLIOGRAPHY
1. Yuzgin, V. S. and Yavelov, B. Ye., "Tritium and Environment," ATOMNAYA
TEK~INIKA ZA RUBEZHOM [Nuclear Engineering Abroad], No 10, 1~73, p 24.
- 2. Moskalev, Yu. I., Okis' tritiya [TritfLUn Oxide], Moscow, Atomizdat, 1968.
3. Telushkina, Ye. L., "Radiation Hygienic Evaluation of Triti~m as a Factor of
Enviranmental Pollution," Report at the 3d All-Union Scient:tfic and Practical
Conference on RadiaCion Safety, Moscow, 1976.
4. Lur'ye, A. N., "Nuclear Power Plants and the Tritium Problem," SB. TRUDOV NP0
- "'ENERGIYA" [Collection of Works~of the "Energiya" Scientifi~ Production Associa-
tion], No 2, Moscaw, 1975, p 144.
50 .
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5. Vakulo"vskiy, S. M., Katrich, Xu. I., Malakhov, S. G., et al., "90Sr, 137Cs
and Tritium Contents in the Baltic Sea in 1972," ATOMNAYA ENERGIYA jNuclear
Paurer], Vol 39, No 3, 1975, p 183.
6. Gedeonov, L. I. and Trusov, A. G., "Environmental Protection in Connection with
the Development of Nuclear Pawer Engineering," ATOMNAYA TEKfINIKA ZA RUBEZHOM,
No 12, 1973, p 22.
7. Gelkin, B. Ya., Gedeonov, L. I., Demidovich, N. N., et al., "Discharge of
Volatile Fission Products into the Atmosphere During Operation of Nuclear
Power Plants and Installations for Regenerating Spent Fuel and Future Prospects
for Trapping these Volatile Products," ATOMNAYA ENERGIYA, Vol 44, No 2,
1978, p 145.
8. Sloth, E. N., "Tritium in Thermal Neutron Fission of 235U," J. INORG. NUCL.
CHEM., Vol 24, No 4, 1962, p 337.
9. Danster, H. G., Warner, B. F., The Disposal of Noble Gas Fission Products
from the Reprocessing of Nuclear Fuel," UN. KINGDOM ATOM. ENERGY QUARTERLY,
1970.
10. Turkin, A. D., DOZIMETRIYA RQDIOAKTIVNYKH GAZOV [Radioactive Gas Dosimetry],
Moscow, Atomizdat, 1973.
11. Smith, I. M., "The Significance of Tritium in Water Reactors," GENERAL ELECTR.
COMP., Sept. 19, 1967.
12. Langekker, K. and Graupe, Kh., "Tritium in Pressurized Water Reactors,"
DOKLAD NA II SIMPOZIUME SEV "VODNYYE REZHIMY VODO-40DYANYKH REAKTOROV,
RADIATSIONNYY KONTROL' TEPLONOSITELYA I SREDSTVA SNIZHENIYA RADIATSIONNOY
OPASNOSTI TEPLONOSITELEY [Report at the 2d Symposium of the CEMA "'Water Regimes
of Water-Moderated, Water-Cooled Reactors, Radiation Monitoring of the Coolant
and Means of Lowering the Radiation Danger of Coolants], German Democratic
Republic, Stral'sund, 1972.
. 13. Ray, J. W., "Tritium in Power Reactors," REACTOR AND FUEL PROG. TE QiNOLOGY,
Vol 12, No 1, 1968-69.
14. Kim ze, B. and Vimund, K., "Problem of Monitoring Tritium in Nuclear Facilities,"
- see reference [12].
15. Sawochka, S. G., "Sampling and Analysis Procedures in Water Reactors,"
PROC. AMER. POWER CONF., Vol 38, 1971, p 741.
16. Lur'ye, Ye. L., "Tritium Dosimetry at Nuclear Power Plants," see referance [3].
17. Abolmasov, Yu. P., "Tritium Content in Liquid Media and the Air of the Work
Places of Nuclear Power Plants," ATOMNAYA ENERGIYA, Vol 41, No 3, 1976,
p 215.
; 51
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18. "Summ~axy of Radioactivity Released in E~~luents from Nuclear Power Plants
during~:.1973," NUCL. SAk'ETY, Vol 16, No 6, 1975, p 734.
- 19. Abkin, A. D., Agap~Cina, N. P., Aleksandrov, A. P., et al., ATOMI~TAYA NAUKA
I TEKHI~IIIKA V SSSR [Nuclear Science and Engineering in the USSR], Moscow,
Atomizdat, 1977.
20. Velikovskiy, A. A. and Chernousov, S. A., "Estimating the Equilibriwn Concen-
tration of Tritium and Its Discharge into the Environment at l~uclear Power
Plants with WER-440," see reference [3].
21. Balonov, M. I. and Kupryashchin, Yu. N., "Study of the Tritium Content in the
~rganism of Workers," TR. IV NAUCH.-TEK~IN. KONF. PO DOZIMETRII I RADIOMETRII
IONIZIRYUSHCHIKH IZLUCHENIY [Works of the 4th Scientific and Technical
Conference on Dosimetry and Radiometry of Ionizing Radiation], Section II,
edited by V. A. Knyazev, Moscow, Atomizdat, 1972.
22. Khitov, K. and Konstantinov, Ye., "Radiation Shielding and Dosimetric Monitoring
at Nuclear Power Plants--First Operating Results," MATERIALY KONF. PO
PROBLEMAM RADIATSIONNOY BEZOPASNOSTI PRI EKSPLUATAZ'SII ATUMNYKH
ELEKTROSTANTSIY [Materials of the Conference on Problems of Radiation Safety
~ in the Operation of Nuclear Power Plants], Czechoslovakia, Prague, 1975.
23. Gedeonov, L. I., Blinov, V. A., Stepanov, A. V., et al., "Sample Collecting
and Tritium Analysis in the Ground Layer of the Air," MATERIALY SIMPOZIUMA
PO NABLYUDENIYU ZA OKRUZHAYUSHCHEY SREDOY VBLIZI YADERNYKH USTANOVOK
[Matezials of the Symposium on Observation of the Er.vironment Near Nuclear
Facilities], Polish People's Republic, Warsaw, 1973.
24. Ledeonov, L. I., Blinov, V. A., Stepanov, A. V., et al., "Set ot Units for
' Taking Samples and Measuring Tritium in Ob~ects of the External Environment,"
ATOMNAYA EN~RGIYA, Vol 42, No 5, 1977, p 361.
25. Lomonosov, I. I, and Soshin, L. D., IZMERENIYE TRITIYA [Measurement of
= Tritium], Moscow, Atomi.zdat, 1968.
~OPYRIGHT: Atomizdat, 1981
10845
CSO: 1861/198
,
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NON-NUCLEAR ENERGY
UDC 532.538.4
- INDUSTRIAL MAGNETOHYDRODYNAMIC EQUIPMENT AND PROCESSES
Kiev TEKHNOLOGICHESKIYE MGD USTANOVKI I PROTSESSK in Russian 1980 (signed to press
21 Oct 80) pp 2, 189-190
[Annotation and table of contents from book "Industrial Magnetohydrodynamic Equipment
and Processes", by Anatoliy Fedorovich Kolesnichenko, Ukrainian SSR Academy of
Sciences, Electrodynamics Institute, Izdatel'atvo "Naukova dumka", 1000 cop~~s,
191 pages]
[Text] In this monograph a study is made of qualitative and spatial conversions in
limited domains filled with electrically conducting drop liquid or electrically con-
ducting gas and placed in physical fields--electric, magnetic and gravitational.
Among such conversions are transformation and transport of energy--conversion of
electromagnetic energy to heat and mechan~cal work, transport of the heat and material
mass of a liquid conductor under the effect of various combinations of electromagnetic,
capillary and thermoconvective effects. The results of theoretical and experimental
studies of a new class of magnetohydrodynamic flaws--ca~illary flows that occur
during arc and induction working of inetals--are discussed. New industrial procedures
for working alloys based on the application of MHD phenomena are described.
The book is designed for scientific and engineering-technical personnel interested
in the development and application of magnetohydrodynamic devices. '
There are 82 illustrations, 1 table and 119 references on p ages 182-188.
Table of Contents
Foreword 3
Basic Symbols used in the Text 6
Chapter I. Basic Equations of Magnetohydrodynamics 9
Chapter II. MHD Phenomena in Electric Arc Welding 18
1. Volumetric and Surface Forces in Electrically Conductin~; Drops,
Jets and Plasma 18
2. Form of the Phase Transformation Interfaces During Weld_Cng Arc
Fusion of A Cylindrical Electrode 30
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3. Quasisteady States of the Free Surface of Drops and Jets of
~lecXrode Metal 40
4. IJrop~Formation and Transport in Arc Fusion of a Cylindrical
Elec~rode 46
5. Drop~Formation and Transport of Electrode Metal During DC Welding
with the Application of Current Pulses 54
6. Drop Formation and Melt Transport in Electric Arc Welding with
Nonsteady Electrode Feed 5~
7. Drop'Forme.tion in Magnetic Pressure Modulation of Noncontracting
Ares 59
Chapter III. MHD Palletizing of Metals 61
1. MHD Methods of Controlling Disintegration of Free Electrically
Conducting Jets 61
2. Free Electrically Conducting Jets under the Effect of Si~
Variable Electromagnetic Forces in Sources 69
3. Disintegration Conditions of Free Jets Placed in a Longitudinal
Variable Magnetic Field 81
4. Conditions of Obtaining Spherical Particles 83
5. MHD Liquid Metal and Alloy Dispersers 90
Chapter IV. MHD Heat and Mass Transport in Induction Melting Furnaces 102
1. Eddy Nature of Electromagnetic Forces and Mechanisms of
Creating Unidirectional Motion 104
2. Pressure Developed by Active Sections of a Channel 108
3. Hydraulic Drag of Induction Channels 113
4. Simulation of Heat and Mass Transport Processes in Induction
Furnace Channels 115
5. Simulation of Industrial Induction Furnaces 134
Chapter V. Gas and Liquid Plug Flow MHD Devices 139
1. 1'hysical Essence of Processer in Accelerated Gas and Liquid 141
rlows
2. Formation of Plug Flows 144
3. Stable Forn~s of Gas-Liquid Tnterfaces. Mass Variation of
- Liquid Dose During Acceleration 150
= 4. Dispersion of Liquid Dose and Mass Transnort 158
5. Interphase Heat Exchange in Accelerator Channels 162
6. Model of the Motion of Liquid Doses of Variable Mass 167
7. Losses During Acceleration of a Plug Flow. Efficiency and
Other Energy Parameters of Plug Flow Accelerators 171
8. Example Calculation of a Device for Accelerating Plug Flows 174
Bibliography 182
COPYRIGHT: Izdatel'stvo "Naukova dumka", 1980
10845
CSO: 1861/195
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INDUSTRIAL TECI~TOLOGY
UDC 539.3+519.2
STRENGTH AND RELIABILITY OF TECHNICAL DEVICES
Kiev PROCHNOST' I NADEZHNOST' TEKHNICHESKIKH USTROYSTV in Russian 1981 (signed
to press 8 Jan 81) pp 2, 184-191
- [Annotation and abstracts of articles from collection Strength and Relia-
bility of Technical Devices", editor-in-chief V. S. Gudramovich, Institute
of Technical Mechanics, UkSSR Academy of Sciences, Izdatel'stvo "Naukova
dumka", 1300 copies, 192 pages]
~ [Text] Current problems of strength and rqliability of technical devices
are discussed. Data are presented on theoretical and experimental studies
of the load-bearing capacity and strength of thin-walled systems under static
and dynamic loads with consideration of plastic deformations and creep strains
with complex loading histories. Contact problems of the theory of.shells
are discussed as well as questions of optimum design of structural members.
Some problmes of the theor}? of accelerated tests are outlined. Some questions
of checking the working condition of technical devices are examined.
The collection is intended for engineers and researchers sp~cializing in the
field of strength and reliability, and also for undergraduate and graduate
students with the corresponding ma3ors in colleges and universities.
UDC 539.374
OSCILLATIONS OF SYSTEM WITH PHYSICALLY NONLINEAR CHARACTERISTICS UNDER PULSE
LOADING
[Abstract of article by Velichkin, V. A., Gudramovich, V. S., Konovalenko,
V. Ya., Makeyev, Ye. M., Pilipenko, V. V., Popov, A. I., Semenenko, V. P.
and Fediy, S. P.]
[~ext] An analysis is made of oscillations of a complex mechanical system
that consists of two masses with plates on springs fastened to an imponderable
hin~ed beam of f inite stiffness under pulsed action with consideration of
physically nonlinear properties of the spring material.
Accelerations are studied as a function of the magnitude and time of applica-
tion of pulses, ratio of stiffnesses and plastic properties of the springs,
and clearances between the plate and the beam.
55
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Calculated and experimental values of accelerations are compared.
Figures references 4 Russian.
UDC 519.2+624.011:539.4
PROBABILISTIC PROPERTIES OF FREQUENCIES AND REDUCED MASSES OF TRANSVERSE
OSCILLATIONS OF THIN-WALLED FLUID-FILLED CYLINDRICAL SHELL
[Abstract of article by Velichkin, V. A. and Fediy, S. P.J
[Text] The paper formulates the problem of finding the probabilistic charac-
_ teristics af a dynamic system by the method of statistical modeling.
An analysis is made of the probabilistic properties of frequencies and reduced
masses of transverse oscillations of a dynamic system based on the example
of a reinforced cylindrical shell filled with liquid.
The conclusions are illustrated by results of nwnerical calculation.
Figure 1, table 1, references 3 Russian.
UDC 539.384.6:624.074.4
STRESS-STRAIN STATE OF CYLINDRICAL SHELL UNDER LOADING VIA CIRCULAR ELASTIC
SADDLE SUPPORTS
[Abstract of article by Gayduchenko, A. P., Katan , L. I. and Makeyev, Ye. M.]
[Text] The authors give the results of an experimental study of the deformed
state ~f a cylindrical shell lying on elastic circular saddle supports of
finite dimensions. Calculated d3ta �ound by a previously developed method
are compared with experimental results, and an estimate is made of the influ-
ence that accounting for the interaction of support an~ she].1 has on the
stressed state of the shell in the vicinity of the support based on the exam-
ple of a tested structure.
Figures 3, table 1, references 6: 4 Russian, 1 Polish, 1 Western.
UDC 539.374
EXPERIMENTAL STUDY OF CYLINDRICAL SHELL LOAD-BEARING CAPACIT'Y UNDER LOCAL LOAD
[Abstract of article by Gerasimov, V. P., Gudramovich, V. S.]
[Text] The paper gives the results of experimental studies of t h e loa d-
bearing capacity of cylindrical shells deformed beyond the elastic limit under
the action of local loads of various classes applied to the surface of shells
by rigid saddle-support punches.
Figures S, tables 3, reference 1 Russian.
56
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UDC 539.374
EXPERIMENTAL STUDIES OF BAUSCHINGER EFFECT FOR AMG-6M ALUMINUM ALLOY
[Abstract of article by Konovalenkov, V.~S]
[Text] The author gives the results of experimental studies of the Bauschinger
effect for AMG-6M. The tests were done on specimens of circular cross section.
Results of processing experimental data are given in the form of tables and
a graph .
Figures 2, table 1, references 3 Russian.
UDC 539.384.6:624.072.4
EXPERIMENTAL STUDY OF DEFORMATION IN BULKHEADS REINFORCING CYLINDRICAL SHELL
UNDER TRANSVERSE LOCAL LOADING
[Abstract of article by Makeyev, Ye. M. and Semenenko, V. P.]
[Text] The paper gives the results of an experimental study of the deformed
state of circular bulkheads that reinforce a cylindrical shell under local
loads of various kinds with different laws of distribution in different com-
- binations with consideration of their deformation together with the shell.
Calculated data are compared with experimental results, and an analysis is
made of the effect of some simplifying assumptions that are generally made
in approximate solutions.
Figures 4, tables 4, references 4 Russian.
~ UDC 624.074.4
- M OD ~L OF MATCHING REQUIREMENTS IN MULTICRITERIAL STRTICTURAL DESIGN PROBLEMS
[Abstract of article by P~chtman, Yu. M. and Skalozub, V. V.]
[Text] The authors discuss the feasibility of using an n-person game model
with a considerable set of criteria in multicriterial problems of cptimizing
structural members. It is notec; that the model permits simultaneous analysis
~ of stages of "external" and "internal" design; some conditions are presented
that are sufficient for constructing a minimax solution with required proper-
ties. A technique is given for reducing certain problems of optimizing beams,
plates and shells to game problems. Numerical examples are given.
Figure 1, table 1, references 10 Russian.
57
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UDC 539.376
~ STABiLIT;t OF STRUCTURALLX ORTHOTROPIC SHELLS UNDER CREEP CONDITIONS
[Abstract of article by Poshivalov, V. P.]
- [Text] The author derives equations of stability of structurally orthotropic
shells under conditions of creep. It is assumed that during creep the stresses
and strains in the shell differ little from the stresses of the basic zero-
moment state. Consideration is taken of the b iaxial stressed state of the
shell and the uniaxial stressed state of the reinforcing elements.
References 4 Russian.
UDC 539.3
- STRESSES IN REGION OF JOINING BETWEEN PIPE AND CYLINDRICAL SHELL
[Abstract of article by Sel'skiy, Yu. S.]
: [Text] An examination is made o~ the stressed state of a joint for small
ratios of pipe and shell radii.
Coefficients of concentration are obtained for cases of internal pressure,
uniform temperature differential, and also the longitudinal force and torques
applied to the pipe.
It is proposed that a ray method be used for problems with arbitrary boundary
shape (as applied to intersections at an arb itrary angle with small ratios
of radii).
Figures 4, tables 2, references 10: 8 Russian, 2 Western.
~ UDC 539.3
PRINCIPAL EQUATIONS IN THEORY OF SHELLS WITH DIFFERENT MODULI IN TENSION AND
COMPRESSION
[Abstract of article by Tamurov, N. G. and Turovtsev, G. V.]
[Text] The paper proposes a model for a material that has ciifferent elastic
characteristics in compression and tension. Without discus:;ing the details
of deriving the model, the authors find the principal equat~.ons of a theory
of thin shells in the geometrically nonlinear approximation that are converted
to the corresponding classical expressions in the case of a single-modulus
material. A method is outlined for solving problems with tY~e proposed physical
re].ations . �
. References 7: 5 Russian, 2 Western.
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UDC 539.384.6:624.072.4
DESIGNING REINFORCEMENT RINGS WITH MAXIMUM BENDING STIFFNESS
[Abstract of article by Tkacheva, T. V.]
[Text] The Prager optimality criterion is used to solve the problem of opti-
mizing the cross sectional shapes of rings that reinforce a cylindrical shell
under loading by a concentrated transverse force. The optimum area of the
supporting layers of a sandwich ring is determined. An example of calculation
is given.
Figures 3, references 6: 4 Russian, 2 Western.
UDC 624.072.4
~PROPERTIES OF OPTIMUM RING FRAME DESIGNS UNDER COMBINED FORCE AND TII~IPERATURE
LOADING
[Abstract of article by Binkevicti, Ye. V. and Dzyuba, A. P.]
[Text] The problem of designing nonuniformly heated ring frames loaded by
concentrated forces is numerically solved on the basis of a maximum principle
with consideration of limitations on deviation of the shape of the axis from
circular, and also strength and structural deviations. The properties of
optimum designs are studied. An analysis is made of the particulars of using
necessary conditions of optimality in design. Results of digital computer
calculations.are given. Figures 4, table I, references 3 Russian.
UDC 539.374
DYNAMIC BEHAVIOR OF PLASTIC SPHERICAL SHELL UNDER REPEATED PULSED LOADING
[Abstract of article by Gudramovich, V. S. and Shatsillo, S. I~]
[Text] A formulation is given for the problem o~ dynamic analysis of the
- motion of a plastic spherical shell under repeated pulsed loading. The factor
that accounts for the nonuniforntity of properties that is acquired during
preceding loading is the fields of residual stresses. The finite element
method is used for the snalysis; the finite elements are selected in the form
of frusta of cones. An examination is made of the problem of motion of an
initially homogeneous shell with pulsed loading on the basis of isotropic
flow theory. References 14: 12 Russian, 2 Western.
UDC 539.3
CYL INDRICAL SHELL BEHAVIOR WITH DYNAMICALLY APPLIED NONUNIFORM EXTERNAL PRESSURE
[Abstract of article by Makarenko, A. D. and Makarenko, N. 8.]
[Text] The authors consider the problem of stability of a cylindrica~. shell
with dynamic application of nonaxisymmetric external pressure. Results of
59
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calculat~ons are given that were obtained with the use of a specially compiled
algorithm enabling inve:3tigation of the process of deformation and loss of
stabilit~ of shells wi*.n respect to space and time coordinates. Figures 3,
references 3 Russian.
~ UDC 519.248:62-192
DETERMINING INVARIANCE IN ACCELERATED TEST THEORY
[Abstract of article by Avramenko, V. I.]
[Text] A definition of invariance of a factory that produces goods by batches
- is proposed on the basis of a statistical p'robability approach that is different
from the existing definition. A def inition of invariance of the properties
of items relative to a load is introduced. An illustration is given of the
feasibility of using these concepts to solve problems of the theory of accel-
erated tests. References 4 Russian.
UDC 519.2.48:62-192
MATHEMATICA:~ MODEL OF SYSTEM DEGRADATION
[Abstract of article by Belosvetov, S. A. and Dyachenko, V. Ya.]
[Text] Based .on an approach to systems analysis as comprising a paired random
process and level, the authors propose mathematical models of processes of
structural and phase degradation of a system as a basis for developing algo-
rittzms and for machine simulation of real degradation processes to evaluate
~ the reliability of a system. Reference 1 Russian.
UDC 629.4:519.2-192
- DEFINITION OF RATIONAL MAINTENANCE SYSTEM FOR ~;OMPONENTS OF TRACTION UNITS
[Abstract of article by Bosov, A. A. and Khandriga, A. G.]
[TextJ A method is proposed for determining the periods of renovating compo-
nents of traction units. An example is given af designing and setting up
a system for maintenance of the DT-9N traction engine. Figures 5, tables
4, references 2 Russian.
UDC 519.24
SOME PARTICULARS OF USING STANDARDIZATION IN COMPUTERIZING REGRESSION ANALYSIS
ALGORITHMS
[Abstract of article by Dolgiy, V. I., Perlik, V. I. and Sokolov, A. S.] .
[Text] An examination is made of the effect of the sequence of linearizing
and standardizing variables in regression analysis algorithms. It is shown
- 60
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that when constructing incomplete second-order models, algorithms based on
standardizing variables before linearizing are more correct in all cases.
Table 1, references 4 Russian.
UDC 539.2
EVALUATING DURABILITY OF MATERIALS UNDER AGING CONDITIONS
[Abstract of article by Pereverzev, Ye. S.]
[TextJ Approximate analytical expressions are given for evaluating the dura-
bility of materials under aging conditions. Approximate methods are considered
for taking account of damage accumulation due to mechanical loading and natural
~ aging. References 4 Russian.
UDC 621.3
RELATION BETWEEN THERMODYNAMIC AND STATISTICAL EQUIVALENCE PRINCIPLES
[Abstract of article by Pereverzev, Ye. S.]
[Text] Based on a thermodynamic approach, conditions are found for which
the thermodynamic principle of equivalence is identical to the statistical
method of equal probabilities. It is shown that in this case the failure
rate is proportional to the rate of increase in entropy. References 4 Russian.
UDC 519.281
STATISTICAL ESTIMATION OF TECHNICAL SYSTEM RELIABILITY INDICES BY EFFICIENCY
FUNCTION METHOD
[Abstract of article by Perlik, V. I.]
[Text] The author considers the statistical aspects of the method of efficiency
fucntions; the main principle~ of the method have been formulated in previous
papers by this author. First an investigation is made of the theoretical
sequence of statistical estimation of the index of reliability of a system,
which in principle enables derivation of the exact solution. Practical imple-
mentation of this sequence is difficult because of insurmountable mathematical
difficulties. The process of approximate statistical estimation recommended
in the article is based on a moment approach. A detailed er.ami.nation is made
of problems involved in getting the ini~ial information. REeferences 3 Russian.
UDC 519.24
STATISTICAL METHOD OF EVALUATING SCALE FACTOR FOR ROCK STRENGTH PROPERTIES
[Abstract of article by Rubets, G. T.]
[Text] Various generalizations of the W eib ull theory of. strength are
examined on the basis of the theory of distributions of ordFred statistics.
6~
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It is shown how these generalizations can be applied to estimation of the
scale factor of rock strength characteristics, as well as those of structural
components in underground structures. References 10: 5 Russian, 5 Western.
~ UDC 519.248:62-192
MODEL TO ACCOUNT FOR A PRIORI INFORMATION IN ACCELERATED TEST PROBLEM
[Abstract of article by Rybalka, K. P.]
[Text] The author examines a method of determining how the ma}hematical ex-
pectation of time to failure (M~) depends on the load vector (e) from results
of accelerated tests using a priori inform~~ion on the distributi~n law for
time to failure and the function of change in M~ as dependent on e. Refer-
ences 3 Russian.
UDC 621:004.17:620.199
BAYES MODEL FOR SEQUENTIAL EVALUATION OF TECHNICAL SYSTEM RELIABILITY INDEX
[Abstract of article by Savchuk, V. P.]
[Text] A mathematical model is proposed for evaluating the lawer confidence
limit of the reliability index for predetermined confidence coefficient. The
model uses a Bayes approach. Inversion of the model enables planning of the
number of demonstration tests when confirming reliability. It is shown that
carrying out tests of long duration considerably reduces the number of test
specimens as compared with the conventional method. Tab1e 1, references 3
Russian.
UDC 621.3.019.3:519
DETERMINING RELIABILITY REQUIREMENTS FOR ITEMS FROM PREDETERMINED LEVEL OF
SYSTEM RELIABILITY
[Abstract of article by Skripnik, V. M. and Spirkov, S. N.]
[Text] An expression is derived for determining guaranteed up times of items
in a complex system with predetermined system reliability requirement when
the distribution of service lives of the items conforms to Weibull law.
_ Figure 1, references 3 Russian.
UDC 620.171.311.2-192
POSSIBLE METAODS OF RECALCULATING RELIABILITY CHARACTERISTICS UNDER DIFFERENT
LOADING CONDITIONS
[Abstract of article by Stepanov, V. V.]
[Text] The author considers some possible methods of recalculating reliability
characteristics of technical devicas under different loadin~ conditions. A
62
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relationship is established between the proposed formulas and known principles
of reliability theory. Figures 2, references 5 Russian.
UDC 62:519.2
READINESS FACTOR AND OVERHEAD COST OF TECHNICAL DEVICE FOR COMBINED MONITOR~NG
AND VARIOUS RATES OF RESTORING OPERABILITY
[Abstract of article by Chumakov, L. D.]
[Text] Analytical expressions are found for the readiness factor, operating
costs and average number of restorations of operability in the case of expo-
nential distribution of the time of fail-free operation and restoration.
References 3 Russian.
UDC 62-192:539.3 ,
llURABILITY EVALUATION METHOD USING SEMI-MARKOV MODEL
[Abstract of article by Shiyan, 0. V.]
[Text] An examination is made of a method of accelerated tests the stepwise
loading method. A method is proposed for estimating the parameters of the
fa ilure distribution function under operating conditionQ using the Korolyuk ,
theorem of inean so~ourn time of a semi-Markov process in a fixed set of states
and the principle of maximum likelihood. References 3 Russian.
UDC 62:519.2
EVALUATING DURABILITY OF' TECHNICAL DEVICE WITH PERIODIC WORKABILITY CHECK
[Abstract of article by Chumakov, L. D.]
[Text] An estimate is obtained for the durability of a technical device with
periodic monitoring of its working state when the mathematical expectation
of the time of fail-free operation is known. References 3 Russian.
COPYRIGHT: Izdatel'stvo "Naukova dumka", 1981
6610
CSO: 1861/189
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TURBINE AND ENGINE DESIGN
UDC 621.165000.5
OPTIMAL LAST STAGE DESIGN OF HIGH-POWER STEAM TURBINES
Kiev OPTIMAL'NOYE �ROYEKTIROVANIYE POSLEDNEY STUPENI MOSH(~iNYKH PAROVYKH TURBIN
in Russian 1980 (sign~~i to press 7 Jul 80) pp 2, 226-227
[Annotation and table of contents from book "Optimal Last Stage Design of High-
Power Steam T~bines", by Leonid Aleksar_drovich Shubenko-Shubin, Anatoliy
Alekseyevich Tarelin and Yuriy Petrovtch Antiptsev, Ukrainian SSR Academy of Sciences,
Proble~ of Machine Building Institute,Izdatel'stv~ "Haukova dumka", 1000 copies,

_n1 228 pages ]
[Text] This monograph discusses the problems of optimal last-stage design of high-
power steam turbines. A study is made of the physicomathematical and engineering
principles of optimal turbine stage design, optimization of the thermal gas dynamic
process and blade design. The follo-+ing are presented: a>mathematical model of the
thermal gas dynamic process and its investigation using the extsting mathematical
methods of finding optimal solutions and modern computer engineeriag means. The
methods of constructing and optimizing the blade cas.cades are presented which permit
economical profiles to be obtained in a wide range of variatiQn of the initial
specifications and geometric characteristics. A formalized blade design process is
proposed which offers the possibility of using simp.~e methods of optimizing various
- types of designs which are convenient in practical application.
The book is designed for scientific and engineering-technic~ personnel specializing
in turbine building.
There are 54 illustrations, 3 tables, and 59 references on pages 222-225.
Table of Contents
3
" Foreword
Basic Symbols used in the Text ~
~ Chapter I. Problems of Optimal Last Stage Design of Pawer Turbines 9
1. Last Stage in Power Turbine Design System 9
2. Last Stage Operating Quality Criteria 13
3. Structural Elements of the Last Stage in its Design System 16
. 64
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Chapter II. Primary Problems of Constructing a Last Stage Mathematical
Model 23
1. Statement of Problem of Optimizing Flow Characteristics in Stage
Clearances 23
2. Fdrmation of the Assignment and Restrictions on Optimal Blade Des~gn 31.
3. General Structure of Optimal Blade Design System 37
4. Problems of Selected Axial Clearance and its Outlines 42
Chapter III. Mathematical Model of Thermal Gas Dynamic Process and
Optimization Methods 47 ~
1. State of the Art of the Problem 47
2. Basir Propositions and System of Flow Equations 50
initial Specifications and Qua.lity Criteria 55
4. Optimization and Limitation Parameters 58
- 5. Meth~ds of Solving the Synthesis and Analysis Problems 60
Chapter IV. Analytical Method of Optimizing Stage Kinematic Characteristics 63
1. Optimization by Quality Criterion Integral Stage Efficiency 63
2. Optimization by Quality Criterion Integral Sta~e Energy Loss
with Output Speed 76
3. Maximum Integral Stage Efficiency with Minimum Integral Energy
Loss with Output Speed 83
4. Some Peculiarit~~s of Considering Restrictions 8~
S. Study of the Basic Laws of Blade ~ist 95
Chapter V. Optimization of the Stage Parameters by Computer Simulation 110
1. Simulation Model of Thermal Gas Dynamic Process of a Stage 112
2. Study Performed by the Scan Method and a Procedure for Processing
the Results 117
3. Optimization of Stage Economy by the Random Search Method 127.
Chapter VI. Designing the Blade Cascades of Tu~bomachines 1~
1. Basic Principles, Initial I3ata, Parameters, Restrictions 1~
2. Simplest nao-Parametric Profile 138
3. Profiles wit~ Given Configuration of trie Output Part of the Outer
Profile ~ 146
4. Profiles with In~.reased Strength Characteristics 156
5. Optimization of Blade Cascades 16Q
6. Blade Cascade Design Considering Peculiarities of Actual Blade
Engineering 166
- Chapter VII. Engineering of the Working Part of Blades 175
1. Basic Principles, Ir~ztial Data, Restrictions 175
2. Analytical Method of Shaping the Blade 185
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3. Numerical Methods of Forming the Blade Surface 196
4. Optimization of Blade Designs 206
5. Research~Results 212
~ibliography' 222
COPYRIGHT: Izdatel'stvo "Naukova d~ka", 1980
- 10845
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NAVIGATION AND GUIDANCE SYSTEMS
- UDC 53i.55:521.1
CORRECTING INERTIAL GUIDANCE SYSTEMS BY USING COMBINED SUBSIDIARY POSITIONAL
AND VELOCITY INFORMATION
Moscow IZVESTIYA AKADEMII NAUK SSSR: MEKHANIKA TVERDOGO TELA in Russian
No 5, Sep-Oct 81 (manuscript received 9 Oct 79) pp 12-19
[Article by V. I. Kalenova, V. M. Morozov, N. A. Parusnikov and A. G.
Shakot'ko, Moscow]
[Text] An investigation is made of thQ problem of correct-
ing an inertial guidance system with combined utilization
of redundant positional and velocity information. In con-
nection with the solution of this problem the authors discuss
one of the possible correcting algorithms that realize decom-
position of the problem with respect to components of the
correction vector.
In present-day navigational systems that are based on the
inertial guidance method, the inertial information is sup-
plemented by information of noninertial nature to improve
accuracy properties. The problem of correcting inertial
guidance systems by using redundant information can be formu-
lated as a problem of estimating the vector of stat:e of
a linear system from given measurements [Ref. 1, 2~.
To evaluate the possibilities of correcting such sqstems
with respect to subsidiary information of a given type,
the linear theory of observability should be used [Ref.
3, 4]. Analysis of the observability of the system enables
discrimination of the variables that can be evaluated, de- .
- termination of the limiting accuracy of evaluation of such
variables, and also construction of a class of suitable
algorithms of evaluation that are subject to further detailed
study. In addition, analysis of observability enables esti-
mation of errors due to various simplifications in construct-
ing the correction algorithm.
In Ref. 2, 5 an analysis was made of observabili.ty, and
some correction algorithms were constructed for inertial
guidance systems using either positional or velocit:y subsidiary
67
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information. We discuss below some modifications of combined
use of this information, and propose corresponding estimation
algoritYuns. In doing so, we use one of the methods of de-
composing the observed space with respect to the components
of the correction vector.
1. Consider the linear system
d~/dt=G~-I-q, a=1f~+r (1.1) ,
- Here ~ is an n-dimensional vector of state, 6 is an s-dimensional vector of
measurement, G, H are fixed matrices of the corresponding dimensionality,
q, r are random vector processes of the white noise type with given intensi-
ties. It is assumed that the pair (G, H) is observable. It is required to
get an estimate of quantity ~ by measurement cr on segment [to, t]. ~
In constructing practical estimation algorithms, it is often advisable to
decompose (split) the estimation problem with respect to the components of
the measurement vector so that this problem is reduced to a series of sub-
problems with scalar measurements. Generally speaking, the decomposition
can be carried out in different ways [Ref. 6J. Let us consider the problem
of decomposition for a two-component measurement vector ~_(Q1, c~2)T, having
in mind a situation where the time interval on which information ctl and cr2 is
used in con~unction is preceded by a long time interval with arrival of only
information Q1.
Let Q~=h,''~, Q,=Iz,''~ and let {g}, {y} be spaces observable by using measurements
61 and o2 respectively.
We use the notation dim =n, rl im {,Y} =m, dim {Y} =l. Obviously tn~n, l`n,
m-I-l>rz. Let us also introduce the notation
_ ~,=h,, nz~G''a.~, . . . , ai+,=G''ai, . . . (1.2)
G,=hZ, bZ=Gr6~,... ~ bf}1=G''bi,
The vector x=(x,.r.Z....r.m)'' made up of components observable by means of quantity
~1 takes the form
x=Lx~, L==C, (a,a.z . . . am)'' (1.3)
where C1 is some nondegenerate square matrix whose selection is not directly
related to the decomposition problem. In virtue of the linear dependence
of vectors a,, a2, am, am+, we have the representation
am+~=f},a,-~flZaZ-f- . . . -f-fl~~~am (1.4)
Here the Ai are constant coefficients. From (1.1)-(1.4) we get
- dx / dt=Gsx+qr, Q,=gx''x+rX (1.6)
0 1 0 0
~x- C19~i ~ 0 0 1... 0
qx = I,xq 9 = . . . . . . .
gxT =(10. 0) Cil 0 0 0... i
fll 9~ ~a...$m
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The standard algorithan that yields estimate x� of vector x is described by
the equation
dx� / dt=G=x�-I-If:(Q,-g=Tx�) (1.G)
Vector KX is unambiguously determined either by the Kalman optimum filtratidri
method, or from the condition of assignment of roots of characteristic equa-
~ tion ~ pE-G~-I-K=ga'' ~=0. At suff iciently low intensity of the measurement error
rX, the initial condition ~(t,) =xo� should be given as ~o�=C, (Q�00 0)''.
- Let us introduce vector z=(z,z,... Zn-m)r, n-m