SOVIET ATOMIC ENERGY VOL. 56, NO. 6

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• Document Creation Date: December 27, 2016
• Document Release Date: September 14, 2013
• Sequence Number: 6
• Publication Date: June 1, 1984
• Content Type: REPORT

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< 'ISSN 0038-531X
Russian Original Vol 56, No 6, June, 1984
-Decem ber; _1984
? ATp111Hill, amprwa_
-(ATOIVINAYA, gNERGIYA)
TRANSLATED FROM RUSSIAN
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SOVIET
ATOMIC
ENERGY
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SOVIET ATOMIC ENERGY
A translation of Atomnaya Energiya
December, 1984
'Volume 56, Number 6 June, 1984
CONTENTS
State-of-the-Art and Development Prospects for Nuclear Power Stations
Containing Pressurized-Water Reactors (VVER) ? G. A. Shasharin,
E. I. Ignatenko, and V. M. Boldyrev
State-of-the Art and Development Prospects for Nuclear Power Stations
Containing RBMK Reactors ? E. V. Kulikov
State-of-the Art and Development Prospects for Nuclear Power Stations
Containing Fast Reactors ? O. D. Kazachkovskii . . ? ? O ...
Engl./Russ.
361 353
368 359
375 365
PAGES OF HISTORY
Critical Assembly of the World's First Nuclear Power Station
? M. E. Minashin
381
382
ARTICLES -
Heat Accumulators at Nuclear Power Stations --V. M. Chakhovskii . ? ?
?
388
389
A Load-Following Atomic Heat and Power Plant --V. M. Boldyrev
and V. P. Lozgachev
397
396
Computer-Assisted Radiation Tomography of Spherical Fuel Elements
? E. Yu. Vasilleva, L. I. Kosarev, N. R. Kuzelev, A. N. Maiorov,
and A. S. Shtanl
402
400
Physicochemical Approach to the Description of the Distribution
of Macroquantities of Pu(IV) in Extraction by Tributylphosphate
from Nitrate Solutions in the Presence of Complex Formers
Applicable to the Regeneration of Spent Nuclear Fuel from Fast
Reactors ? A. S. Solovkin and V. N. Rubisov
410
406
LETTERS TO THE EDITOR
Contribution of Nuclear Interactions to the Distribution of Absorbed
? Energy in Thin Plates Bombarded with Fast Charged Particles
? S. G. Andreev, I. M. Dmitrievskii, and I. K. Khvostunov
- 418
413
Stability of Scintillation Detectors Vis-A-Vis y Radiation
? V. V. Pomerantsev, I. B. Gagauz, Yu. A. Tsirlin,
and O. V. Levchina
421
415
Radiation Stability of Scintillating Polystyrene ? I. B. Gagauz,
A. P. Meshman, V. F. Pererva, V. V. Pomerantsev,
and V. M. Solomonov
423
416
Yield of Electron Bremsstrahlung from Thick Targets ? V. I. Isaev
and V. P. Kovalev
425
417
The Russian press date (podpisano k pechati) of this issue was 5/24/1984.
Publication therefore did not occur prior to this date, but must be assumed
to have taken place reasonably soon thereafter.
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STATE-OF-THE-ART AND DEVELOPMENT PROSPECTS FOR NUCLEAR POWER STATIONS
CONTAINING PRESSURIZED-WATER REACTORS (VVER)
G. A. Shasharin, E. I. Ignatenko, UDC 621.311.2:621.039:621.039.57
and V. M. Boldyrev
At present, nuclear stations containing pressurized-water reactors PWR such as the VVER
have become most common in nuclear engineering in the USSR and elsewhere. In all, there are
27 stations containing VVEE operating with unit electrical power levels from 70 to 1000 MW
and total installed power of 12.14 GW (Tables 1 and 2) [1].
ATTAINMENT OF DESIGN PARAMETERS, OPERATION, AND REPAIR
Table 3 shows that stations containing VVO reactors work reliably and consistently.
The power utilization factor (PUF) in these VVER units, on the whole, is higher than in for-
eign ones (Table 4) and in some cases attains 90-96%. In the main units, the PUF increase
steadily as the working time increases at first and attain stable values in the third year,
while in later units they stabilize in the second year. There have been substantial reduc-
tions in PUF in individual years for certain units because of long downtimes due to accident
situations or control constraints. There are also differences in internal electricity consump-
tion at units of the same type because of differences in output, load level, and local work-
ing conditions. When a station is operating stably, the internal use varies around a certain
level, but all the parameters deteriorate when there are prolonged downtimes. The efficiency
is dependent mainly on local conditions, as well as on the state of the equipment. As a rule,
the efficiency increases after maintenance, condenser cleaning, the replacement of worn equip-
ment, and other such measures.
The Ministry of Energy of the USSR operates a system for planned prophylactic mainte-
nance (PPM) at nuclear stations, which includes periodic major, medium, and current repairs,
whose sequence and duration are determined by the planned maintenance cycle. Each year,
each unit is shut down for major or medium overhaul. The fule is also changed during this
period. Working results show that the times taken in planned maintenance correspond to the
standards (Table 5). Information is not given on nuclear stations containing VVER-1000 be-
cause they have not been operating long.
One of the obvious factors resulting in improved economics at nuclear stations is the
reduction in the equipment upgrading time in annual PPM. This is dependent primarily on the
organization of the operations, the staff qualification, and the equipment of the station
with the necessary servicing facilities. The main operations in PPM are devoted to the equip-
ment in the first and second loops (Table 6).
About 280 items of equipment and apparatus are used in maintaining the nuclear steam-
producing plant at a nuclear station containing two VViR-440, which means that it is impor-
tant to upgrade the maintenance operations on the equipment in the first loop. The following
are required for high-grade and rapid upgrading and servicing:
1) a set of standardization documents laying down the specifications for checking the
metal in nuclear station equipment, including the corresponding methods of checking the metal;
2) a set of means of checking the metal for use with those units and components whose
checking is laid down by the standardization documents;
3) criteria for accepting or rejecting defects found in the metal;
4) methods (technologies) for repairing defects in the equipment; and
5) repair equipment.
Translated from Atomnaya Energiya, Vol. 56, No. 6, pp. 353-359, June, 1984.
0038-531X/84/5606-0361$08.50 ? 1984 Plenum Publishing Corporation
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TABLE 1. Nuclear Units Containing VVER
Reactors as of Jan. 1, 1984
Nuclear
station
Number and
type of tur-
bines
I
Installed elec-
tric power, MW
Installed
district heat-
tapoff, Gcal/h
Connec-
tion to
grid
Novyi Voronezh
first unit
3 A1-70-11, 210
3 TVF -100-2
IX.19611
second
5 It-75-30, 365
5 TVF -100-2
XII.1969
third
2 H-220-44, 417
2 TVV-220-2
2x25
XII.1971
fourth
21c-220-44, 417
TVV -220-2,
TVV -22(1-2A
2 x 25
XII.1972
fifth
2 H-500-60/1500 1000
2 x 30
V.1980
Kola
first unit
2H-22)-44, 440
2 TVV -220-2A
2 x 25
VI.1973
second
The same 440
2 x 25
XII.1974
third
2 1c-220-44-3, 440
2 TVV -220-2A
2 x 50
111.1981
Armenian-
first
2 1-220-44, 407,5
2 TVV -220-2A
2X25
XII.1976
seCond
The same 407,5
2x25
1.1980
Rovensk
first
2 1-220-44-3, 392
2 1VV. 220-2AUZ
2 X 50
XI1.1980
second
The same 416
2X50
XII.1981
South Ukranian
first unit
It-1000-60/1500, 1000
TVV-1000-4
200
XII.1982
Reinsberg (GDR)
_._ 70
?
V.196?
Nord
first unit ?
21-22044-3, 440
2rjv .220-2AJZ
2 X 50
X11.1972
second
The.same 440
2x50
XII.1974
third
4 ? 44')
2x50
XI.1977
fourth . ,
Kozlodui (Bulgaria)
first unit
? ? 440
2 It-220-44-3, 440
2TVV 220-2AUZ
2 x 50
2 x 50
VI I I .197E
m ,,s,.,
v I?luiq
second
The same 440
2x50
XIII.1975
third
r ? 440
2x50
X11.198(
fourth ',
? ? 440
2 X 50
I V .1982
Lovisa (Finland)
first unit
2 H-220-44-3, 440
2TVV 220-2ADZ
2 x 50
11.1977
second
The same 440
2 X 50
X1.198(
Bogunice
(Czechoslovakia)
? first unit
2 It-220-44-3, 440
2 TVV-220-2AUZ
2x50
XII.197E
secind
The same 440
2x50
111198(
Faits (Hungary)
first unit
2 1-220-44-3, 440
2 TVV 220-2AUZ
2 x 50
X I 1.1982
TABLE 2. Basic Technical Characteristics
of Reactor Systems Containing VVER
Characteristic
o
c_
VViR-210 I
in
>
o
o
o
Reactor thermal .
power, MW
265
76,1
1320
1375
30,0
Number of circulation
loops
3
6
8
6
4
Pressure, MPa;
in reactor
9,8
9,8
10,3
12,3
15,7
in steam generators
3,1
3,1
3,2
4,6
6,3
Temp., t
at reactor inlet
250
245
248
268
288
at reactor outlet
266
266
274
296
317
Coolant flow through
reactor, m3/h
160)0
33.00
50,00*
45900
88000
Internal diam. of re-
actor body, mm
2640
3560
3561
3560
4139
Core
equiv. diam., mm
1900
2880
2880
2880
3110
height in working
staft, mm
power density, kW/
liter
2500
38
2500
47
2460
83
2460
86
3560
111
Number, of fuel as-
semblies in core
148
343
349
349
151
Number of fuel pins
90
90
126
126
317
Fuel pips:
outside diameter,
mm
thicluiess of Zr + 1%
10,2
0,6
10,2
0,6
9,1
0,65
9,1
0,65
9,1
0,67 .
Nb sheath, mm
mean linear power.
80
99
122
127
176
W/cm
Uranium:
loaded into reactor
17,0
40,0
41,5
41,5
66,0
specific power, kW/
kg U
enrichment in new
pins on replacing 1
of assemblies, 5
mean burnup, MW ?
day/kg
tkrumber of CPS units
15,5
2,0
13
19
19
2,0
14
37
32
3,0
28
73
33
3,5
73/37
45,5
3,3/4,4
27/40
1.,9
Unit efficiency. To .,
26,5
27,7
27,7
32,0
33,3
*With seven loops working (one loop reserve).
Up to now, the monitoring of metal at nuclear power stations has been undertaken on in-
dividual programs agreed annually with the corresponding organizations. From July 1983,
unified instructions apply for the monitoring of the state of the main metal and welded joints
in,equipment and pipelines in the first and second loops in nuclear power stations containing
VVER, which make the fullest use of experience with metal monitoring at existing nuclear sta-
362
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TABLE 3. Unit Parameters in Nuclear Sta-
tions Containing VViR
Nuclear
station
Power pro-
duction,
billion
kW 'h
Power
use fac-.
Dar, To
.
Internal
power
consump-
on To
Net unit
eL. To
Cl
CO
CO
.4.4
CO
00
CO
.....
CO
CO
03
CO
00
CO
Cl
CO
CO
CI
CO
CA
Cl
CO
CC
.c-c
CO
00
CD
...1
Novyi Vor-
onezh:
first unit
1,591
1,602
86,5
87,1
6,96
6,57
25,61
25,29
second
2,8
2,851
87,6
89,1
6,53
6,04
25,98
25,36
third ,
fourth
3,012
3,057
2,682
3,219
82,5
83,7
73,4
88,1
8 ,02
8,49
8,03
7,96
15,94
26,22
25,82
26,12
fifth
5,338
7,085
60,9
80,9
5,59
4 ,76
29,77
29,94
Kola:
first unit
3,080
3,471
79,9
93,0
7,53
6,67
28,87
29,16
second
2,6483,311
68,7
86,9
6,99
6,60
29,09
29,14
third
2,049
2,632
53,2
68,3
7,72
5,95
28,3r
29,82
Armenian
first unit
2,304
0,873
64,6
24,4
7,8
12,1(
26,48
25,02
second
2,179
3,073
61,0
86,1
8,2
8,32
26,25
26,01
Rovensk
first unit.
1,895
2,233
55,2
65 ,0
8,93
8,92
25,89
25,53
second
2,266
2,018
62,2
55,4
8,96
8,40
24,44
15,81
South
Ukrainian
first unit
-
2,763
-
31,5
-
4,76
-
-
TABLE 4, PUF for Foreign Nuclear Stations
Containing PWR, % [2]
Region
1980-
All countries apart
from Comecon
members
59,3
1 1981 1982
61,9 1 60,0
1983
(first half
year)
59,2
tions.* These instructions cover not only the volume'of work and the periodicity in monitoring;
the individual units and components in the two loops but also enumerate the main standardiza-
tion documents on metal monitoring, including methods of metal monitoring for various pur-
poses (ultrasonic, visual, magnetit-powder, color, etc.), in addition to the monitoring fa-
cilities and the basic criteria or standards for evaluating metal state.
One mainly uses manual monitoring facilities at nuclear power stations, although this
tends to lead to high staff doses. Future facilities should be highly specific, for example
for monitoring the metal in the reactor body, the collectors and tube bundles in the steam
generators, the volume compensators, etc. Here we may note developments in this country (at
the All-Union Nuclear Power Station Research Institute) and in Czechoslovakia, including the
use of the miniature Prognoz-11 TV system for monitoring inaccessible locations, the UNIKOP
specialized remote-sensing systems for monitoring steam generators, the KONAP units for check-
ing reactor piping, and the UKOZ for monitoring volume compensators.
The metal defects at our nuclear power stations have the following percent distribution
by cause: constructional 20, technological 40, metallurgical 16; installation 14, and oper-
ation 10.
There are ongoing studies designed to improve the reliability,and safety, particularly
on the basis of studies of failures. Long-term experience with VVER-440 reactors is con-
firmed by data on PWR stations from other countries indicates that failures (defects) in
equipment are distributed in the following percent proportions: first-circuit equipment
*In January 1984, the unified instructions on operational monitoring of the state of the
malm metal and welded joints in equipment and pipelines in nuclear power stations containing
VVER-1000 reactors were confirmed and put into operation.
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TABLE 5. Durations, of Downtimes for One
Unit Containing VVER-440 for Fuel Chang-
ing and Equipment Upgrading
Maintenance
Nuclear station
Novyi
oronezh
Kola
Arme-
nian
Average from standards:
30
30
30
actual average
34
40
39
actual minitnal
18
24
30
Major from standards
55
55
55
actual average
66
66
52
actual minimal
53
45
45
TABLE 6. Structure of the Costs for PPM
in the Volume of Major Equipment Mainte-
nance in the Main Sections of Nuclear
Power Stations
Section
Proportion of
costs in overall
maintenance
costs. To
Proportion of
total labor
cost. To
Reactor
Turbine
Electrical
Thermal automatics and
measurement
Others
40-46
25-30
6-8
7-10
4-5
45-55
18-20
8-10
10-15
2-3
15-20, turbines 25-30, electrotechnical and conversion equipment 35-45, and auxiliary thermo -
mechanical equipment and pipelines 15-20. In 15-20% of the cases, the faults are due to
staff errors. One unit on average show 20-25 faults a year leading to power loss. For ex-
ample, in 1981 this power loss from these causes was about 2% at nuclear stations in this
country.
The most characteristic faults are leaks in condenser pipes, sealing failures in the
high-pressure and low-pressure heaters, unsatisfactory operation (overheating or sparking)
in the brush apparatus in turbine generators, sealing failures in the pipelines and other
equipment in the second loop, failures in steam generator pipes, and incorrect operation of
protection and interlock equipment.
The All-Union Nuclear Power Stations Research Institute has set up a system for acquir-
ing and processing information on these failures, and this has been accompanied by an analysis
of the earlier stages of operation, with the result that the design of power stations and
equipment for them has been improved in many ways, which have substantially improved the re-
liability, where we may particularly note the following: more defined welding in steam-
generator collectors for the water-steam interface section to protect it from corrosion; a
modified throttling control for K-220-44 turbines and changes in the design of the high-
pressure heaters; design changes in the SPP -220; and anticorrosion coating in the volume
compensator to reduce corrosion.
The radiation backgrounds at operating power stations have been maintained within the
limits set down by the standards for staff safety in any operations. The overall specific
activity of water in the first circuit due to fission products is not more than 10-4-10-3
Ci/liter (1 Ci = 3.7 x 1010 Bq). The low activity level and the good sealing in the first
circuit have meant that the activity levels in the air in the sealed spaces are low. The
mean annual collective dose to staff in systems containing VVER-440 reactors has been 362
ber (1 ber = 0.01 Sv), the average number of staff being 380. The fluctuations in annual
dose are due to differences in the volume of servicing operations. The annual collective
dose in the normal operation of a VVER -440 unit together with planned preventive maintenance
and fuel reloading is distributed as follows: 130, 217, and 35 ber, respectively.
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The radioactive discharges from power stations containing VVER-440 to the atmosphere
have been stabilized at the level of 2.1-3.3 Ci/MW(el)yr, which is close to the average
for power stations containing PWR. The activity of liquid discharges to open bodies of water
has not exceeded 300 mCi/yr. The radiation backgrounds in areas around nuclear payer stations
in the main are determined by global fallout and the natural ionizing-radiation background.
In normal operation, the power stations have virtually no effect on the environment.
At present, the staff numbers at nuclear power stations exceed the standard level by
15%, mainly due to maintenance staff, which account for about 53% of the total. The staff
numbers at Soviet nuclear power stations are higher by factors of 2.5-3.5 than those at nu-
clear power stations in the developed capitalist countries, while the numbers Of maintenance
staff are larger by factors of 3-5, mainly because in capitalist countries the equipment is
maintained usually by the maker, whose staff are not included amongst the operating staff.
The approach used in this country in organizing nuclear power station servicing is due to the
lower equipment reliability, the absence of spare power, and the consequent higher equipment
use.
UPGRADING VVER DESIGN
Improvements in nuclear systems containing VVER reactors follow the lines of increasing
of the following:
1) unit power: in the 20 years since the first mit at the Novyi Voronezh power station
was commissioned, the unit power of the VViR has increased from 210 to 1000 MW(ed.);
2) first-circuit pressures and steam parameters: during this time, the pressure in the
reactor has been raised from 9.8 to 15.7 MPa, and the steam pressure in the steam generators
from 3.1 to 6.3 MPa;
3) the power density in the core has been raised from 47 to 111 kW/liter by equalizing
the power distribution over the radius and revising the neutron-physics and thermophysics
characteristics of the core; and
4) increasing the burnup from 13 to 40 W.day/kg U by improved fuel-pin
of boron regulation.
Some initial designs have been retained:
1) railroad transportability of the reactor body;
2) the use of hexagonal fuel assemblies in the core containing rod pins filled with
uranium dioxide sheathed in an alloy of zirconium with 1% niobium;
3) the use of high-tensile chromium-molybdenum steels for the body; and
4) the use of horizontal steam generators to produce saturated steam.
We now mention some changes that have been made in the basic designs as a result of ex-
perience with the VVER.
Reactor and Enclosed Equipment (EE). The displacement of the thermal shield at the
first unit in Novyi Voronezh power station in 1969 led to reconsideration of the flow and
mounting conditions or all the EE components. The thlrmal screen has been completely
eliminated in the VVER-1000 and later models of the VVER -440.
Originally, the body was made from 15Kh2MFA steel without anticorrosion coating, where
suitable water treatment ensured a satisfactory state of corrosion in the inner surface.
However, anticorrosion coating was applied beginning with the first unit in the Lovisa power
station, in accordance with world practice and to simplify the water-treatment requirements.
Core and CPS. To increase the core power, the outside diameter of the fuel pins has
been reduced from 10.2 to 9.1 mm, while the number of pins in an assembly has been increased.
The reactivity margin in the first reactors was compensated by mechanical CPS. Starting with
the third unit at Novyi Voronezh station, the reactivity margin compensating for burnup was
compensated along with slow reactivity changes by introducing boric acid into the coolant,
which reduced the number of mechanical CPS units in the VVgR-440 from 73 to 37.
/gg., Originally, the units were fitted with low-inertia sealed pumps working at 1500
rpm, which were supplied when the external line was disconnected from the turbine genetator.
use
design and the
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In the VVER-1000 and the later designs of power stations containing VVER-440, pumps are used
in which surges are absorbed by a special flywheel.
Steam Generators. The main design changes have been related to providing access to first-
circuit collectors for examination and maintenance of the tube-mounting points directly from
the central bay, and there have also been improvements in the collector unit at the boundary
between the water and steam in the second circuit.
Safety Systems. In the first VVER, the maximum design emergency was taken as the in-
stantaneous failure of a pipeline of diameter about 100 mm bearing a unilateral flow. In
the current VV/R-440 and VVER-1000, the protection and localization devices provide safety
in emergencies extending to instantaneous failure of the main circulation pipe coinciding
with complete current failure. Systems are also envisaged for emergency core cooling (hy-
draulic vessels connected in pairs to the inlet and outlet pipelines, with groups of low-
pressure and high-pressure pumps), which prevent the fuel-pin sheath temperatures from rising
above 1200?C.
Fission products escaping from the main circulation loop are localized in new power sta-
tions containing VVER-440 units by a system of sealed boxes, in which the maximum pressure
is 0.15 MPa. It is guaranteed that this pressure is not exceeded by condensing the steam
during the first period of a maximal design emergency using a special bubble condenser.
In power stations containing the VVER-1000, there is a protective containment around
all the sections of the main circulation circuit and the reactor hall, which is designed
to withstand a total pressure such as would arise from the escape of all the coolant (0.4
MPa) with provision for reducing it with a sprinkler system. A high level of independence
is also provided in the duplicated protective and localizing systems by locating them in
different buildings with separate power supplies etc.
THE UNIFIED PROJECT
A unified project has been drawn up for nuclear power stations containing VVER-1000
reactors. This design provides for flow production of the units and should substantially
reduce the installation time, which will greatly increase labor productivity because the
same operations are reproduced and are executed by specialized teams. The main and auxiliary
items of equipment have been standardized to provide power stations with similar designs no
matter where the equipment is manufactured, which will also increase labor productivity by
leading to longer runs. The start of flow production began at Zaporozhe power station, where
four reactor units are being installed simultaneously. Single-block styles were used for the
first time in Soviet power stations, which eliminates the gap between the reactor and ma-
chine sections, and this reduces the loss in steam parameters and increases the unit power
by 6 MW. The turbines have been fitted with modified condensers, which has reduced the size
of the machine bay by 3 m. These and other design modifications have reduced the volume re-
quired for the main housing by 20%, with the consumption of reinforced concrete reduced by
6%, building labor involved by 30%, equipment mass by 9%, and pipelines by 12%. The unified
project differs from the fifth unit at the Novyi Voronezh station also in having better phys-
ical characteristics. The number of control and protection units has been reduced from 109
to 61, while the number of absorbing components in one unit has been increased from 12 to 18,
while the number of fuel-pin assemblies has been increased from 151 to 163 because jackets
have been abandoned.
The Atomic Heat and Elestricity Design Institute has compared the unified project for
power stations containing VVER-1000 with foreign power stations. This has shown as follows:
1. The specific working area of analogous foreign power stations (France and the Fed-
eral German Republic) is less by 34-35% than that in the unified project because the prin-
ciples used in determining these areas differ particularly as regards the set of buildings
and structures required for auxiliary and maintenance services (there is no nitrogen-oxygen
system or acetylene station with stock of carbide, nor are there mechanical maintenance work-
shops and stores). Also, foreign nuclear power stations are often built without access to
railroad transport, which also reduces the area required.
2. A comparison has been made of volume characteristics for the reactor sections in
the unified project and for power stations in France, the USA, and the Federal German Re-
public, which has shown that stations such as Paluel and Belfonte are substantially better
(specific volumes 14% less, consumption of reinforced concrete and metal 30-35% less). In
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these stations, there is high unit power in the main equipment within dimensions ,enabling
one to fit them into the same building volumes as the unified project for the VVER-1000.
The reductions in reactor-section volume in power stations such as Bouget and Tricastin has
been attained also be using reactors with three circulation circuits and vertical steam
generators. At the Mhlheim-Kerlich and Biblis stations (in the Federal German Republic),
better parameters have been obtained in particular because more compact electrical engineer-
ing and ventilation systems have been employed. In all cases, foreign power stations use
closer spacing in the reactor section, which adversely affects maintenance and servicing on
site.
3. The unified project differs considerably from the Biblis power station regarding
volume of the machine bay: respectively, 300 and 150,000 ms. This very substantial differ-
ence occurs because the turbines at Biblis are much smaller, as at most foreign nuclear power
stations, including as regards the diameter of the low-pressure cylinders, while the numbers
and dimensions of the separators, steam superheaters, and regenerative heat exchangers have
been reduced, while the low-pressure heaters are built into the condenser connecting pipes
and do not occupy any additional volume.
A substantial difference also is that foreign power stations provide minimal maintenance
areas in the machine bay, in particular because the equipment is installed without modifica-
tion at the building site (on wheels).
4. The area of special facilities is estimated as 7800-8200 m2 for foreign stations and
9500 m2 for the unified project, the difference being due to foreign stations not having long-
time stores for radioactive wastes and having much smaller volumes of domestic accommodation
and also the absence of laundries.
The unified project is inferior on specific parameters to foreign power stations also
on certain auxiliary equipments, which is due primarily to the numbers of operating and main-
tenance staff, whose accommodation requires additional buildings, medical services, etc.
On the whole, this comparison of the unified VVER-1000 project with foreign stations
has shown that the design parameters are comparable with current levels. Nevertheless, the
designs for Soviet power stations still leave some margins that should be utilized in the
next generation of designs.
SCOPE OF IMPROVING UNIT POWER ECONOMIC PARAMETERS
At present, various research and development organizations in the country are examining
three basic ways of providing essentially nsw designs capable of giving a new level of economic
performance in power stations containing VVER: fuel-cycle improvement, reactors with super-
critical coolant, and substantial increases in unit power.
We consider the last of these as it is most fully developed. As traditional VVER have
attained a high level and nearly optimum values have been obtained for parameters such as
coolant pressure, steam production, and power level, one of the ways for improving reactors
of this type is to increase the unit power substantially. All previous power-engineering
developments in this country and abroad have been accompanied by the enlargement of power
stations.
The main factors responsible for this tendency are firstly the economic advantages of
concentrating production deriving from increased labor productivity, as is characteristic of
other branches of heavy engineering, and secondly the difficulties in providing the required
rates of increase in installed power without increasing unit power levels. However, as in
thermal power stations, increasing the unit power on nuclear systems not only reduces the
specific capital costs per station but at ?the same time adversely affects the system reliabil-
ity, and to maintain this at a fixed level there must be additional costs involved in increas-
ing the emergency backup and maintenance facilities. Therefore, increasing the unit power,
including in nuclear systems, is effective only if the reduction in the specific power-gener-
ation costs is at least not less than the costs for providing the additional backup. This
problem will have to be dealt with in the designs presently being drawn up by the Atomic
Thermal Electric Project Organization for a unit containing VVER-1500 reactors.
LITERATURE CITED
1. F. Ya. Ovchinnikov et al., At. inerg., 54, No. 4, 249 (1983).
2. L. Howles, Nucl. Eng. Int., 28, No. 347, 36 (1983).
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STATE-OF-THE ART AND DEVELOPMENT PROSPECTS FOR
NUCLEAR POWER STATIONS CONTAINING RBMK REACTORS
E. V. Kulikov UDC 621.039.577
Over a comparatively short period, there has been a substantial increase in the generat-
ing capacity in the country on account of nuclear stations containing RBMK-1000 reactors,
which have been rapidly run up to nominal power and which have provided stable and safe oper-
ation, so one can say that this reactor type is promising for nulcear engineering in the next
few decades. The following advantages are responsible for the major plans for building nu-
clear stations containing RBMK [1]:
1) the RBMK units and equipment are made at existing plants in the country and have not
required, apart from construction in Yugoslavia, the building of new industrial organizations
with the unique equipment for making large components working at high pressures;
2) for these reactors there are virtually no limits to the unit power associated with the
manufacture, transportation, and installation;
3) the branching in the circulation system increases the overall safety because it elim-
inates a complete coolant loss from the core and enables one to build reliable protection
systems and devices for localizing leaks;
4) the good physical characteristics of the reactor and the continuous fuel recharging
make it possible to provide efficient use of low-enriched fuel together with extensive burnup
giving low contents of fissile uranium isotopes in the spent fuel, and in addition there is a
fairly substantial increment in the burnup as a result of the incidentally produced plutonium;
and
5) the high reliability in the heat-engineering units is supported by wide parameter
control ranges with monitoring in each channel.
The basis of a nuclear station containing RBMK is provided by two units of electrical
power 1000 MW each with a common machine hall. Each unit is a reactor With its circulation
system and auxiliaries, steam and condensate-feed units, and two turbine generators of power
500 MW each. The essential scheme of a unit is shown in Fig. 1.
The reactor is located in a concrete pit on welded metal structures, some of which are
used simultaneously for radiation protection, and which in conjunction with the jacket form
a sealed space filled with helium-nitrogen mixture (the reactor space), in which the graphite
stack is located. The stack contains the fuel channels (FC) and the CPS channels, which run
through the upper and lower metal structures.
A fuel channel (Fig. 2) is a welded tube construction intended to take a fuel assembly
(FA) and to organize the coolant flow. The upper and lower parts of the channel are made of
stainless steel, while the central part within the core is made of zirconium-niobium alloy
having good mechanical parameters and corrosion resistance together with a low neutron ab-
sorption cross section. The central part of the channel is coupled to the upper and lower
ones by special couplers.
Twelve units containing RBMK -1000 are now operated: 4 at Leningrad Lenin nuclear power
station, 4 at Chernobyl, 3 at Kursk, 1 at Smolensk, and 1 unit containing an RBMK -1500 at the
Ignala nuclear station. The overall installed power of nuclear stations containing RBMK is
over 60% of the total nuclear power in the Soviet Union. These nuclear stations have very
good performance parameters, on which they are in no way inferior to the best nuclear power
stations in the most developed countries in the World. As an example, Table 1 gives the
economic parameters of the units at Leningrad and Kursk stations.
Translated from Atomnaya inergiya, Vol. 56, No. 6, pp. 359-365, June, 1984.
368 0038-531X/84/5606-0368$08.50 ? 1984 Plenum Publishing Corporation
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Fig. 1. Scheme for a nuclear station containing RBMK-1000: 1) stack; 2) wet gas hold-
er; 3) delay gas holder; 4) helium purification system; 5) compressor; 6) gas loop
condenser; 7) FC integrity monitoring system; 8) pumping and heat transfer system in
CPS; 9) cladding sealing monitoring system; 10) separator; 11) control unit; 12)
emergency feed pump tanks; 13) FC; 14) CPS channel; 15) reactor; 16) MCP; 17) cool-
ing pumps; 18) regenerator; 19) pipe to purification system; 20) cooler; 21) emerg-
ency reactor cooling system; 22) compressed air; 23) emergency pump; 24) technolog-
ical condensers; 25) superheaters and separators; 26 and 27) turbine generators 1
and 2; 28) condensers; 29) KN-1 condensate pumps; 30) condensate purification; 31)
apparatus for burning explosion mixture; 32) pipeline to delay gas holder; 33) low-
pressure heaters; 34) deaerators (0.7 MPa); 35) electric feed pumps.
The high reliability of the RBMK-1000 is confirmed by the operation of the units for 52-
reactor-years and is indicated by numerous research studies, calculations, and design studies
performed in the early stages of operating the units. Improved reliability in fuel recharg-
ing with the reactor working has been provided by modifying the sealing plugs in the FC, the
ball flowmeters, and the control valves, while upgrading in the units within the containment
has extended to the separator drums in the main steam pipes, which has provided more uniform
loading on the separator drums and the required steam wetness under stationary and transient
conditions.
Reliable core cooling is provided when there is emergency reduction in the feedwater
flow rate, which can extend as far as complete stoppage, is provided by an automatic system
for reducing the power involving constant automatic comparison of the thermal power with the
feedwater flow rate [2]. Many experiments have been performed at the Leningrad station to
determine the cooling parameters arising from the natural circulation directly with the unit
working, which have shown that previous calculations and testbed results are reliable, and
this has provided proposals for means of switching off the main circulation pumps MCP at an
appropriate time and for accelerating the turbine unloading. The final stage in the research
on the natural circulation conditions was carried out at the Kursk station in 1981.
Nuclear stations containing RBMK have safety systems that eliminate any hazardous es-
cape of radioactive material even in the case of unlikely accidents, including failures in
major pipelines [1]. From this viewpoint, the main hazard lies in failure in the pressurized
MCP pipeline, since this halts the flow of coolant to the FC in the half of the reactor af-
fected by the emergency. This hypothetical accident determines the characteristics of the
emergency cooling system (ECS), including the response rate of this and the maximum capacity.
The water from the ECS is supplied to each distributing group collector (DGC), and to
avoid water escaping through the failed section there are nonreturn valves at the inlet to
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TABLE 1. Economic Parameters in the Oper-
ation of Nuclear Units Containing RBMK-,
1000
nit
Nuclear station
1982.
power pro-
duction,
million
kW ?h
PUF, To
*
eff.,
net, To
Leningrad
Kursk
1 . 6185,6 70,6 28,74
2 7864,7 89,8 29,58
3 7007,9 80,0 28,38
4 7298,1 83,3 . 29,19
1 7548,7 86,2 29,98
2 6305,0 - 72,0 29,36
Nuclear station
Unit
1983
power pro-
duction,
kWhmillion
PUF, To
.
eff.,-
net, To
Leningrad
Kursk
1 7739,4 88,3
2 7356,9 84,0
3 6234,7 71,2
4 7487,8 85,5
1 7237,5 82,6
2 6183,5 70,6
'Without allowance for heat use.
29,51
29,20
29,23
28,78
28,96
28,26
Fig. 2. A fuel channel: 1) protec-
tion unit; 2) FA suspension; 3) FC
head; 4) pressure tube; 5) upper
protection plate; 6) thermal shield;
-. 7) Steel-zirconium joint; 8) FA; 9)
stack block; 10) supporting vessel;
11) lower plate; 12) channel sec-
tion; a) steam-water mixture outlet;
b5 water inlet.
the DCC. The ECS consists (Fig. 3) of a major subsystem containing hydraulic accumulation
unit and i.long -time cooling subsystem having special pumps and water stocks in tanks. The
cooling water is passed through pumps to the ECS collector in each half of the reactor and
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Fig. 3. Essential scheme for circulation loop and ECS:
1) reactor; 2) separator; 3) MCP; 4) pressurized collec-
tor; 5) DGC; 6) hydraulic accumulation unit in ECS; 7)
feed pump; 8) ECS pumps; 9) water stock in condensation
device; 10) ECS collector; 11) restriction nozzle; 12)
intermediate throttling link; 13) cutoff float valve;
14) fast ECS valve; 15) DGC insert; a) steam to turbines;
b) condensate return.
then along pipes to each DGC. Fast slide valves are mounted in the water feedlines and col-
lectors, which open when the signal to actuate the ECS arrives. The algorithm for activating
the main subsystem in the ECS provides for cooling the core when there is complete or partial
failure in large pipes and eliminates incorrect operation in accidents not involving failure
in the circulation loop.
The RBMK gives extensive burnup with low initial enrichment, which is provided by con-
tinuous fuel recharging with the reactor working. In all the nuclear stations containing
RBMK, there is ongoing fuel recharging at power by means of an unloading and loading machine.
The continuous-recharging mode enables one to roughly double to burnup by comparison with
complete fuel change in one operation. The 295U concentration is reduced from 18-20 to about
3.7 kg per ton of uranium, while the amount of fissile plutonium attains about 2.8 kg per ton
uranium. This change in isotope composition results in substantial changes in the core neu-
tron physics characteristics.
In the steady recharging state, only the local characteristics such as the power in the
channels alter, while the characteristics of the reactor as a whole remain virtually unchanged,
whereas during the first operation of the reactor loaded with fresh fuel and additional ab-
sorbers, there are fairly substantial changes in the physical characteristics, in particular
in the reactivity coefficients (steam and temperature ones). The values of these coefficients
are dependent not only on the fuel isotope composition but also on the number of absorbers in
the core. Experience with the RBMK-1000 has confirmed the theoretical conclusions that the
reactivity coefficients increase and that the stability in the power distribution decreases
as the fuel burns up and the absorbers are extracted. The radial-azimuthal energy distribu-
tion is the least stable, where the form of the nonstationary deformation is determined by
several of the lower harmonics. The distribution is stabilized in two ways:
1) improving the automation level by means of a branched reactor control system; and
2) increasing the fuel enrichment.
Under the first approach, essentially new systems have been introduced for local auto-
matic control of the power distribution (LAC)and local emergency protection (LEP), which op-
erate from transducers within the core [3]. The LAC system automatically stabilizes the
lower harmonics in the radial-azimuthal distribution. This system maintains the overall set
reactor power level, by using individual effectors to provide automatic power control in the
individual core regions. The LEP system provides for emergency power reduction when there
are impermissible local power rises, in spite of the action of the LAC. The LAC and LEP use
groups of effector mechanisms (from 7 to 12 of them) uniformly distributed over the core and
containing control rods each surrounded by two LAC transducers. The averaged and corrected
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lv
46 10'
7:I :0-8
C.)
111-9
V4?
0 1
Operating
2
time, yr
Fig. 4. Distributions of radioactive
products in deposits (Ci/cm2) (1) and
in coolant (Ci/kg) (2).
signal from the LAC transducers is used to control the rods. The transducers in the LAC-LEP
system are triaxial chambers placed in central sealed sleeves in the FA.
Calculations on the performance expected from the second approach have shown that in-
creasing the initial 235U content of the fuel improves not only the dynamic parameters but
also the economic characteristics by increasing the extent of burnup and reducing the specific
fuel consumption. It has been found that there is a substantial dependence of the time con-
stant for the first azimuthal harmonic on the steam reactivity coefficient. The less the
positive steam reactivity coefficient, the higher the power distribution stability and the
simpler the reactor control. The most rational way of reducing the steam coefficient is to
increase the ratio of the 23513 concentration to the moderator concentration in the core. The
reduction in the steam coefficient as a result of going to fuel with 2% enrichment is about
1.3 0.* These conclusions served as basis for increasing the RBMK enrichment.
We give below the basic characteristics of the RBMK-1000 fuel cycle for 1.8% and 2%
initial 285U contents (first and second values correspondingly):
Uranium burnup, MW.day/kg 18.5 22.3
Final 235U content in unloaded
fuel, kg/ton 3.9 3.5
Reduction in steam reactivity
coefficient, 0
1.3
Annual input of enriched uran-
ium, ton/GW (with PUF = 0.8)
50.5
42
Annual consumption of fuel pins
to supply reactor (with PUF =
0.8), 102 per GW
16.0
13.3
Annual consumption of natural
uranium, ton/GW*
169
158
Mean FA use, effective days
1100
1350
*With PUF = 0.8 and 235U content in spent fuel of 2-3 kg/ton.
Since the first unit at Leningrad station began to operate, there has been continuous
monitoring of the radiation environment at the station and in the surroundings, which has
confirmed that the designis correct and has provided detailed data required in upgrading the
radiation safety systems in stations containing RBMK. These studies have shown that the ir-
radiation of the station ,staff on average is considerably below the permissible level: the
annual dose for over 35% of the operating staff does not exceed 0.2 ber, and only for 2% is it
between 4 and 5 ber. The collective dose to the staff in one unit after about 5 years of op-
eration was about700 man-ber (1 ber = 10 mSv), which is virtually the same as the dose at
foreign nuclear stations of the same power. The largest contribution to the irradiation
(40-50%) comes from prophylactic maintenance operations and operations associated with mon-
itoring the metal; up to 60-65% of the dose to the station staff is due to operations dur-
ing prophylactic or major servicing, i.e., with the reactor shut down.
*0 is the effective proportion of delayed neutrons.,
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Fig. 5. Block construction of the control room at Ignala nuclear power station.
Fig. 6. Loading the RBMK-1500 with fuel at Ignala power station.
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During this period, the radiation environment around the first circuit is determined
by the emission from radioactive corrosion products deposited on the inner surfaces (Fig.
4). The contributions from the y rays from different nuclides to the dose rate vary during
the operation, and after about 5 years the radiation environment is determined by the y
rays from "Co. The deposits on the equipment contain fission products, but their contri-
bution to the y-ray dose rate is not more than 10-15% on operating the station for 10 years.
During the entire period of operation for nuclear stations containing R3MK-1000, there
have been no instances where the discharges of radioactive gases and aerosols have exceeded
the permissible values laid down by SP AES-79 (or the earlier SP AES-68). As the permissible
discharges have not been exceeded, the radiation background in the surroundings is satis-
factory. Direct measurements on the y-ray dose rate in the locality indicated no increase
over the natural background throughout the area.
The definitive parameters restricting RBMK power are the fuel temperature, the tempera-
ture of the graphite and the metal construction, and the margin from the heat-transfer crisis.
These parameters have not yet attained the limiting permissible values at existing RBMK-1000
reactors.
The margins of the temperature of fuel, graphite, and metal have led to the suggestion
of building a more powerful reactor based on the RBMK-1000 without change in general design
and dimensions. This requires a design that enables one to increase the critical channel
power without changing the dimensions or number of the FA, i.e., to increase the margin up
to the heat-transfer crisis. The solution is seen as FC with heat-transfer intensifiers.
A design has been developed for a new FC for the RBMK-1500 with special devices enabling one
to increase the heat loading, which is characterized by a high level of standardization in
the units based on RBMK-1000 ones. This means that the production of FA of a single type
will not only simplify the process by reducing the number of different types to be produced
but will also give a considerable economic gain from using the new FA in existing RBMK-1000,
which will increase the core reliability, improve the stability in the power distribution,
and raise the fuel burnup. The FA with intensifiers enable one to increase the thermal
power of these by 20-25% in the RBMK-1000 with separator drums of diameter 2600 mm. This
broadens the scope for the wider use of the reactors in district heating.
The RBMK-1500 at the Ignala power station has been commissioned and is being run up to
nominal power, the unit electric power being over 1500 MW. This unit has represented a start
on a new generation of channel reactors, as more economical ones should ultimately replace
the highly successful 1000 MW ones. Stations containing 1tBMK-1500 will reduce the specific
capital investment by 20-30% relative to ones containing RBMK-1000 and will also reduce the
fuel costs. Experience with designing, building, and operating the RBMK boiling-water re-
actors has shown that the correct decision was taken on building a large series of nuclear
stations containing reactors of this type, and that there are good prospects for developing
them further.
LITERATURE CITED
1. N. A. Dollezhal' and I. Ya. Emel'yanov, A Channel Nuclear Power Reactor [in Russian],
Atomizdat, Moscow (1980).
2. I. Ya. Emel'yanov, S. P. Kuznetsov, and Yu. M. Cherkashov, At. Ener., 50, No. 4, 251
(1981).
3. I. Ya. Emel'yanov et al., ibid., 49, No. 6, 357 (1980).
4. A. P. Aleksandrov and N. A. Dollezhal', ibid., 43, No. 5, 337 (1977).
374
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STATE-OF-THE ART AND DEVELOPMENT PROSPECTS FOR NUCLEAR
POWER STATIONS CONTAINING FAST REACTORS
O. D. Kazachkovskii UDC 621.311.2:621.039+621.524.526
The scope for extending nuclear fuel breeding in fast reactors was predicted in the
1940s. In essence, this was a conceptual extrapolation from thermal reactors to fast ones on
the basis of the scanty experimental data then available, i.e., the parameters of the inter-
action of fast neutrons with matter. Subsequently there were studies designed to confirm
or reject this suggestion. It should be noted that these studies began before the first nu-
clear power station in the World had been commissioned, when it had not yet been shown that
nuclear power stations were really feasible. However, the importance of the problem led to
considerable effort being devoted to determine the scope for extended breeding. In essence,
this corresponds to a requirement in the scientific revolution: the transition to operations
in a new stage need notwait onthe endof theprevious one, since otherwise rates of progress
would be inadequate. An unambiguous positive solution was obtained at the end of the 1950s,
when it was established that the breeding factor in a fast reactor can be greater than one.
This meant that fast reactors in principle allow one to use all of the mined uranium. The
raw-material base for nuclear power was thereby increased by factors of tens or hundreds rel-
ative to the case where only thermal reactors are used.
The next stage involved determining the engineering feasibility of industrial fast re-
actors. This was concerned mainly not so much with construction of high-power fast reactors
but instead the commercial fast reactors meeting certain requirements, particularly from the
viewpoint of fuel-cycle economy: heat production density of 500-1000 kW/liter of core, and
burnup of 10% of the heavy atoms in a run of more.
Right from the start, researches on fast reactors were directed to using sodium as the
coolant. Water is unsuitable for fast reactors because of its nuclear-physics parameters.
Sodium on the other hand has good thermophysical parameters and acceptable nuclear-physics
characteristics. An important point also is the high boiling point (about 900?C), which
means that a sodium-cooled reactor does not need to use high pressures, which is a consid-
erable advantage from the engineering viewpoint. Also, simultaneously but mainly incidentally
one attains a considerable advance in efficiency by comparison with a water-cooled reactor.
For some time there was also a discussion on the use of sodium-potassium eutectic, although
this is worse in thermophysical properties and was found to be less suitable than sodium.
Therefore, one was involved in developing an entirely new industrial technology for using
sodium as a coolant. The problem was solved in a short period. In any case, it was shown
that there were no essential difficulties in setting up a large-scale industrial sodium
technology.
At the start of the 1960s, when the BN-350 began to be developed as the first commer-
cial fast reactor, an experimental fast reactor with sodium cooling, the BR-5, had already
operated at Obninsk. The thermal power of this was only 5 MW, whereas the design thermal
power of the BN-350 was 1000 MW. It must be emphasized that this was a very large step,
which was not decided on at once, but it corresponded to the general requirements in the
scientific revolution and the requirements for rapid devleopment. It is true that apart from
the power level, the BN-350 parameters were moderate: the coolant temperature, the tempera-
ture of the working body (steam), and the pressure were low. This was deliberate, since one
of the main purposes of the reactor was to determine the effects of the scale factor on the
working characteristics of such systems while avoiding any additional difficulties. Almost
simultaneously, work began on the BOR-60 research power reactor, which was designed to have
operational determinations of the effects from high power levels, high temperatures, and
other factors on the working characteristics of fast-reactor components and to define reason-
able safety margins for the engineering parameters. The development of the BN-600 began
_
Translated from Atomnaya Energiya, Vol.: 56, No. 6, pp. 365-370, June, 1984.
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somewhat later. This was already a reactor with high and not excessively conservative thermal
parameters. At the same time, the power was such as to approach the basic parameters of future
standard fast reactors.
The BN-350 began to work at full power in June 1973, and the BN-600 in April 1980.
It is now possible to survey the operation of both reactors. Firstly, the design val-
ues were attained for physical parameters such as the critical mass, the distribution of
heat production over the core, and the kinetic, thermophysical, and hydrodynamic character-
istics. The basic engineering parameters also corresponded to the design values, apart from
the power in the BN-350, whose nominal value was set somewhat below the initial design value
on account of difficulties with the steam generators. At present, the BN-350 is operated at
720 MW (thermal) with the following parameters:
Electrical power, MW up to 130
Distillate output,
kg/sec; 1000
Sodium Temp., ?C:
at reactor inlet 283
at reactor outlet 425
Superheated steam pressure, MPa 4.5
Maximum burnup, MW.day/kg:
in low-enrichment zone 54
in high-enrichment zone 60
Power use factor, % N 88
The refined neutron-physics characteristics and other data obtained during the operation
of the BN-350 provided a basis for modifying the core and screen reloading programs. Improve-
ments in the CPS increased the working time between reloading cycles from 54 to 73.5 days.
The burnup was thereby increased by about 15% relative to the design value.
The basic working parameters of the BN-600 are as follows:
Electrical power, MW
Sodium Temp., ?C
at reactor inlet
at reactor outlet
Superheated steam pressure at
turbine inlet, MPa
Maximum burnup, MW.day/kg:
in low-enrichment zone
in high-enrichment zone
Power-use Factor, %
600
377
550
12.7
41.0
61.0
N71.8
The equipment in the sodium system (apart from the steam generators) worked almost en-
tirely without faults in both of these reactors. In essence, the failures and unplanned shut-
downs were due to the third loop (steam-water one).
The scale factor was substantial in relation to the steam generators. In the BN-350 and
BN-600, there were water leaks into the sodium and corresponding failure in some steam gener-
ator.sections. These failures were not due to any essential features of the steam generators,
but they were due to the development of defects or rather to the nuclei of defects at welding
points, which on account of their smallness could not be observed by monitoring during manu-
facture. An essential point was that the defect growth rates up to a certain critical size
became large. This required the introduction of especial systems for early defect diagnosis.
This also imposed certain requirements as regards promptness in action by the staff. Failure .
in the steam generators did not lead and could not lead to any catastrophic consequences.
The instances of steam generator failure in either reactor occurred during the initial per-
iod of operation, essentially in the running-in period. The subsequent operation was free
from faults. Nevertheless, there remains the problem of improving the reliability, or rather
improving the degree of detection of defects during manufacture and also the scope for cor-
recting errors and mistakes during operation. Here there is obvious scope for applying des-
ign improvement and for using more suitable materials for the pipes.
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The scale factor also made itself felt to some extent in relation to the BN-600 fuel
pins, which work under heavy stress. The working life of the pins and correspondingly the
burnup are at present less than the planned values. Here there are some major factors
characteristic of fuel-pin working conditions in fast reactors, including some that were not
known in advance. Some constraints at present are imposed by the vacancy-swelling effect in
constructional materials produced by intense fast-neutron fluxes.
Further improvements in fuel-pin life are based on advances in constructional materials.
One can safely say that it is possible to improve the working life by a substantial factor,
but the researches and the tests on the systems inevitably require long time, and in addition
the studies after irradiation in hot laboratories are usually laborious and lengthy. This
means that measures have to be taken to substantially expand the scale of the operations.
It can thus be said that clearly that the engineering feasibility of commercial fast
power reactors has been demonstrated. After a certain initial startup period (which was
very short for the BN-600), both reactors have been operated reliably with large power-use
factors.
There are two major technical problems requiring research in order to improve the work-
ing characteristics of fast reactors further: improving fuel-pin life and increasing steam
generator reliability.
The future fast-reactor development program must be based on the economic parameters.
Economic aspects are ultimately always the last and final criteria for the desirability of the
large-scale development in any area. Here the situation is as follows. Of course, we can-
not determine the economic performance of future large standard commercial fast reactors by
extrapolation from existing stations with any great accuracy. Inaccuracies always appear in
extrapolation. However, at present we can say that the present fast reactors are foreseen
as developing much the same thermal power per specific capital cost; an exact figure cannot
be given but one can speak of a difference of 30-40%.
On the other hand, the fuel component for the fast reactor is less, since the
fuel breeds at a higher rate than in a thermal reactor. Estimates in relation to
the fuel component are even less reliable, since at present there is no experience with rou-
tine commercial fuel reprocessing. However, it seems that one can conclude that existing
prices for natural uranium mean that the advantage over the fuel component (relative to ther-
mal reactors) cannot compensate for the disadvantage over specific capital costs. In the
future, as the cost of natural uranium rises, the cost relation will steadily improve in
favor of fast reactors. At some time there will be an inversion, and fast reactors will be
economically better than thermal ones. No date for this can be given. It should however
be remembered that the working life of a nuclear power station is long at 30-50 years. As
this life is long, it will be economically desirable to include fast reactors extensively
in the nuclear power program before the time of economic inversion has been reached. On the
other hand, it should be borne in mind that there are certain and by no means minor oppor-
tunities for reducing the specific capital costs. This scope is bettter for fast reactors
than for thermal ones. Fast reactors have high efficiencies, which are almost 30% more than
for thermal ones, which is already a great advantage. There is no particular point in rais-
ing the thermal parameters. Although the efficiency then increases somewhat, the specific
capital costs will most probably not thereby be reduced but increased. However, the main
point is that raising the temperature substantially reduces the fuel-pin life. Here one
expects a sharply varying relationship, and this will certainly have a very adverse effect
on the fuel component. On the other hand, it is desirable to increase the unit power, since
the specific capital costs are thereby reduced. Fast reactors enable one to raise unit power,
since they are compact and do have have pressures within the containment. One can say that
it is already technologically possible to make fast reactors of power 2000 MW and more. There-
fore, it is better to compare fast and thermal reactors not at identical power levels but at
the power levels that are accessible with the existing manufacturing technology. Finally,
there is scope for improving and simplying fast reactors, which extends to abandoning the
intermediate circuit, heating the main pipelines, etc.
It should however be noted that although the unit power will increase, we at present
cannot envisage and probably for a long time will not be able to envisage enlarging the steam
generators in fast reactors. At present, the nuclear station containing the BN-600 employs
a modular principle, with many sections (modules) of comparatively low power. When a module
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fails, it is switched out without shutting down the reactor, as has been demonstrated with
the BN-600. Then the module is replaced by a new one during the next planned preventative
maintenance. The modular principle has been found to be fully justified and should be used
in the future.
As regards the fuel cycle, a solvent-extraction technique has been developed for re-
processing the fuel rods. This will evidently be used in the first stage of large-scale fast-
reactor development. It may be that a nonaqueous technique will subsequently be devised.
Here there are no criticality constraints, and the wastes are obtained at once in solid form,
which opens up good scope for shortening the external fuel cycle. It is possible that non-
aqueous technology may not provide a sufficiently high degree of purification, but this is
not necessary from the physics viewpoint, since the fast-neutron absorption cross sections
in fission products are comparatively small. Instead, remote handling is required for the
fuel. On the other hand, the inherent activity of the recirculated plutonium will be so
high that remote handling will always be used. Also, the large scale of fuel-pin manufac-
ture in the future will necessarily require automation, which naturally fits in well with re-
mote handling.
We now consider the breeding rate. A breeding factor greater than one is already good,
and this is the essential feature that distinguishes the scope for using all mined uranium
completely from the partial use. However, higher values are desirable, because of the need
for self-sufficiency in fuel in developing nuclear power. The required breeding rates at
various stages may differ. Evidently, in the first stage the breeding rate should be high,
in order that such reactors can rapidly take over a considerable part of the total nuclear
power program. In the first stage, it is not essential to provide self-sufficiency in fuel,
since the plutonium accumulated in thermal reactors can be used to fuel new fast ones. Plu-
tonium is most valuable when it is used in fast reactors. It is clearly undesirable to burn
it in thermal reactors. It is thus very necessary for the first stage, the stage where fast
reactors take on an asymptotic mode of development. It is also possible that to fuel them,
i.e., to provide the initial loading for new nuclear stations, one should use enriched uran-
ium if the rate at which new capacity is introduced is high. Here one must remember that
we have a certain enrichment capacity for thermal reactors, and this capacity can be used
for fast reactors when these become the main line of development.
When the asymptotic growth mode has been attained for fast-reactor power, the rate will
be determined by the required rate of power engineering development. Fast reactors could
provide asymptotic development with self-sufficiency. It is also possible that they will be
required to produce additional plutonium for other power systems such as thermal reactors
not working in baseload mode. It is also possible that additional fuel will be required
for commercial high-temperature reactors, etc. Fast reactors should be capable of providing
cfluis additional plutonium. It is true that in that case one cannot speak of the breeding
rate, since this rate is the ratio of the excess plutonium produced to the total load in
the cycle of a fast reactor. If the plutonium is diverted to other uses, there is no need
to calculate the breeding factor, and it will be incorrect to refer the amount of excess
plutonium is diverted to other uses. There is no need to calculate the breeding factor, and
it will be incorrect to refer the amount of excess plutonium to the load in a cycle.
The breeding rate is determined primarily by the breeding factor and the specific heat
production, i.e., the specific power per unit amount of fuel. The latter is controlled
mainly by the scope for cooling,the core and by the acceptable thermal stress on the fuel
pins and does not exceed 1000 kW/kg of fuel. The breeding factor is also dependent to .a
substantial extent on the form of the fuel. When fast reactors began to be developed, they
were oriented to a use of uranium-plutonium metal fuel. However, it soon became apparent
that there were difficulties with metal fuel: swelling of the uranium and hazardous inter-
action between the metal fuel and the cladding. It was therefore necessary to abandon this
temporarily and to use ceramic oxide fuel, which is less satisfactory from the viewpoint of
the fuel breeding coefficient. There was a large reduction in the breeding factor because
of the neutron moderation at the oxygen and the reduced fuel density. An important task is
to increase the burnup, which is also important from the viewpoint of improving the breed-
ing rate. The larger the burnup, the greater the fraction of plutonium formed in that run
that is burned and correspondingly accelerates the breeding. The time spent in the external
fuel cycle is also economized, and there are benefits over the chemical processing and fuel-
pin manufacture. Table 1 gives some parameters characterizing the performance from exten-
sive burnup.
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At present, all these reactors:here and abroad are designed to use ceramic oxide fuel,
but studies on metallic uranium have not stopped, and it is currently evident that it should
be possible to attain the necessary fuel-pin parameters with metallic fuel. In that case,
one could provide a breeding factor up to 1.8-1.9, which would satisfy any concealable fuel
requirement in the future. If all the samethe transfer to metallic fuel is held up, and the
demand for fuel increases, one could use the principle of a heterogeneous (hybrid) core, i.e.,
simultaneous use of fuel pins containing oxide :or metal. From the physics viewpoint, this is
equivalent to removing some of the oxygen from the core, with the corresponding partial in-
crease in the breeding factor.
It is also desirable to use fast reactors to produce not only electricity but also heat
for industrial purposes. Experience In that respect has been acquired with the BN-350, where
part of the energy is used for desalinating sea water. Further research on this topic is
favored by the fact that fast-reactor stations produce high-potential heat (over 500?C),
which is very useful for some industrial technologies. Also, the industrial use of fast
reactors at constant mode is equivalent to a baseload operation, which is most favorable
from the economic viewpoint and which provides 'the largest amount of additional plutonium.
Many Comecon member-nations are interested in the fast-reactor researches. Those coun-
tries have collaborated in this area for 13 years via the Fast-Reactor Council. The collab-
oration has accelerated the resolution Of problems in designing large and economically viable
fast reactors. The program in he USSR envisages building the BN-800, which is essentially
a modified and ungraded BN-600, with the work then extending to the next generation with the
BN-1600. The basic reactor parameters are the following:
BN-800 BN-1600
Power,
thermal
electrical
No. of cooling loops
Coolant parameters in first
flow rate, ton/h
Temp. at heat exchanger inlet,
?C
Temp. at core inlet, ?C
2100
800
3
31 000
547
' 354
4200
1600
4
62 000
547
354
Coolant parameters in second
circuit:
flow through one steam gener-
ator, ton/h
inlet temp. for steam
generator, ?C
outlet temp. from steam gen-
erator, ?C
10
000
505
309
15 000
505
309
Feedwater temp. at inlet to
steam generator, ?C
210
210
Steam temp. at outlet from steam
generator, ?C
490
490
Steam pressure at outlet from
steam generator, MPa
13.7
13.7
Run time between reloads, days
120
150
Considerable attention is being given to the fast-reactor work in this country. There
are statements on the need to accelerate the work and to make early use of fast reactors in
the proceedings of the Twenty-sixth Congress of the CPSU and the plenary meeetings of the
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TABLE].. Effects of Burnup on High-Power
Reactor Characteristics (type BN-1600 with
oxide fuel) and Influence on the Fuel
Cycle
Characteristics
Maximum burnup. To of
heavy atoms
5
10
15 20
Reactor characteristics
Breeding factor losses:
from capture by fission
products
due to chemical repro-
cessing (loss in cycle 20/0 )
0,1133
0,047
0,067
0,025
0,101
o,of6
0,135
0,012
Overall losses in breeding
factor
0,080
0,092
0,117
0,147
Fuel cycle
Breeding rate (external
cycle 1. yr)
3,8
5,7
6,7
7,3
Specific volume of repro-
cessed fuel, kg/MW(el)?
yr
Specific volume of repro-
cessed plutonium, kg/
20,9
3,5
10,4
1,9
7,0
1,4
5,2
1,12
MW(ely?yr
Specific scale of fuel pin
production. items/
90
45
30
22
MW(el) ? yr
Central Committee of the CPSU. Soviet scientists and engineers, and workers and technicians,
and all who are involved with the problem will make efforts to carry out these resolutions.
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PAGES OF HISTORY
CRITICAL ASSEMBLY OF THE WORLD'S FIRST NUCLEAR POWER STATION REACTOR
M. E. Minashin UDC 621.039.524
Among the many questions and problems facing the developers of the world's first nuclear
power station and particularly its reactor, one of the principal ones must be designated the
determination and choice of physical characteristics of the reactor and also the study of its
behavior during runup, operation, transitional regimes, cooling, and in emergency situations.
All these characteristics can vary during a reactor run, as burnup of uranium and buildup of
plutonium takes place. Fission products are accumulated, the ratios of the number of nuclei
of fissile isotopes to the amount of structural materials and coolant vary and, consequently,
the intensity of competing processes of the interaction between the neutrons and these mater-
ials varies. As a result of all these changes, as a rule, the effective neutron multiplica-
tion factor also varies, which determines the control ability of the reactor.
The determination of the nuclear-physics characteristics of reactors of selected design
and power, as is well known, starts with the determination of the amount and enrichment of
the uranium charge in order, in the final count, to ensure an acceptable duration of opera-
tion (or the amount of power produced from the reactor) and to have satisfactory character-
istics with respect to controllability and the necessary economic indices.
When designing and constructing the world's first nuclear power station, the questions
of economic operation of the station were not studied in detail, since the design did not pur-
sue an economic goal, but the problem of the proof of the feasibility of construction of the
nuclear power station and its operational capability was resolved. Only after startup and
during further work on the creation of more powerful nuclear power stations were investiga..
tions on the economics and conditions for achieving economic competitiveness initiated. Just
as for the first nuclear power station, so also for subsequent stations, the main attention
was paid to questions of the operating reliability of each unit of the plant. The choice and
development of designs, including such new elements of the reactor as the fuel channels (FC)
and fuel elements, including the choice of the type of uranium fuel for it, were subordinate
to a considerable degree to this problem. Prior to the contruction of the first nuclear
power station, there was no design of fuel channel and fuel elements suitable for operation
at the high temperature and high pressure of the coolant, nor even data published about this.
The construction of the fuel channels and fuel elements was the most difficult problem in
designing the first nuclear power station. In addition to the physicists' calculations of the
reactor, experiments were also conducted on the assessment of the suitability of the alter-
natives being proposed by the designers and technologists for the designs of the fuel chan-
nels and fuel elements with different types of uranium fuel. When calculating the reactor
alternatives, difficulties arose in the manufacture of graphite components of large dimensions
for the reflector. However, these and many similar calculations associated with the choice
of designs were only tasks of an incidental nature, whereas the search for designs of the fuel
channels and particularly the fuel elements was conducted constantly right up to the end of
1953. Together with the use of tubular fuel elements, during planning the possibility of us-
ing rod-shaped fuel elements was considered. But from the point of view of reliability, the
tubular fuel elements were assessed to be the most suitable, as they could be previously
tested on a thermal test-rig under load. Other positive properties of these fuel elements
also were noted. Because of the nonmastery in 1951 of zirconium production for the channel
tubes, necessary for the use of rod-shaped fuel elements, the tubular type of fuel elements
was assessed to be the main alternative. Towards the middle of 1953, teat-rig trials of new
samples of tubular fuel elements, which had been started in March 1953 by a group of special-
ists under the direction of B. A. Zenkevich, were mainly completed, and for which a uran-
ium-molybdenum alloy dispersed in magnesium was used, developed by a group of specialists
under the direction of V. A. Malykh. These samples were subjected to thermal tests with
loads of more than 2.3 Mire (2.106 kcal/m2.h). In the quantity of materials --nonproduc-
Translated from Atomnaya Energiya, Vol. 56, No. 6, pp. 382-386, June, 1984. Original
article submitted March 2, 1984.
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0
0
0
G
0
0
A.3
9
6
0
430
0?0000000
0
00?00?
*000
000000000
-0
0?04
9 0. o-,_Ae
a
:p:
?000004)00,
o
00
e e
0 *000*o
oe
?
0
0
0
0',
0
00
403000403V
*
? eh,
00000
i
1 1
?
1 ?
0-1
C-2
0-3
0-4
4175
Fig. 1. AMF reactor (plan view): 1) loaded fuel chan-
nels; 2) fuel channels loaded to 2/7; 3) gaps filled
with graphite samples; 4) unloaded gaps; 5) rod ab-
sorbers of scram system; 0) reflector.
Fig. 2. Fuel element (a) and fuel channel (b) of the
AMF reactor: 1) filling of U908 powder; 2, 3) 1Kh18NYaT
steel tubes, with diameter 13.4 x 0.2 and 9 x 0.4 mm,
respectively; 4) soldering; 5) steel sleeve; 6) water;
7) steel base with thickness 2 mm;- 8) confining ring of
tin plate.
tive absorbers of neutrons in the fuel elements and channels of this type (molybdenum, steel,
water) ? and in geometrical dimensions, this fuel element version differed little from that
used in April 1952. For this fuel-element version, taking account of the availability at this
time of-new values of the constants (resonance integrals of absorption in 238U, fission prod-
ucts, absorption cross-sections in molybdenum, etc.), new values were determined for the charge
and enrichment of the uranium (570 kg and 5%, respectively, instead of 600 kg and 3% in the
original design version). Although these values, just as the choice of the uranium compound
for the fuelelements considered in April 1952, underwent almost no changes right up to the
manufacture of a regular batch of channels, not one of the versions of fuel elements up to
382
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. TABLE 1. Volume of Materials and Gas Cav-
ities Per cm height of Cell, cm3/cm
Materials and cavities
Lattice pitch a, cm
14
20
4,84
184,5
4,84
384,5
1120
3,7
3,7
1Kh18N9T steel
1,84
1,84
Gas cavities
9,12
13,12
Total in cell
204
408
TABLE 2: Values of Constants Used in the
Calculations, for Neutrons with E0 = 0.025
eV, b
Material
va
?as28'i
as
(Ftr?
_
cm
20,
t
cnr?
C
?
?
0,0042
4,8
4,5
?
0,37
H20
Za -=-
?Itot''
.----- 0,021
=4,33
=2,35
235U
2,07
545
645
8,3
8,3
?
?
2"U
?
?
2,6'
8,3
8,3
?
?
1Kh18N9T
?

0,9
0,9
steel
=0,247
Note. For neutrons with E > 0.025 eV, oao
iliTTE. Here va is the number of secondary
neutrons per absorption event; of is the
fission cross section; aa is the absorption
cross section; Y
-a = Paa; It0t = PatOt is the
total cross section; p is the density of the
nuclei.
the end of 1953, however, could be considered as acceptable. This was explained by the fact
that attempts to manufacture test samples of certain versions of fuel elements ended in fail-
ure (these samples were not even subjected to thermal tests: their surface was covered with
pits, cracks, etc.) and the version placed as final still had not pressed reactor tests.
In view of the absence of a fuel-element version acceptable for the first nuclear power
station project, it was not possible to order an experimental batch for the testing of a
physics assembly. It was only in mid-1953, based on thermal tests (including also tests with
thermocycling) of samples manufactured by a technology developed under the direction of V. A.
Malykh, that confidence emerged that this type of fuel element was promising. Now the first
nuclear power station reactor project became more substantiated. All developments could be
refined. In June 1953, on the instructions of I. V. Kurchatov, the Commission of the Institute
of Atomic Energy was formed at the Physicopower Institute (FBI) composed of V. S. Fursov, G. N.
Kruzhilin, V. I. Markin, and S. A. Skvortsov, which examined the design data of the first nu-
clear power station reactor. In particular, the Commission noted that the physics calculations
of the reactor had no experimental verification and therefore could be erroneous.
Because of these comments by the Commission, verifying calculations were performed and a
report was issued. It was shown in it that according to the concepts existing at that time
about the constants, the margin of reactivity was sufficient for accomplishing the assumed
running period. At the same time it was noted that, taking account of the entirely possible
errors of the constants used in the calculations (and also the moderation length and diffu-
sion length, existing for a reactor of small dimensions, resonance absorption of neutrons in
2381J, etc.), due to inaccuracy of the cross-sectional values, the relatively large amount of
383
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water in the lattice of the reactor and the complex geometry of the fuel channels, there
might not be a reactivity margin above the calculated value of the running time (100 days).
On the other hand, it was not possible to increase, for example, the uranium enrichment,
since the number of rod-absorbers necessary for compensation of the excess reactivity, cal-
culated with consideration of the assumptions mentioned above about the accuracy of the con-
stants (and the estimated value of Ak.=:28%), would be increased significantly and an ex-
cessively expanded grid of rod-absorbers would be used. This implied that it would be
necessary to assign an additional number of cells for these rods and have coolant feed and
outlet lines to them from both the main system of heat removal and from the heat-removal
system from the control and safety rod channels, add to have feeder tubes for the cables of
the rod actuators, etc. All this concerned the design of the reactor and the control and
safety rod system as a whole. Therefore, it would be necessary to look for confirmation or
discrepancy of the calculated and factual data on examples of critical assemblies of other
reactors. Testing and verification of the procedures and constants used before and after
this period were carried out by means of a calculation of the RFT reactor of the Institute
of Atomic Energy and of other critical assemblies. The results obtained were quite good,
and this checked the designers before a new change of design of the first nuclear power
station reactor and its systems.
However, the results of the verification of the procedure and constants were obtained
as applicable to reactors with significantly different fuel-element designs. Therefore, the
question of accuracy of the calculations of the first nuclear power station reactor still re-
mained open.
At the beginning of 1954 in the Physicopower Institute, an assembly of experimental fuel
elements was received for physics models, which in their dimensions were similar to the fuel
elements of the first nuclear power station reactor. The evalution calculations showed that
this assembly of fuel elements (435 pieces) was sufficient for the construction of a uran-
ium-graphite reactor of zero power and of small dimensions, with fuel channels similar in
design to the channels of the first nuclear power station reactor. More detailed calculations,
performed with the participation of Yu. A. Sergeev, V. Ya. Sviridenko, and G. Ya. Rumyantsev,
confirmed this possibility. Z. M. Kurova, S. I. Shagalina, L. Yu. Dol'skaya, and V. M. Stroi-
kova also participated in these calculations.
The spacing grid for the rod-absorbers was also chosen. The actuators for the rods of
this test-rig were designed by G. N. Ushakov, and with the approval of D. I. Blokhintsev, the
reactor was built in the Physicopower Institute with the vigorous participation of A. K. Kra-
sin, B. G. Dubovskii, A. V. Kamaev, M. N. Lantsov, E. I. Inyutin, L. A. Matalin, etc. The
schematic form of the reactor (designated ANF) is shown in Fig. 1, and a cross section of ?a
fuel channel and fuel element in Fig. 2. Table 1 shows the quantity of materials occurring
per cm of height of the cell, and in Table 2 the values of the constants assumed in the cal-
culations are shown. The physics model was not an exact copy of the AM reactor, but in the
design of the fuel channels, cell dimensions and quantity of materials in it, it was the
closest to the reactor model by comparison with the one having been used earlier.
The distribution of thermal neutrons in the cell of fuel channels (0) and the thermal
neutron utilization factor 0 were computed on the basis of the solution of the diffusion
equation for each zone of the cell:
(1)
where Li and Di are the length and coefficient of diffusion in the zone i, respectively, and
qi in the density of sources due to moderation.
At the boundaries of the zones, the fluxes 0 were "joined" and the resulting fluxes were
Di di/dr.
The coefficient of resonance absorption escape was determined by the formula of M. B.
Egiazarov [M. B. Egiazarov, V. S. Dikarev, and V. G. Madeev, in: Session of the Academy of
Sciences of the USSR on the Peaceful Utilization of Atomic Energy. Conference of the Division
of Physicomathematical Science [in Russian], Academy of Sciences of the USSR, Moscow, (1955),
p. 53] which, after its transformation for a cell of the first nuclear power station reactor,
acquired the following form:
exp
n
F 4Vu
Y1+2.66
e
(2)
5.4 I-7e,
vc + 22.8 (17191120
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where F is the surface of the uranium turned towards the moderator (water or graphite); I
is the mean free path of neutrons in the uranium; 7 - 2dr = 2(1.3-0.9) = 0.8 cm; c
1022 is the relative density of uranium nuclei; Vu, Vu, and Vuo is the volume of uranium,
graphite, and water in the cell, cm'/cm; n is the number of fuel elements in the fuel chan-
nel, and y is the density of the water in the fuel channel.
The square of the neutron moderation length up to energy E in the lattice and in the
reflectors was calculated for a homogenized medium on the basis of a computation only of
elastic collisions (in view of the small concentration of heavy nuclei): ?
Bo
Va 9f(EO
T c3e9 (Et)) X { dEVE'
(E V hEtrk) kg2shVO 2112
E cell
k - h
where f(E0) is the distribution of fission neutrons with respect to energy 80;
1 (E0) dE0? e-E? sh 1/-2E,
dE0;
So-E? sh ja?E0 dEo
(3)
Esk pkask and Etrk = Pkatrk are the cross sections of scattering and transport of neutrons
by the nuclei of material of the kind k;: occupying a volume Vk in a cell of the core or in
the reflectors; r(Eo)/6 is the square of the mean free path of a fission neutron with energy
E0 before the first collision, as the average value of all mean free paths r;
0-28(E?)r Ea (E,) dr
71(E0) = o
6
Se
0
ER(E?)r 28 (Es) dr
The critical size for a given neutron multiplication factor K. (or K. with specified
dimensions) is determined on the basis of the formula of age approximation
where
(4)
(1+ x2L1) exp (xilrf) (5)
x2= / 12 12
H?1./ R-+-6/1 /
(6)
here H and R are the height and radius of the core, respectively, and go = 2.401.
The values of the equivalent additions to the size of the core due to the end reflec-
tors (6s) and the lateral reflector (6R) were estimated on the basis of the solutions of the
two-group diffusion equations of age approximation-, obtained on the assumption of an arbitrary
separation of the spatial realtions of the fluxes 0(r, z) = 0(r), O(z). For this, in 1951-
1952, a universal equation of critical sizes was developed and used for the first nuclear
power station reactor, suitable for 10 different geometrical shapes of one-dimensional re-
actors, at first for LI = Tf (in the reflectors), and then also for L2 = Tf.
The calculated data obtained by using these equations and formulas are given in Table 3,
whence it follows that with the number of individual fuel elements available in our arrange-
ment (435 pieces), a critical reactor with a lattice pitch of 20 cm could not be achieved,
but with a lattice pitch of 14 cm criticality should be achieved with 49 channels. The
physics model of the AMIE reactor achieved criticality on March 3, 1954, when 512/7 fuel chan-
nels were loaded with the presence of gaps in the lateral reflector. According to the ex-
perimental estimates, loading of the gaps with graphite reduced the critical mass to 50 fuel
channels. The results of a comparison of the calculated and experimental data indicated the
necessity for the use in the calculations of joining of the generation density of neutrons
moderated in the core and in the reflectors, with weighting factors averaged over the square
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TABLE 3. Calculated Values of the Physical
Parameters for the AMF Reactor
Parameters
Core, channels
with water
Reflectors
a = 14CITI a=2m
lateral end
T
K.
Z', cm
Ei-Ot, cm
CM2
TT, CM2
(al:,)-', cm
RERZirrI, "12
go
gl
go
A f a=14 cm
a-20 cm
cr
ncr
0,923 0,956 -- --
0,792 0,759 -- --
1,511 1,503 -- --
124 259 3260 3260
2,56 2,78 2,76 2,76
106 239,7 3000 3000
300 360 370 370
13,9 15,43 17,4 17,4
60,5 69 67,4 67,4
1,08 0,984 -- __
1,25 1,128 --
1,11 0;976 -- --
-- -- 32 28
-- -- 39 32
55,8 132 -- --
49 134 -- --
Note. Rcr is the critical radius of the core)
ncr is the critical number of fuel channels.
20
q? -- (7c)therrnal region
Ti
dT
Ifig 'IriG/TnE8 ; '72- 4
--1- nE ; li_ dtlote
e,
Tr 0
01 - If
ftlEirtEs
I 0 ' o
Symbols with primes refer to the reflectors.
of the moderation length, but not for 1 eV, as this was sometimes used in calculations prior
to 1954.
The experiments conducted on the AMP showed that by using the procedure and constants
assumed for the first nuclear power station reactor there were no large errors at least at
the start of the reactor run. However, agreement with the experimental data on the AMY test-
ring was obtained only for a cold reactor state. However, the first nuclear power station re-
actor was designed for operation at a high temperature, and there could be no such agreement.
At the start of the 1950s, experimental and theoretical investigations in the field of
nuclear reactors, both in the Soviet Union and abroad, continued to expand. In this period
reactor physics was rapidly enriched with new data about the values and behavior of the inter-
action cross sections of neutrons with the nuclei of different substances in different neutron
energy ranges, and the values of the yield of secondary neutrons as a result of absorption
were refined, etc. Under the influence of these new data, in May-June 1954, the procedure for
reactor calculations of the first nuclear power station (vainly in the part of the calculation
of the effects of absorption in the region of moderated neutrons) and the constants were re-
fined. These refinements were caused by the necessity for detailing the neutron multiplica-
tion process, in order thereby to allow for the difference in the capture and fission cross
sections in the thermal and epithermal energy regions, and to clarify the influence of other
effects. From a comparison of the method of calculation of the first nuclear power station
reactor [D. I. Blokhintsev, M. E. Minashin, and Yu. A. Sergeev, Atomnaya inergiya, 1, 24
(1956)], carried out in the second half of 1954, with that used for the AMY test-rig and the
first nuclear power station reactor in the first half of 1954, it follows that besides the cal-
culation of the epithermal absorption, a method was recommended for determining the cutoff en-
ergy between the thermal and epithermal regions, partially used even at the present time.
Present-day calculations differ first and foremost in that the thermal energy region is div-
ided into several intervals. The calculations are performed on a computer, which allows many
details of processes to be taken into consideration which previously were either unknown or
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could not be considered. It must be mentioned that the fundamentals of knowledge concerning
reactors, accumulated by many Years Of work by the scientists of the Soviet Union headed by
I. V. Kurchatav, have remained unchanged Since the time of planning of the first nuclear
power station despite the Complexity of the-methods. Our debt and the debt of subsequent
generations consists in simultaneously displaying those integral effects enumerated in
Table 3, for all the complexities and refinements of the physical characteristics, which the
physicists of previous generations obtained at the cost of an enormous expenditure of labor.
387
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ARTICLES
HEAT ACCUMULATORS AT NUCLEAR POWER STATIONS
V. M. Chakhovskii UDC 621.355
From the early days of nuclear power, the equipment and fuel for nuclear power stations
have been designed for baseload operation, which is economically justified by the large cap-
ital investment in nuclear power stations and the requirement to maintain high rates of pro-
duction for secondary nuclear fuel. However, the conditions in the unified power system (UPS)
for the country have become such that the scope for Increasing the control range in the out-
put power from traditional plants arevirtually exhausted. It has therefore become necessary
to provide essentially new economic and technically accessible facilities. The traditional
approach here is to devise load-following nuclear power stations based on specialized fuel
pins and equipment stable under repeated thermal cycling, which is the method to be applied
when other and more economical methods of covering load variations have been exhausted.
The engineering and technological problems in designing load-following nuclear power
stations are so complicated that another approach has become necessary. This is to incor-
porate heat accumulators (HA). The main purpose of HA is to retain a high power use factor
(PUF) in the nuclear power plant (NPP) when the station works under varying load. There are
papers [1.-4] on the working conditions in nuclear power stations containing HA, which give
the essential schemes and designs. At present, the development of HA is taking various lines.
Table 1 gives the characteristics of HA for daily load graph control (HA of hot-water
accumulator type (HWA), double steam-water accumulators (DSWA), phase-transition ones (PTA),
ones containing chemicals (CA) or organic materials (OA)) and also for weekly load graph
control (HWA, DSWA, PTA, and CA). Here we present results on research on nuclear power
stations containing HA of HWA, DSWA, and PTA types.
Methodological Aspects of Efficient HA Use. The nonuniformity in the load graph for a
power system makes it necessary to provide the optimum types of load-following system. When
one combines load-following and baseload systems, one employs principles for equalizing the
production of load-following and baseload power at the level of the station or system.
To make it possible to use nuclear power stations with HA in accordance with load-curve
requirements, one has to consider various ways in which they can participate in the variable
part of the load curve. When one determines the relation between the accumulated heat and
output power from an HA, one has to know the load graphs for working and other days, together
with the control characteristics of the system equipment. These data are used in establish-
ing the load coefficients and the power increment in relation to the power of the reactor,
which is taken as the nominal load of NPP with HA. To evaluate the participation of nuclear
stations in power regulation and to choose the optimum equipment composition for the UPS, we
use a modified form of model, whose basic concepts have been given in [5].
Two approaches can be used in determining the costs at nuclear stations with HA. Firstly,
one identifies the components and systems in the HA loop (peak power) that provide for work-
ing the station with a variable load, and the costs for this system are referred to the addi-
tional energy production (peak) during the hours of maximum load (HA discharge), when the
turbine power is above the nominal reactor power. On the other hand, the cost of the basic
equipment is referred to the total power output less the peak output. In the second approach,
the costs of a nuclear power station containing HA include those for the peak system and are
referred to the total power production. In both approaches, the total costs are compared
with those for alternative load-following and baseload systems. The forms may be compared
separately on the production of load-following and baseload electricity. The alternatives
may be nuclear power stations with load-following fuel pins, nuclear power stations with
pumped-storage stations (PSS), nuclear power stations with gas-turbine systems (GTS), and
other combinations.
Translated from Atomnaya tnergiya, Vol. 56, No. 6, pp. 389-396, June, 1984. Original
article submitted February 28, 1983; revision submitted February 24, 1984.
388 0038-531X/84/5606-0388$08.50 1984 Plenum Publishing Corporation
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TABLE 1. Basic HA Parameters
Parameter
HA type
HWA IDSWA
PTA
CA
OA
Working body Water
Working pres- :0,1-2,54,0-7,0
sure, taa
Thermal ener- 170-.600
gy capacity.
&U/nis
Working temp Ao 200
t
Elect. energy 15-45
capacity pfP1A
kW .h/m'
Cost of unit H 50-250300-500
volume,
hibles/m3
HA cost, 40-200
rubles/kW
Cost of stored 0,8-2,81,2-3,6
energy. ko- .
pecks/kW .11-
Water
670-800
JO 280
55-65
60-250
Salt
0,1
300-
3030
JO 1000
30-300
25-600
80-340
1,35-
3,85
' acti ? ?
h ? ?
cemi-
cal mat
ter
0,1-4,00,1-1,5
500-
5000
,Ito 1500
60-600
200-900
100-400
1,5-4,40,8-3,2
Organic
matter
100-300
Ao 400
10-30
60-200
50-260
*The range in stored energy cost in HA is
dependent on the mode of operation in the
power system.
An important condition for the nuclear station is to provide the maximum PUF, and con-
sequently high rates of secondary nuclear fuel production (under conditions of a closed fuel
cycle). Figure la shows a possible integral daily load graph for nuclear power stations in
the UPS in 'relative units, which is characterized by pronounced load peaks. With this graph,
the load on the nuclear power stations should vary continuously, which involves ongoing
changes in the technological process, which firstly are complicated to provide and which
secondly reduce the station reliability. Therefore, in order to follow the graph for the
real load (real graph) one is justified in organizing each nuclear station on the basis of
a simple graph with the minimum number of load changes, for example as shown in parts b and
c of Fig. 1 for nuclear stations without and with HA. The daily power outputs from the real
and simple load graphs should be identical, as should be the conditions for running at the
minimum and covering the maximum load, namely in other words the graphs should have equal
areas, equal maximum heights, and equal minimum depths. Figure lc shows a simple load graph
for a station with RA in which the nominal load on the NPP corresponds to the mean daily
nominal load in the real graph.
The following condition must be met for the station in the system when a set of simple
load graphs As drawn up: the station must provide a given load Pkv and therefore produce
power during time interval v, including the time of minimum load and the time of maximum:
wik?Ni= Pkv, (k 1, n; v = 1, V), (1)
where Tikv is the relative load on nuclear power station j on day k during interval v, the
number of these being V when a station works n days a year, while Ni is the nominal NPP power
at nuclear station j, the total number of these in the power system being J.
The PUF for NPP with and without HA may be defined by the following for identical daily
power outputs on the working and other days of the week with equal morning and evening peaks
in the load with the nuclear stations operating on a simple load graph (subscript j is omit-
ted):
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10
10
Somin
-
nd
0
402m
Orm 1-1 -11
rin7,
7.2 _
nd
nd 4?2mr1-1
r wim rl
, nd
wine/7
0
12
1,0
492m
r--7_
I I
II I I
,j 1 .1 V in 1
I I
L-
__-___,_ _-_\7____
C/8 I 1
I nd , ,
I
40,,,, nd ; '. 1m :
I/3min 0-1 glin i L
____r__
0 2 -nd
9
I
0.15-0.2 M, 1.e, up to compositions of the aqueous system in which Pu(IV)
exists in monomeric form.
As already noted, in describing chemical equilibrium, it is necessary to use the activity
coefficients of individual ions (but not the stoichiometric activity coefficients). On the
basis of an analysis of published data on the activity coefficients, the transfer numbers,
the electrical conductivity, and the density of the aqueous solutions of electrolytes, phe-
nomenological equations were proposed in [1, 2], allowing the activity coefficients of indi-
vidual ionsin aqueous solutions of 1,1-, 1,2-, 1,3-, and 1,4-electrolytes and their mixtures
' 411
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TABLE 3. Equilibrium Constants of the Re-
action in Eq. (2) at.25?C
K
K '
K:
K;
Diluent
ange?of ap-
plication of
ULS-2--)--
v)
m(-1 ?
11
j?o
;'
;V-.1
:1,:2
L.:3
1IN03, Till*
IA vol.)
Zr
0,6
14
5
?
Kerosene
< 6
< 60
Zr
0,8
24
7
?
o-Xylol
< 6
< 60
Pu
190
41750
1105
0,4
n-Alkanes
< 5
10-40
U
300
22000
?
?
Kerosene
< 3,5
20-30
Th
150
?
?
?
Kerosene
< 3,5
20-30
to be calculated sufficiently correctly over a broad range of concentration of components of
the salt background. The equations of [1, 2] are based on a qualitatively new approach to
the phenomenon of ionic hydration, developed in [13]. The values of the parameters of ionic
hydration p are given in [2]; from th experimental data of [4], values p.. = 0.047 and 0.039
are obtained for the ions Pus+ and N2115, respectively. In using the equations of [1, 2] for
the identification of the processes occurring in aqueous solutions of Pu(IV), and determining
the thermodynamic equilibrium constants of the reactions of hydrolysis and complex formation
and the distribution, the limited number of assumptions given in the notes to Tables 1 and
2 is employed.
On the basis of thermodynamic analysis of the present data, and those in the literature,
on the relations between composition and properties, for the nitrates M = Zr(IV), U(IV),
Np(IV), Pu(IV) it was concluded in [10] that the ions M(OH)1.-i (i = 0-3) do not react with
NO; ions in the first coordination sphere, and the nonideality of aqueous nitrate solutions
of M(IV) is due to the change in structure and extent of their hydrate shells with change in
composition of the solutions. This conclusion, important for both theory and practice, was
later confirmed in many works [9, 14].
Using the method developed in [1, 2], the thermodynamic constants of hydrolysis of M41-
ions shown in Tables 1 and 2 have been calculated, together with the constants of complex
formation of M(OH)1- with certain organic and inorganic adducts in aqueous nitrate solutions.
It follows from Table 2 that, in complex-forming reactions, the ions which react with the
organic ligands are not teat but M(OH)7:11, which have a considerably weaker hydrate shell be-
cause of the reduction in charge and increase in radius of the hydrolyzed ions in comparison
with M4* [10, 15].
aa
In [16], the thermodynamic characteristics of the system Pu(IV)-HNO3-H20 were given.
It was established that, with increase in [HNO3] and [Pu(IV)], there is a significant change
in the activity coefficients of the particles Pu(OH)1-1 (i = 0-3), II+, NO3 in the system,
the yield and distribution of hydrolyzed forms of Pu(IV), the concentration of H+ ions (es-
pecially when [HNOs] < 1 M), and the degree of dissociation of HNO3; the method of calculat-
ing the degree of dissociation, taking account of the activity coefficients of nondissociated
HNO3 molecules, was outlined in detail in [1]. It follows from the data of [16] that the
description of chemical equilibrium with variation in the Pu(IV) concentration in the solu-
tions from 1-2 g/liter (fuel elements of water-cooled-water-moderated reactors) to 20-30 g/
liter and more (fast-reactor fuel elements) both when [HNO3] 0 const and when [HNOs] = const
is impossible without taking account of the change in the given parameters as a function of
the state of the solution. It is more significant that the use of the method of [1, 2] al-
lows the chemical reactions occurring in complex multicomponent - and heterogeneous - systems
to be identified.
On the basis of the method of [1, 2], using the results of the present. work and litera-
ture data [7], it may be shown [10, 18] that the extraction of M(IV) by tributylphosphate
(TBF) solutions occurs according to the equation*
*Experimental proofs are given in [10].
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TABLE 4., Comparison of Experimental [20,
21] and Calculated Data on the Extraction
of Pu(IV) at 25?C
Expt.
Cale.
IHNO31/11/,
M
[Put,. n,
g/Iflar
woo,
g/riter
(rub.
g/liter
0,52
1,41
1,58
1,73
0,53
4,28
4,79
4,79
0,52
13,6
17,1
15,9
0,51
0,51
19,6
33,5
21,2
37,7
21,2
36,7
0,99
0,63
1,72
1,83
0,99
2,27
5,8
5,55
1,00
5,73
14,7
13,3
1,01
8,3
19,2
19,2%
1,20
13,8
30,0
33,4
3,89
0,014
0,34
0,29
3,98
0,129
2,27
2,6
3,94
0,145
8,6
8,4
3,93
0,98
16,7
16,8
4,01
1,77
25,0
29,8
TABLE 5. Comparison of the Calculated and
Experimental [19] Values of the Distribution
Coefficients of Pu(IV) in the System HN0s-
Pu(IV)-RC0011-1120-30% TBF Solution
DIN031ant
M
MCOOHJa-,
g
m
DPu
DIWOOlii
t dau of
? - -
calc ?
____
data of
......... _
,.t,
0,246
0,49
0,17
0,15
0,49
0,46
0,246
1,01
0,084
0,088
0,37
0,37
0,246
2,02
0,043
0,042
0,27.
0,27
0,246
3,02
0,024
0,024
0,23
0,22
0,56
2,05
047
0,18
0,27
0,26
0,56
3,1
0,12
0,105
0,22
0,24
0,8
1,0
0,74
0,88
--
0,35
0,8
2,0
0,48
0,44
--
0,26
0,8
3,0
0,3
0,27
,-
0,21
1,05
0,56
1,7
1,81
0,36
0,38
1,05
1,09
1,24
1,25
0,30
0,32
M(01-14-i) + (4- ONO; +2TBF
;2: MODIIMNT3)i2 TBF
where i = 0-3. The values of e for the corresponding reactions (Table 3) are not strictly
thermodynamic, since they are calculated by an empirical method, using the parameter git from
Eq. (3) to take account of the nonideality of the organic phase [1, 2, 10, 18]. Nevertheless,
the approach to the interpretation of extractive equilibrium developed in [1, 2, 10, 18] gives
a sufficiently rigorous quantitative description of the distribution of M(IV) (and other
metals) in multicomponent heterogeneous systems, since the empirical Eq. (2), taking account
of nonideality of the organic phase, is obtained on the basis of statistical analysis of num-
erous literature data [17] on the distribution of uranyl nitrate and M(IV) in the systems
140-extractive agent-HNOs-salting-out agent-TBF-diluent, taking account (this is most im-
portant) of nonideality of the aqueous phase [1,2]. The validity of Eq. (2) hasbeenprovenfor -
many examples [2, 6, 10, 15, 18].
The given approach to the interpretation of equilibrium in homogeneous and heterogeneous.
systems has been used to develop a mathematical model of the distribution of Pu(IV) and HNO3
in reflux processes of the purification of plutonium in the regeneration of spent fast-reactor
fuel.
(2)
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a
2
9
12
13
14
17
c
18
19
27
Fig. 1. Diagram of the extractional purification
of plutonium in nominal conditions: a) spentex-
traction agent > 1) appear in the moments of the distribution. The first two moments calculated
by taking account and not taking account of nuclear interactions differ appreciably in spite
of the small value of y = 10-' [9]: (A = Zion)/Zion = 0.01; (A2-A10) /ZI on = 0.55, where
A/4 =h
AlionA,k= 1,
' f d 2. Figure 1 also shows that the distribution.of AE differs
greatly from the distribution of the energy lost by a fast particle in the layer, taking
account of elastic nuclear scattering. This is a result of the high probability of the escape
of scattered protons from the layer. The difference between the AE and the energy loss dis-
tributions, which results from the escape of secondary particles from the volume of material,
indicates a substantial correlation of energy release processes in adjacent layers of the:ma-
terial. This fact probably muet be taken into account in processing measurements in-multi-
layer ionization detectors. It is natural to expect that analogous effects should occur -
also for inelastic nuclear reactions of high-energy particles, and also for more complicated
bombarding geometry, for example a "layer in matter" [8].
(4)
tFor simplicity we assume they are all of the same kind as the primary particle. The gener-
alization for particles of different kinds leads to a more cumbersOme notation for the last
term in Eq. (2).
420
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LITERATURE CITED
1. V. A.Lyubimov, in:ElementaryParticles [in Russian],No. 2,Atomizdat,Moscaw(1980), p.91.
2. V. I. Ivanov and V. N. Lystsov, in: Problems of Microdosimetry [in Russian], inergoatom-
izdat, Moscow (1982), p. 3; I. M. Dmitrievskii, ibid., p. 88.
3. A. G. Alekseev et al., in: Third All-Union Conf. on Microdosimetry, Moscow Eng. Phys.
Inst. (1979), P. 122; L. N. Zartsev, Physics of Elementary Particles of the Atomic
Nucleus [in Russian], Vol. 11, No. 3 (1980), p. 525.
4. L. D. Landau, Collected Works [in Russian], Vol. 1, Nauka, Moscow (1969), p. 482.
5. P. V. Vavilov, Zh. Eksp. Teor. Fiz., 32, No. 4, 920 (1957).
6. N. P. Kalashnikov, V. S. Remizovich, and M. I. Ryazanov, Collisions of Fast Charged Par-
ticles in Solids [in Russian], Atomizdat, Moscow (1980), p. 61.
7. S. G. Andreev, Author's Abstract of Candidate's Dissertation, Moscow Eng. Phys. Inst.
(1981); S. G. Andreev and N. P. Kalashnikov, in: Third All-Union Conf. on Microdosi-
metry [in Russian], Moscow Eng. Phys. Inst. (1979), p. 12.
8. S. G. Andreev, in: Shielding Against Ionizing Radiations of Nuclear Industrial Plants.
Proc. Third All-Union Scientific Conf. on Shielding Against Ionizing Radiations on
Nuclear Industrial Plants [in Russian], Tbilisi State Univ? Vol. 3 (1983), p. 15.
9. V. S. Barashenkov, Interaction Cross Sections of Elementary Particles [in Russian],
Nauka, Moscow (1966), p. 15.
STABILITY OF SCINTILLATION DETECTORS VIS-A-VIS RADIATION
V. V. Pomerantsev, I. B. Gagauz, UDC 539.1.074.3
Yu. A. Tsirlin, and O. V. Levchina
Scintillation detectors which are used to monitor background radiation or radiation
in geological prospecting or other research work are in a continuous flux of y-radiation.
We consider in the present work the influence of I57Cs y-radiation (Ey = 0.662 MeV) of
various flux densities (105, 106, and 107 quanta?cm-2?sec-4) upon the spectrometric char-
acteristics of detectors of the most frequently employed type (diameter 40 X 40 mm) with
various activator concentrations. The absorbed dose was constant and amounted to ru0.1 Mrad
(1 rad = 0.01 Gy). The detectors were supplemented by 2"Am a sources (Ea = 5.486 MeV) with
a flux of 400 particles.sec-I to study the a/y ratiowhichinthe crystaldevelops from the
light yield of the a line on the energy scale of the y radiation subsequently termed "y
equivalent").
We irradiated in our experiments 37 detectors in lots of 5-10 detectors; the detectors
had the standard activator concentration of 10-2 mass %. The influence of the radiation was
assessed from the relative changes of the light yield on the 447CS y line (dCy) and the y
equivalent (Ca/) and on the basis of the changes in the intrinsic a and y line broadening
of the detectors (ARa and ARy). The average changes and the greatest changes in the param-
eters at the probability level P = 0.95 are listed in Table 1.
Since the experiment lasted for a long time (3000 h at a flux density of 105 quanta.
cm-2.sec-I), changes in the characteristics of three detectors which had not been irradiated
were monitored. All detector characteristics remained constant within the error limits of
the measurements; more specifically, the variation in light yield did not exceed ?1.5%. Fig-
ure 1 illustrates the changes in the light yield obtained via the y line in three lots of de-
tectors. It follows from Fig. 1 and Table 1 that the detector light yield decreased (by
about 40%) in ally radiation fluxes considered; the average change Of the resolution of the
y line was significant when the flux density of the irradiation was 105 and 107 quanta.cm-2.
sec' and statistically probable at a flux of 106 quanta-cm-2.sec-I.
This means that y radiation produces in the crystal volume damage which may change the
spectrometric characteristics of the detectors on the y line. The statistical evaluation
of the data concerning the change of the y equivalent in individual sets of detectors shows
' that changes of the y equivalent are either probable or statistically significant. But only
Translated from Atomnaya tnergiya, Vol. 56, No. 6, pp. 415-416, June, 1984. Original
article submitted October 24, 1983.
0038-531X/84/5606-0421$08.50 1984 Plenum Publishing Corporation 421
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TABLE 1. Influence (%) of y Radiation
Upon the Detector Parameters
No. of irradiated detectors
106
1064
101
26
6
5
Relative change
Change in the in-
irinsit detector reso-
lution
light yield
oC (meas.
error ? 5%)
y equi-
valent
oCcvey
(meal.
error ?3%)
4612
(meas.
error
i5/)
bRa
(meas.
error
?0.5%)
!,4.1
?16,1
?28,(
?3,4
1,2
1,7
0,0
0,3
?39,2
?47,1
3,5
5,2
0,3
0,8
0,2
1,0
?39,9
?43,6
5,9
6,5
1,1
1,7
0,4
1,0
10 20
Time, h
Fig. 1
110
100
.90
110
a
1 nv
> 130
! 120
110
100
.90
10 20 JO
Time, h
Fig. 2
Fig. 1. Dependence of the detector light yield of the 157Cs y line upon
the time of irradiation at identical irradiation doses and flux densities
of a) 105, b) 106, and c) 10' quanta.cm-5.sec-1. In this figure and in
Fig. 2, the numbers at the curves denote the detector numbers in each lot.
Fig. 2. Dependence of the y equivalent of the detectors upon the time of
irradiation for a particular absorbed dose and flux densities of a) 105,
b) 106, and c) 10' quanta-cm-5.sec-I.
at a flux density of 107 quanta-cm-5.sec-1 is the measurement error exceeded by the greatest
possible change of the y equivalent at the probability P = 0.95 (see Table 1 and Fig. 2).
Thus', the ah ratio of the crystals is practically not affected by gamma radiation with
a flux density below 10' quanta.cm-5-sec-1, Accordingly, when detectors are operated in con-
tinuous y-radiation fluxes with a density not exceeding the above-indicated value, the y lines
under inspection must be identified with the aid of the y equivalent. ?
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A change in the resolution is at the a line statistically insignificant for all y ra-
diation fluxes employed. The greatest possible change at the probability P = 0.95 does not
exceed the measurement error (?0.5%), i.e., this parameter is also stable. Accordingly,
the a linecanbeused to stabilize the measuring channel ofa scintillation y spectrometer op-
erated in a y-radiation field.
The light yield of the detectors is restored after y irradiation in individual form and
takes place in more than 1 month but, in the majority of cases, takes one year and more, i.e.,
the restoring processes can compete with aging [A. A. Burshtein et al., Metrologiya, No. 2,
59 (1981)].
Similar results concerning the influence of weak y radiation upon detector characteristics
were obtained with activator concentrations of as high as 3.10-1' mass % in a single crystal
(18 samples). A further increase in concentration results in coloring of the crystals and a
substantial change of all detector characteristics by radiation. A decrease of the concentra-
tion to 5.10-3 mass % increases the stability of the detectors.
The authors thank E. A. Bugai, L L. Vinograd, L. S. Zubenko, V. F. Lyubinskii, and A.
P. Meshman for their collaboration and help in the present work.
? RADIATION STABILITY OF SCINTILLATING POLYSTYRENE
I. B. Gagauz, A. P. Meshman,
V. F. Pererva, V. V. Pomerantsev,
and V. M. Solomonov
UDC 539.1.074.3
Scintillating polystyrene is widely employed in 0 radiometry. In some applications,
for example in monitoring the contamination level of various objects, the behavior of the
detectors must be checked while the dose absorbed during the detector operation increases
cumulatively.
We consider in the present work industrial polystyrene samples of standard composition
(2% PT and 0.1% POPOP) exposed to the radiation of a 0 particle source consisting of "Sr +
'Y with an average energy Esr = 0.196 MeV and ty = 0.936 MeV [1] and the activity A = 9.25*
10' Bq.
The rate of the absorbed dose is
(1)
where N denotes the flux of 0 particles with the energy E, which are absorbed in the scin-
tillator; V denotes the volume of the irradiated scintillator mass; and p denotes the density
of the scintillator material (p = 1.06 g.cm-3).
The irradiated volume was given by the sum of the volumes of 1) a cylinder of diameter
d = 3 cm (equal to the diameter of the active spot of the source) and a height equal to the
range / of 0 particles of average energy E, and 2) the body formed by rotation of a curve
around the cylinder axis:
2 cot a+ cos a; sin cx,
where h denotes the distance between the source and the sample (h = 0.02 cm); and a denotes
the angle under which a 0 particle from the edge of the source is incident on the sample.
The curve connects the ends of the trajectories of 0 particles which leave from the edge
of the source surface under an angle a (amin a