SOVIET ATOMIC ENERGY VOL. 45, NO. 6

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• Publication Date: June 1, 1979
• Content Type: REPORT

..~' A. ~.
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v~-.~
`_
Russian Original Vol. 45, No. 6, December, 1978
,June, 1979
SATEAZ 45(6) 1147-1256 (1978)
SOVIET-
ATOMIC.
ENERGY
ATOMHAfI 3HEP~NH
tATOMNAYA ENERGIYA)
TRANSLATED FROM RUSSIAN
r
CONSULTANTS BUREAU, NEW YORK
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- {
Soviet Atomic Energy is acover-to-cover trans{ation of Atomnaya
Energiya, a publication of the Academy of Sciences of the USSR:.
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- Editorial Board of Atomnaya ~nergiya:
Editor: 0. D. Kazachkovskii
Associate Editors: N. A.-Vlasov and N: N. Ponomarev-Stepnoi
SOVIET
ATOMIG
ENERGY
I. N. Golovin
V..I . ~I I'ichev
V. E. Ivanov
V. F. Kalinin
P. L. Kiri lov
Yu. I. Koryakin
A.K. Krasin '
E. V. Kulov
B. N. Laskorin~
V. V, Matveev
,I. D. Morokhov
A. A. Naumov
A. S. Nikiforov
A. S. Shtan' "
B. A. Sidorenko
M. F. Troyanov
E. I. Vorob'ev
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SOVIET ATOMIC ENERGY
A translation of Atomnaya Energiya
June, 1979
Volume 45, Number 6 December, 1978
CONTENTS
'Engl./Russ.
ARTICLES
Prospects of the Utilization and the Basic Problems of the Introduction
of HTGR in Technological Processes and Electric Power
Generation - V. A. Legasev, N. N? Ponomarev-Stepnoi,
A. N. Protsenko,.Yu. F. Chernilin, V. N. Grebennik,''
and A. Ya. Stolyarevskii .............. ............. ... ........ 1147 411
Comparison of Calculations for a Standard Fast Reactor (Baker Model) ?
- A. I. Voropaev, A. A. Van'kov, and A. M. Tsybulya .... .. ............ 1155 4i9
Power-Reactor Fuel-Production Problems - F. G. Reshetnikov,
Yu. K. Bibilashvili, and V. I. Kushakovskii ......................... 1162 426
Zirconium Alloys in Nuclear Power - A. S. Zaimovskii ........................ 1165 430
Swelling of Steels and Alloys Irradiated in the BOR-60 Reactor
to a Fluence of 1.1 ?1023 Neutrons/cmz - N. P? Agapova,
V. S. Ageev, I. N. Afrikanoy, N.I. Budylkin, V. A, Krasnoselov,
E. G? Mironova, V. D. Onufriev, Z. E. Ostrovskii, V. I. Prokhorov,
and Yu. N. Sokurskii . .. .. .... ........ ..... .... 1169 433
Measurement of the 2s7Np/23ePu and 2?~Am/23sPu Fission Cross. Section
Ratios for 0.13-7.0 MeV Neutrons - V. M? Kupriyanov, B. I. Fursov,
V. L Ivanov, and G. N. Smirenkin .. , . ....... ..................... 1176 440
? ...
Experimental Research on the Production and Storage of Ultracold Neutrons
- V. I. Morozov ..... ... ..... ......... .... .... ..... 1179 442
... ..
Cyclotron X-Ray Spectral Analysis of the Elements from Titanium
to Cesium - V. A? Muminov, R? A. Khaidarov, and Kh. Isaev ............ 1185 449
LETTERS
Measurement of Neutron Total Cross Sections and Resonance Parameters
of 147Pm - V. A. Anufriev, T. S. Belanova, Yu. S. Zamyatnin,
A. G, Kolesov, S. N. Nikol'skii, V. A. Poruchikov, S. M. Kalebin,
1189 453
V. S. Artamonov, and R. N. Ivanov .. ....................... ...... .
Use of Small Deuteron Accelerator for Neutron-Activation Determination
of Fluoride - Yu. I. Bondarenko and V. S. Rudenko . , ........... ....... 1192 456
Radial Reactor Stability and Automatic Regulator - B. Z. Torlin, ........ ?... ?........ 1194 457'
Experimental Determination of Neutron Leakage from a Manganese Bath
- A. V. Sorokina, E ? A. Shlyamin, K. A. Petrzhak, G. E. Lozhkomoev,
A..G. Prusakov, .and Ya. M? Kramarovskii .... ...... ............ 1197 459
Possibility of Acoustically Detecting the Boiling of Sodium in a Fast
Reactor by Means of a Pulsed System - B. V. Kebadze; K. A? Aleksandrov,
and V. V.Golovanov ....... .. .??? ???? ???????????? 1200 461
ANNIVERSARIES
Nikolai Nikolaevich Ponomarev-Stepnoi ..... . ? ? ? ? ? ? ? ? ? ? ? ? ? ? ? ? 1.203 465
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. CONTENT8
cro~f~~~>
Engl./Buss.
CONFERENCES, MEETINGS, SEMINARS
Fourth International Conference on the Application of Zirconium in the Atomic
Industry-B. S,RodchenkovandA,N.Ivanov ,..,?.,,,?,,?,,,,,,,,?? 1204 466
All-Union Seminar on the Processing Technology for Ores, and Rare,
Dispersed, and Radioactive Elements - V, A. Pchelkin
and E. A.Semenova ,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,, 1207 468
Problems of Fuel Utilization in the Power Industry - B. E. Novikov , ,,, , , , , , , , , , , , , , 1209 469
International Meeting on Fast-Atom Injectors for Thermonuclear
Facilities-N.N.Semashko,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,, 1210 470
Soviet-Italian Seminar on Plasma Research in Tokamaks
- L. G. Golubchikov .............................................. 1213 472
Meeting of IAEA Experts of Prevention of Ocean Pollution
-L. I,GedeonovandV.M,Flegontov ,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,, 1214 473
Third International Conference on Collective Methods of Acceleration
-V. P, Sarantsev,,.,.,,, ,
. ..................................... 1216 474
M. Kh. lbragimov, V. I. Subbotin, V. P. Bobkov, G. I. Sabelev, and G. S. Taranov.
Structure of Turbulent Flow and the Heat Exchange Mechanism
in Channels -Reviewed by Yu. Klimov . ............................... 1219 476
Z. A. Al'bikov, A, I. Veretennikov, and O. V. Kozlov, Detectors of Pulsed
Ionizing Radiation -Reviewed by E, A.?Kramer-Ageev ....................... 1220 477
Yu. V. Gott. Interaction of Particles with Matter in Plasma Research
- Reviewed by A.I. Akhiezer ........................ ................ _1221 477
A. D. Frank-Kamenetskii. Modeling Neutral Trajectories in Monte Carlo
Calculations of Reactors -Reviewed by L. V. Tochenyi ............. ........ 1223 478
V. I. Davydov (editor), M. N. Gamrekeli, and P. G. Dobrygin.
Thermal Processes and Apparatus for Obtaining Oxides of Rare
and Radioactive Metals -Reviewed by I, G. Slepchenko ....................... 1224 479
Il~TDEX
Author Index, Volumes 44-45, 1978 ........................................ 1229
Tables of Contents, Volumes 44-45, 1978 .................................... 1235
The Russian press date (podpisano k pechati) of this issue was 11/24/1978.
Publication therefore did not occur prior to this date, but must be assumed
to have taken place reasonably-soon thereafter.
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ARTICLES
PROSPECTS OF THE UTILIZATION AND THE' BASIC
PROBLEMS OF THE INTRODUCTION OF HTGR'
IN TECHNOLOGICAL PROCESSES AND ELECTRIC
POWER GENERATION
V. A, Legasev, N, N, Ponomarev-Stepnoi,
A, N, Protsenko, Yu, F, Chernilin,
V. N. Grebennik, and A. Ya, Stolyarevskii
UDC 621,038.52,034,3
An extensive and comprehensive discussion of the various problems of the future development of power
generation has originated with the increasing difficulties for providing the Soviet Union with cheap power and
energy resources, for which the level of demand has reached significant scales,
The difficulties of the development of power generation are due to the constant price increase of organic
resources and, in the first place, of petroleum and gas, which are the most suitable and most widely used
sources of power, and for which reserves are limited; the necessity for taking ecological factors into con-
sideration, which make the production of power more expensive and carry additional limitations on the- scale
and disposition of power-generating stations; and also the large "inertia" of power generation -the most time-
consuming and investment-capacious branch of the economy, in consequence of which it is necessary to con-
struct prototype facilities long in advance of the widespread introduction of stations with new power-generating
trends into the fuel-Powerbalance (FPB).
All this requires an extension of the range of applicability of nuclear power generation in the fuel-power
structure of the country and its maximum utilization efficiency [1, 2]. Most important in the development of
power generation in the near future is the gradual replacement of liquid and gaseous organic fuels [1, 3, 4].
At the present time, nuclear power generation in practice is being used for the production of electric
power and is replacing the organic fuel necessary for its generation, The development of nuclear power gen-
eration in the European part of the Soviet Union will allow the deficit of energy resources in this region to
be reduced significantly, However, the functions of nuclear power stations in supplanting the deficient hydro-
carbon fuel are limited. As the functional power reactors are intended for use mainly in the base section of
the power-system loading, then their introduction will supplant the condensation power stations, operating on
coal, in the energy balance.
It can be seen from the data of Table 1 that the greatest demand on the fuel-energy resources, including
petroleum and gas, is for the generation of medium- and low-potential heat and steam, high-p~ential heat for
technological processes (metallurgy, chemistry, etc.), and for the provision of motor fuel to the national econ-
omy. In addition, gas-petroleum residue fuel is used in the production of peak and semipeak power.
The demand for gas-liquid fuels in some of these fields can be partially reduced by the use of nuclear
heat-supply stations (NHSS) in the production of low-potential heat in functional nuclear power stations in
certain technological processes, for the complex provision of low-potential heat and electric power, The
possibilities of using nuclear power based on functional reactors for the purpose of supplanting gas-liquid
energy resources are limited, Wider prospects are being opened up by the construction and introduction of
high-temperature gas-cooled reactors (HTGR),
The principal feature of HTGR is the production of heat at a temperature of ~ 1000?C and higher, This.
temperature allows the introduction of these reactors into various central-heating, power-technological and
other processes, and will allow deficient hydrocarbon fuel to be supplanted. in Tables 2-4 the possible fields
of application of HTGR are considered. The potential scales of development of HTGR are considered by the
example of the use of high-temperature heat for the steam conversion of methane and are given in Table 2.
*High-temperature gas-cooled reactor.
Translated from Atomnaya L`nergiya, Vol, 45, No, 6, pp. 411-418, December, 1978, Original article
submitted April 17, 1978,
0038-531X/78/4506-11=i7ti07,50 ?1979 Plenum Publishing Corporation 1147
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TABLE 1, Structure for the Requirement on Fuel-Energy Resources, %FEB
Period is7o-isso
Long-term$
Potential fraction of .
nuclear power
Field of requirement
-
raction o
petroleum
fraction of
total"
and gas (at
Total
petroleum
LWR' ?
HTGR
end of
~
g
and as
riod)fi
Generation of electric power
25
13
30-35
~ 10
p to20-25
~ Up to 25-28
Generation of heat and steam of
32
22
25-30
~ 20
Up to 8-10
Up to 25
medium and low potential.
Generation of high-potential heat
19
14
14-i6
10
-
Up to 12-14
Mobile and fixed power ,
18
14.
i6
15
~ -
Production
faci)ities ~
of synthe-
tic fuel
In the.chemical and petrochemical
6
5
10
~ 8
-
-
industries as raw material
Total _ I
100 I
68 I
100 I
- 60
*According to data of [5], average for the period,
'[According to authors' estimates, effected by taking account of the structure of the demand
for petroleum and~gas according to data of [5], and the scales of demand for petroleum and
gas i-n 1980 from the data of [6], _
$ According to authors' estimates, taking account of the main trends of change of the FEB
and forecasts, for example [3], -
* * Eight-water reactors (VV>~R and RBMK) [Water-cooled/water-moderated power re-
actor, and high-powered water-cooled channel reactor, respectively], _ -
About 20% of the total-organic fuel which will be extracted in 1980'should be consumed by the high-tem-
perature heat industry, Some 60-70% of the fuel consumed for this purpose comprises deficient hydrocarbon
fuel, and this fraction will not change significantly during the-next 15-20 years: High-temperature potential
is essential for the production of ammonia and ammoniacal fertilizers, synthetic alcohol, hydrogen, etc? and
also in metallurgy in the direct reduction of iron, blast-furnace processes, etc.
A key problem of the majority of high-temperature processes. is the production of the various reducing
agents, and particularly hydrogen, which can be obtained by means of HTGR, Hydrogen, as raw material; is
obtained from organic fuel in the' steam conversion of methane or the gasification of coal (see Table 2), In
the long-term, the thermochemical or thermoelectrochemical decomposition of water can provide an inorganic
source of hydrogen (Table -3). The production of reducing agents and, particularly hydrogen, by means of HTGR
will allow all the organic fuel consumed to be supplanted in this field of power generation,
Thus, the use of HTGR in power-technological factories will open up a wide prospect for the develop-
ment and use of nuclear power generation, which will have a positive effect on the solution of the problem of
protection, of the environment, _ _ - _
As mentioned earlier, the greatest portion of the extracted organic energy resources, including hydro-
carbons, is consumed in the production of medium- and low-potential heat steam. Part of the concentrated
consumers of low-potential heat can be provided by NHSS based on functional reactors, A high portion of.gas-
black oil fuel is necessary for decentralized and industrial heat supply with a small concentration of power
requirement; the transmission of which by coal is made-difficult in view of the technicoeconomic and ecological
reasons, One of the possible routes for solving this problem is opened by nuclear long-distance heat-supply
stations (NLDHSS), In this case,-HTGR are used for the steam conversion of methane with the transmission
of the cooled conversion products. (CO and HZ) (Fig, 1) through gas pipelines to the sites of heat requirement,
where the reverse methanization reaction with the release of-heat is carried out, The temperature during
methanization is ~ 450_650?C, As_a result, CO and HZ are converted almost completely into methane, which
can be returned-to the reactor through pipelines.
in all the schemes considered for the use of HTGR inpower-technological-processes, the use of part of-
the heat generated by the: reactor is intended for the generation of electric power, Owing to the high temper-
ature of the coolant in this case,=steam turbines can be used with. modern high=steamparameters (550?C and
170/240 bar) and net effective efficiencies of ~ 40%._ _ - - _
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 Declas C 700120001-4
sified and Approved For Release 2013/03/22: IA-RDP10-021968000
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0,5145
0,2864
0,559
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0,0143
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0,00075
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0,558
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0.00075
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1.85!
0,375
0,0456
0,566
0,283
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0,1313
0,587
0,278
0,0265
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0,0090
0,0016
0,00092
0,274
0.496
0,0265
0,0072
0,0087
0,0015
0,00093
0,535
0,511
0,274
0,489
0,0266
0,0071
0,0086
0,0014
0.00095
1,843
1,788
0,353
0,0458
0,132
5,350
1,167
0,134
0,274
0,0256
0,0116
0,0077
0,0014
0,00079
0,271
0,511
0,0254
0,0113
0,0073
0,0014
0,00079
0,511
0,576
0,270
0,501
0,0256
0,0111
0,0073
0,0013
0.00080
1,817
0,369
0,0485
3. The deviation of the critical-charge calculations for BNAB-M and ENDF/B N from the average cal-
culated values for KFK-iNR and CARNAVAL-N is 27 kg (~ 1,4% OKeff) in all variants, These values are
close to the difference between the BNAB-M and ENDF/B N calculated values and the experimental values
for the ZPR-6-7 assembly,
4, We observe a lower critical charge in the calculation based on FD-5 in variants A and B, For ex-
ample, the difference between the ENDF/B N and the FD-5 calculated values is 44 kg (~ 2,4% OKefg). How-
ever, in variant C, with high 240Pu content, the data for FD-5, CARNAVAL-N, and KFK-iNR are close to one
another, As Table 4 suggests, this is due to the high 240Pu capture cross section in the FD-5 system in com-
parison with the other systems of constants,
5. It follows from Table 3 that the dispersion of the excess reproduction factor and the physical repro-
duction factor in the BNAB-M, CARNAVAL-N, KFK-INR, FD-5, and ENDF/B N calculations is fairly small
(~ t 0,02), and the relation G ~ (B-1) +0,06 holds, The excess reproduction in the BNAB-70 calculations, as
could be expected [4, 5], is ^~ 0.06, The main contribution to the discrepancy comes from the difference in
the reproduction in the active zone,
6. The adjustment of the OSKAR-75 constants assumed what we consider an unduly high accuracy in
the reaction rates determining the neutron balance (primarily e$/fy (t 2%)) found in the experiment conducted
on the ZPR-6-7 large plutonium assembly, This probably is one of the reasons for the low reproduction in
the OSKAR-75 calculations, it has already been noted that the discrepancy between experiment and calcula-
tion on the basis of the KFK-INR, ENDF/B N, and BNAB-M constants amounts to 6-8%. It is also possible
that there was a systematic error in the experiment [21],
Table 5 shows the blocked group cross sections for Z~Pu and 2~U and the value of v, We observe a con-
sideiable dispersion of of (Z~Pu) (~ 15% in groups 14 and 15) and vc (238U) (~ 25% in groups 14 and 16), The
dispersion in Qc (z~Pu), in our view, is less than might have been expected, taking account of the accuracy
of the microscopic data.
We do not have atry analogous data from France and Great Britain, Therefore, it is useful to compare
.the active-zone averages of the cross section, which are easily obtained from the neutron balance (see Table
4), A comparison shows that the calculations of the integral spectra in the active zone are close to each other,
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TABLE 5, Blocked Group Cross Sections of the Active Zone (variant A)*
iJO. Of
V (23BPll)
of (2sspu)
Q~ (23BPU)
Q~ (238 (J)
giOllP
!
I 2
I 3
4
i
l 2
3
I 4
1
I 2
I 3
I 4
f
I 2
I 3
4
1
3;86
4,03
4,09
4,01
2,21
2,23
2,09
2,22
0,01
0,0244
0,0210
0,0360
0,0260
0,0256
0,0250
0,0236
2
3,51
3,63
3,62
3,59
1,72
1,79
.1,72
1,73
0,02
0,0266
0,0223
0,0215
0,012
0,011
0,0290
0,0292
3
3,27
3,33
3,32
3,32
3,32
1,86
1,87
1,85
0,03
0,011
0,0250
0,0232
0,024
0,021
0,021
0,024
4
3,12
3,14
3,14
3,14
1,97
1,93
1,95
1,93
0,04
0,022
O,D11
0,0294
0,060
0,049
0,055
0,059
5
3,01
3,01
8,03
3,02
1,76
1,78
1,74
1,75
0,04
0,052
0,022
0,025
0,13
0,11
0,13
0,11
6
2,95
2,93
2,96
2,95
1,59
1,65
1,61
1,61
0,10
0,112
0,065
0,091
0,13
0,12
0,13
0,11
7
2,91
2,89
2,92
2,91
1,53
1,51
1,55
1,51
0,16
0,154
0,16
0,18
0,14
0,13
0,13
0,12
8
2,89
2,88
2,94
2,89
1,50
1,50
1,53
1,53
0,23
0,20
0,22
0,22
0,18
0,16
0,18
0,16
9
2,88
2,87
2,90
2,88
1,47
1,61
1,58
1,60
0,26
0,30
0,25
0,35
0,26
0,26
0,28
0,26
10
2,87
2,87
2,89
2,87
1,59
1,61
1,69
1,62
0.,48
0,49
0,51
0;50
0,44
0,45
0,44.
0,41
11
2,87
2,86
2,89
2,87.
1,74
1,74
1,84
1,74
0,83
0,84
0,83
0,86.
0,62
0,61
0,,54 ?
0,55
12
2,87
2,86'
2,89
2,87
2,16
2,12
2,31
2,08
1,67
1,51
1,60
1,46
0,80
0,78
0,70
0,72
13
2,87
2,86
2,89
2,87
2,88
2,99
3,25
2,73
2,88
2,70.
2,93
2,49
1,14
0,98
1,09
0,95
14
2,87
2,86
2,89
2,87
4,01
4,04
4,44
4,26
3,72
.3,58
3,97
3,50
1,12
0,78
0,9f
0,98
15
2,87
2,86
2,89
2,87
'6,85
7,44
6,49
6,78
5,28
5,81
5,51
5,36
1,07
0,96
0,92
1,09
i6
2,87
2,86
2,89
2,87
10,7
10,4
10,G'
9,59
-8,52
8,82
8,10
8,00
1,22
1,14
0,95
1,12
17
2,87
2,86
2,89
2.,87
15,8
13,0
14,1
14,6
12,8
9,63
9,19
10,2
2,04
1,78
1,60
1,67
18
2,87
2,86
2,89
2,87
38,7
38,5
39,7
38,3
31,2
17,9
16,2
15,8
1,79
1,26
1,62
1,35
19
2,87
2,86
2,89
2,87
14,8
13,5
8,72
6,22
21,1
12.,9
15,6
20,4
3,32
3,26
3,22
3,90
20
2,87
2,86
2,89
2,87
70,2
53,4
41,4
26,8
42,8
29,9
28,2
16,1
4,50
3,39
5,73
20,&
*1) BNAB-70; 2) BNAB-M;
3) KFK-INR; 4) ENDF/B 1V,
TABLE 6, Comparison of Calculation Results
Physical
parameter
197p ~ .1976 ~ 1970 ~ 1976.
M izsepp), kF
B
Ba.z
G
~slfs
as
f8/f9
M ~z3epu),.kg
B
Ba.z
G
~slfs
a ulfa
falfa
975
1,34
0,74
0,37
0,161
0,301
0,0247
1,82
1006
1,36
0,74
0,44
0,162
0,270
0,414
0,0268
1,75
968
1,30
0,72
0,35
0,155
0,306
0,0246
1,89
963'
1,31
0,71
0,37
0,156
'0, 282
0,298
0,0262
1,82
952
1,25
0,67
0,30
0,147
0,334
0,0224
1,89
956
1,26
0,67
0,39
0,149
0,310
0,316
0,0236
1,83
Winfrith
1970 ( 1976
946
1,31
0,70
0,38
0,151
0,319
0,0239
1,84
939
1,26
0,69
0,34
0,146
0,306
0,0221
1,92
948
1,31
0,68
0,38
0,153
0,299
0,286
0,0256
1,79
929
1, 25
0,66
0,31,
0,148
0,281.
0,198
0,0236
1,83
929
1,32
0,71
0,37
0,147
0,302
0,0255
1,87
990
1,33
0,71
0,39
0,149
0,282
0,317
0,0267
1,82
1970 1976
961
1,31
0,74
0,38
0,160
0,312
0,0226
1,87
965
1,28
0,70
0,36
0,162
0,290
0,240
0,0242
1,81
945
1,32
0,73
0,39
0,153
0,298
0,0236
1,90
933
1,31
0,70
0,37
0,156
0,279.
0,222
0,0246
1,85
1970 ~ 19'76
978
1,41
0,79
0,46
0,164
0,253
0,0239
4,89
973
1,28
0,71
0,30
0,152
0,304
0,0239
1,86
951
i ,42
0,75
0,46
0,162
0,237
0,238
0,0264
1,84
and therefore we may expect the discrepancy between the average cross sections. to be mainly associated with
the dffference in the group constants, Noting the data of Table 4, we can observe the following:
the active-zone average capture cross sections for 238U in the systems fitted under integral data (OSKAR-
75, CARNAVAL-1V, FD-5) is ~ 6% lower than in the BNAB-M', KFK-INR, and. E NDF/B IV systems, which are
based primarily on microscopic data;
a -for 2~Pu in the OSKAR-75 calculations is ~~ higher than the data;
the dispersion of the average capture cross sections for Fe, Ni, Cr, and Na is ~ 50%;
the average cross section for the fission products in the CARNAVAL-IV calculations differ considerably
from the other data, It may be noted that the values shown in Table 4 (0.496 for variant B and 0,489 for variant
C) also differ considerably from the analogous values and the calculation based on the previous version,
CARNAVAL-III [22] (0.522 and 0,519), These changes are mainly due to the change in the fission-product con-
stants, since the neutron spectrum did not change much in the transition from the third to the fourth version,
In Table 6 we compare the results of the calculation of the main parameters for the standard reactor
(variants A and C) shown in this article with the data obtained fn 1970 [1], We can see that the critical charge
in variant A has changed little; there is a substantial reduction in the average dispersion fn the reproduction
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parameters [ (B=1,32 t 0.07 (1970); B =1,31 f 0.02 (1976), variant A], From this, of course, it does not follow
that the real accuracy of the calculation of the physical reproduction factor is this high. Because of the un-
questionable correlation between the principal nuclear data used in different laboratories, the estimate of the
accuracy (f 0,02) should rather be considered a lower bound, it should also be noted that the considerable
dispersion in critical charge (973 kg for the ENDF/B 1V, 929 kg for the FD-5) in itself signifies an indefinite-
ness in the physical reproduction factor (t 0,05); the average dispersion of the data on the critical charge in
variant C remains large (M =961 t20, 1970; M =959 f 18, 1976); the change in the ratios of the active-zone
averages of the reaction rates is the following: c8f f9 = 0.156 * 0,07, ~y = 0,301 f 0.020 (variant A, 1970); c$/ fy =
0,152f0,02, ay=0,306f0,010 (variant A, 1976),
The observed closeness in the results of the calculations of different laboratories objectively indicates
an increased reliability in the calculation of fast reactors in different countries. There remains a consider-
able degree of indefiniteness in the prediction of the main physical characteristic, namely, the critical charge,
it is desirable ~to make a representative international comparison of the results of the calculation of a two-
dimensional model, which is closer to a real reactor than the one-dimensional model considered here,
The authors are deeply grateful to the staff of the laboratories in the various countries who were good
enough to send the results of their calculations, and also to M. F. Trayanov for his support and his comments
at every stage of the work.
LITERATURE CITED
1, A, Baker and A, Hammond, Calculations for a Large Fast Reactor. Risley, TRG Report 2133 (A) (7971),
2, L, P. Abagyan et al? Group Constants for the Calculation of Atomic Reactors [in Russian], Atomizdat,
Moscow (1964).
3, A, A, Van'kov,A. I, Voropaev and L, N, Yurova, Analysis of aReactor-Physics Experiment [in Russian],
Atomizdat, Moscow (1977),
4, A, I, Voropaev et al? in: Problems of Atomic Science and Technology, Nuclear Constants Series [in
Russian], Vol, 2, No, 20, Atomizdat, Moscow (1975), p. 112
5, A, I, Voropaev et al? in: Problems of Atomic Science and Technology, Nuclear Constants Series, No,
25, Izd, TsNilatominform, Moscow (1977), p. 69,
6, L. P, Abagyan et al? "Calculations of the characteristics of a "standard" fast reactor," Preprint FEi-
525 Obninsk (1974),
7, Yu, G, Bobkov and L, N, Usachev, "Results of the calculation of a standard reactor by the OSKAR-75
system," Preprint F~`I-659 (1976),
8, B, G. Bobkov et al? in: Proceedings of the "Neutron Physics" Conference, Part I [in Russian], izd,
Ts DTIlatominform, Moscow (1976), p. 64,
9, V, F, Khokhlov, M. M, Savos'kin, and M, N. Nkolaev, in: Nuclear Constants, No, 8, Part 3 [in Russian],
Atomizdat, Moscow (1972), p, 3.
10, J. Chaudat and J. Courchinoux, Characteristiques d'un It~acteur Rapide "Etalon" Calcul8 avec le Formu-
laire CARNAVAL-IV, Note Technique, Cadarache (]977),
11, I, Barre, I, Bouchard, and I, Chadar, in: Proceedings of the Fourth Conference on Nuclear Cross Sec-
tions and Technology, Washington, March 3-7,1975, Vol, 1, p. 51,
12, W. Barron and J. Mann, A Comparison of FD-5 and FD-4 Data for a Spherical Reactor 1VIode1, Technical
Note, Risley (1975); J, Rowlands, Calculations for the Standard Fast Reactor, Technical Note, Winfrith
(1975).
13. J. Rowlands et al? in: Proceedings of the IAEA Symposium on Physics of Fast Reactors, Tokyo, 1973,
Vol, 3, p. 1133,
14, E, Kiefhaber and D, Thiem, Institut ~iir Neutronenphysik and Reaktortechnik. Technical Note, Karlsruhe
(1976),
15, E, Kiefhaber, KFK-1572, Karlsruhe (1972),
16, H, Kusters, KFK-1632, Karlsruhe (1973).
17, W, Oesterkamp, Trans, Am, Nucl, Soc? 16, 262 (1973),
18, R, Bucher et al? Calculation of the Breeding Properties of a Standard Fast Reactor, Technical Note,
ANL (1975).
19, R, McKnight, Nucl, Sci, Eng? 62, No, 2, 309 (1977),
20, P, Hammer and F, Plum, Physics Investigations of Sodium-Cooled Fast Reactors Core Z1 Masurca in
SNEAK Assembly, 6D, CEA-N-].561, KFK-1581 (1972),
21, P, Collins and M, Lineberry, Trans, Am, Nucl, Soc? 24, 481 (1976),
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 Declassified and Approved For Release 2013/03/22 :CIA-RDP10-021968000700120001-4
22, d, chaudat, J, Courchinoux, and J, Barre, Caracteristiques d'un Reacteur Rapide Etalon Calcule avec
le Formulaire CAR NAVAL-III, Note Technique, Cadarache (1975).
POWER-RE-ACTOR FUEL-PRODUCTION PROBLEMS*
F, G. Reshetnikov, Yu. K, Bibilashvili, UDC 621,039,54:621,311,2:621.039
and V, I. Kushakovskii
The quality of fuel-pin cores largely determines the viability and reliability of any nuclear-power. sys-
tem, Particular attention is therefore needed in the technology of pin production in order to ensure high and
consistent quality, Although there are now closely defined specifications for UOZ, and these ensure high via-
bility, improvements to the specifications should not only provide higher reliability but also should indicate
where specifications are too stringent on particular parameters, which might assist in cheapening the process.
Fluorine and Water
it has previously been pointed out [1] that one of the major specifications for a uranium oxide fuel (the
acceptable fluorine content) has been formulated without due allowance for the relationship to the water con-
tent in the pin, The acceptable fluorine content limit has been set as 0,006%. Dry fluorine at that level can-
not produce any substantial pitting corrosion in the sheath, whereas water will convert such fluorine to a
highly corrosive form, Therefore, a particular need is to reduce the water content, the more so since water
is a basic source of hydrogen, The manufacture of UOZ tablets is now highly advanced, and the water levels
do not usually exceed 0,003-0,0005%, On.the other hand, the specifications for fluorine have been tightened,
which has given the impression that the specifications do not have a proper experimental basis, There is
therefore some substantial practical interest in demonstrating the relationship between the standards for
acceptable fluorine and water levels in fuel pins,
Specifications for Mechanical Parameters
Some specifications for UOZ concern the mechanical features; the failure force is an inadequate char-
acteristic for any tablet, and other criteria must be used to evaluate the fracture resistance during the vari-
ous manufacturing and transport operations, For example, appropriate vibrational tests must be performed,
Then it would be possible, e,g? to use'a relatively small number of tablets and detect ones with latent flaws
such as microcracks, which would mean that not so many expensive assembled pins might be rejected, These
cracks are particularly hazardous when the pins have been assembled, because the damage in the cores may
increase on irradiation, which can result in increased forces on the sheath and ultimately in premature failure
of the pin, Therefore, preliminary careful testing of the material may be economically justified.
The absence of cracks does not necessarily mean that the sheath will not be affected, e,g? by thermal
expansion or~swelling in the fuel, Although the effects may be of very different types, the loads on the sheath
may become very considerable. Of course, the loading can be adjusted to a certain extent via the density of
the tablets, the volume of the end recesses,. and the diameter of the holes, There may also be other effects
on the interaction from the ratio of height to diameter, which should be within the range 1,1-1,2.
However, any adjustment to the above parameters is restricted by the specifications for a fairly high
mean or effective density of the fuel in a pin, It has therefore been suggested that fuel should be made reason-
ably plastic at the irradiation temperature in order to eliminate substantial forces on the sheath,
The plasticity of sintered UOZ can be increased by special technologies and by the addition of other
substances, e,g? other oxides that form low-melting eutectics, The choice of such substances is not an easy
matter, since the amounts must be small; also, they must not form solid solutions with UOZ, while the low-
temperature eutectics must lie along the grain boundaries, Finally, these additives must not have any marked
effects on the physical characteristics of the reactor itself, Nevertheless, the suggestion is valuable, and it
should be given appropriate attention, ,
*Journa.l form of a paper read at the Reactor Materials Science Conference, Alushta, 1978.
Translated from Atomnaya nergiya, Vol, 45, No, 6, pp, 426-429, December, 1978,
1162 0038-531X/78/4506-116? $07.50 ?1979 .Plenum Publishing Corporation
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 Declassified and Approved For Release 2013/03/22 :CIA-RDP10-021968000700120001-4 .
The production of cores with improved plasticity is particularly important for any power reactor that is
to work with highly variable or peaky loads,
Fuel-Pin Design .for Power Stations Working
Under Variable-Load Conditions
The power-control problem becomes more important as the output power of a nuclear station increases,
as there is more difficulty in matching the load curve. The specifications for control characteristics of nuclear
power station units are presently formulated mainly from user requirements, The control specifications lay
down that the power levels of the units should be adjustable at rates of up to 4-6% per minute. These specifica-
tions envisage power stations with thermal reactors, because the economic performance of a fast reactor is
adversely affected by operation with peaky loads, and this relates particularly to the entire fuel cycle, not
feast because of the direct effect on the rate of accumulation of plutonium,
The development. plans for nuclear power in the USSR and other countries envisage preferential construc-
tion of reactors using rod systems with cores composed of compact UOZ sheathed in zirconium alloy. The
loads on such sheaths under steady conditions are closely known, The forces arise from the external pressure
of the coolant, the internal pressure from the gaseous fission products and filling gases, the swelling in the
fuel proper, and thermal expansion, The physicomechanical characteristics of the material are affected by
the radiation damage occurring during operation, in addition to any effects from mechanical fatigue, inter-
actfon with the coolant and the fission products, and so on,
Ail of these factors reduce the reliability of the fuel rods considerably, Airy new effects superimposed
on the long-term processes, such as ones arising from sharp changes in the temperature and heat production,
naturally reduce the fuel-rod viability.
It is now considered firmiy established that the main mechanism responsible for fuel-rod failure on ac-
count of power-level variation is the mechanical force exerted on the sheath by the fuel consequent on the
thermal expansion, This interaction occurs in part because the temperature of the core is different from that
of the sheath, while the thermal-expansion coefficient of zirconium alloy is less by a factor 1,5-2 than that of
UO2, The danger of sheath failure under thermal cycling is made worse by the tendency of the zirconium
sheaths to show corrosion cracking under stress in the presence of corrosive fission products, in particular
iodine,
Reliable operation under variable conditions has been researched as follows.
1. The limiting permissible thermal loads have been determined for rates of change of power level up
to 4-6% per minute, Preliminary estimates show that the linear thermal load in that case should not exceed
400 W/cm.
2, The optimum rates or runup have been determined for systems with rods operating at 550-600 W/
em. In this connection, it has been necessary to consider the scope for increasing the threshold failure power
when the reactor is run up to nominal power by stages. This suggestion involves maintaining the rods at two
or three intermediate power levels for times sufficient to allow 40-60% relaxation in the stresses in the sheaths
consequent on relaxation in the fuel,
3. Cores are being developed showing high creep rates or low temperatures of brittle-plastic transition,
which should reduce the relaxation times and also the stress levels in the sheath. This implies the addition
of minor components to the UOz, which have almost no effect on the melting point but which do increase the
creep rate of UOz in the range 700-1000?C. There are also other possibilities for increasing the plasticity
of the cores.
4. Detailed studies are being made on the corrosion cracking of zirconium sheaths; in particular, one
needs to establish the main factor responsible for such cracking, namely whether the limiting tensile stress
or the threshold plastic-strain rate is the more important factor, In the first case, alloys with low yield points
would be better, whereas stronger alloys would be preferable in the second. However, this is not the only pos-
sible criterion, because other factors influence the corrosion cracking: The composition and structural state
of the alloy, the working temperature, the general quality of the sheath, etc.
Mixed Oxide Fuel
The above arguments apply equally to mixed uranium-plutonium oxide fuel, but here there is also an-
other very important specification, viz., that the uranium and plutonium should be uniformly distributed. There
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are two essentially different methods of producing mixed oxide fuel: coprecipitation and mechanical mixing,
The first is used with high-temperature sintering, which gives an ideal solid solution of the two oxides,
In the second method, the uniformity of the distribution is less perfect and is dependent onthe grain
size of the initial powders, the care taken in the mixing, and to some extent on the mode of sintering, since
high temperatures and fairly long times are required to homogenize mechanically mixed oxides.
However, the advantages of the coprecipitation method involve considerable expense, since the repro-
cessing of large amounts of fuel containing plutonium is made more complicated. Also, the coprecipitated
oxides cannot be stored for long periods. The mixture must be used almost immediately, since the accumula-
tion of zaiAm hinders the subsequent handling and may require fresh purification.
in practice, it is often difficult to organize the production cycle for immediate use of prepared materials,
particularly if the cycle involves a plant concerned with reprocessing uranium-plutonium fuel from spent rods,
i,e,, the plutonium may be repeatedly purified from zaiAm, In that case, the cost of repeated fuel purification
will be much less if the i?uOz is not mixed with UOz.
All of these difficuities are much reduced if one uses mechanically mixed UOz and PuOz, with prepara-
tion directly before manufacture, Research has shown that the specifications previously laid down for the
homogeneity of mixed fuel are too stringent. Comparative testa have shown that identical degrees of viability
are obtained by coprecipitation and mechanical mixing [2, 3J, There is a certain amount of redistribution of
the plutonium that has no appreciable effect on the physics of the reactor during the irradiation, no matter how
the fuel pins are produced, Therefore, mechanical mixing is considered as the technique of preference for
ma.ktng uranium-plutonium fuel,
Granulated Fuel
So far we have discussed fuel pins in which the cores are composed- of seta of tablets of average density
10,5=10.6 g/cros, i,e? about 96'1c of the theoretical value. The effective fuel density in the pins for a thermal
reactor is then about 9.55 g/cros when allowance is made for all the permissible gaps, viz? about 87?!0 of the
theoretical value for the VV~R-1000 reactors, or correspondingly 10 gems or 91q of the theoretical value
for the RBMK reactors,
The effective fuel density is somewhat lower at 8,8 g/cros, or about 84~ of the theoretical value, for
fast reactors; such densities are readily attained by the usual techniques for loading tablets with central holes
or recesses, On the other hand, there are also other loading techniques based on vibrational consolidation of
granulated fuel, and some experience is already available with such rods,
Granulated fuel has certain advantages over tablets; the sol-gelmethod of producing such fuel eliminates
the dust hazards of powder techniques, as well as, tablet grinding, and it also reduces the mechanical inter-
action between the fuel and the sheath and improves the heat transfer,
However, the published evidence indicates that vibrational consolidation of granulated fuel has not yet
been widely used, although the technique is attractive in certain respects. One reason for this appears to be
that good results have been obtained in making cores from sets of sintered tablets. Also, there are the follow-
ing major aspects to be examined in any proper evaluation of the desirability of transferring to vibrational
consolidation of granulated fuel.
1. An efficient and economical technique has to be developed for producing the granulated fuel, in par-
ticular densities close to the theoretical value, Such a technique should provide granules of closely controlled
size, and it is necessary to have a set of sizes, including a fine fraction of grain size about 20 ?. Unless these
conditions are met, the necessary effective fuel density cannot be attained, particularly for thermal-reactor
rods, However, it appears difficult to attain even 90~ of the theoretical density even if optimal characteristics
are attained in the granulated fuel.
2. It is necessary to define a mode of vibrational consolidation such as to ensure a reasonably uniform
distribution of the fuel throughout the length of a rod, The length of the active zone of a rod in a BN-350 re-
actor is about 1100 mm, while it is about 3500 mm for thermal reactors (RBMK and VVER-1000), The devi-
ations from the specified density should not exceed 3-5%. It will not be an easy matter to attain this over
such large lengths.
3. Research is required on the response of sheaths, particularly of zirconium alloys, during vibrational
consolidation of such long rods, since dense sintered UOz is very abrasive,
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Some researchers consider that most of these problems can be solved fairly readily, and they assume
that rods containing vibrationally consolidated fuel, particularly of mixed uranium-plutonium type, may re-
place the tablet type [4], However, major researches must be undertaken before any serious comparative
evaluatfion'of;the techniques can be performed,
1, F, G, Reshetnfkov et al? At, Energ? 43, No, 5, 408 (1977),
2, G, Karsten, Trans, Am. Nucl, Soc., 19; No, 1, 19 (1974),
3. R. Hilbertetol, Am. Nucl, Soc., 19, No. 1, 135 (1974) .
4. W. Helfrid, Lahr, Nucl. Tech., 31, No. 2, 7.83 (1976).
ZIRCONIUM ALLOYS IN NUCLEAR POWER*
A, S, Z a i m o v s k i i UDC 546,831 +621,039.5
Presently there is a tendency for the unit power to increase in water-cooled reactors; this has meant
more stringent specifications for the corrosion resistance and mechanical parameters of zirconium sheaths
and tubes, Further, there has been an increase in the interest in the behavior of zirconium alloys under
emergency conditions (COCA), Finally, there is a tendency to operate reactors with more peaky loading.
The development program for the RBMK reactors in particular envisages the need for alloys of high corro-
sion resistance capable of working in superheated steam at 400-500?C.
For this reason, the Soviet Union has researched zirconium alloys since 1950, in particular the phase
diagrams and structures of about 50 ternary systems. Transmission electron microscopy and x-ray analysis
have been employed to define the solubility of niobium in a-2r, which is about 0.7 at, %, and studies have been
made on the formation of metastable phases and the sequence of transitions in Zr-.Nb alloys quenched from
the ~-phase region [1],
Niobium as an Alloying Element for Zirconium Alloys, Niobium is used in alloying zirconium for sev-
eral reasons,
The oxide film on zirconium is usually considered as a semiconductor with anion vacancies, which in-
dicates that cations of valency 5 or 6 will enter lattice nodes and reduce the number of anion vacancies, which
,will reduce the scope for oxygen ions to penetrate and thus tend to inhibit corrosion, Niobium is far better
than tin as a hardening agent, and is inferior only to aluminum and molybdenum, which adversely affect the
corrosion resistance. Niobium also has a relatively small thermal-neutron capture cross section,
Niobium also offsets the adverse effects of N, C, Al, Ti and other impurities (Fig, 1), and also substan-
tially reduces the absorption of hydrogen; it is highly soluble in ~-Zr and has reasonable solubility in a-Zr,
which makes it favorable to mechanical treatment and means that the parameters of the alloys can be varied
widely by heat treatment, The high melting point facilitates casting zirconium alloys, the more so since there
is only one alloying element, in contrast to the four elements used in zircalloy, for example,
Properties of N-1 Alloy. Classical data on the corrosion of N-1 and N-2.5 alloys outside reactors [2,
3j indicate that alloys containing 1 or 2,5% Nb differ from zircalloy in showing no stepout in the oxidation
kinetics on testing in water and steam-water mixtures for times up to 20,000 h, The oxide films adhere
firmly to these alloys, while microcracks tend to heal up,
Numerous studies have been made on the effects of irradiation on the corrosion resistance of N-1, and
it is clear that irradiation of itself has little tendency to accentuate corrosion, provided that there. is no oxy-
gen in the water or steam-water mixture, or a~ other oxidizing agent arising from radiolysis of water, as
Fig, 2 shows [4, 5j.
It has been concluded [6] from tests lasting up to 20,000 h under boiling conditions free from deposition
in the RBMK reactors that N-7. and N-2.5 are corrosion-resistant and retain their plasticity; e,g., the total
residual strain in N-7 was not less than 10%, Local signs of nodular corrosion were seen on tubes of this
alloy, which were similar to those seen in zircalloy,
Journal form of a paper read at the Reactor Materials Science Conference, Alushta, 1978.
Translated fTOm Atomnaya ~nergiya, Vol, 45, No, 6, pp, 430-433, December, 1978,
0038-531X/78/4506- ?1979 Plenum4 Publishing Corporation
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 Declassified and Approved For Release 2013/03/22 :CIA-RDP10-021968000700120001-4
At Ti C SL fe N~
~~ ~.
iw
1105 1D0 Hh ~ ~ w
N
?~ 6175 61~ 67 9U '15 1a 75 19 ~ ~ 7d ~ ~ p 1 ~()
~ ea--f~h
??~? o00 ?~e?m =~m? o
soooo ooo~ s~~., m_.,~o
~~ dddo' dc3o ~ dcf'ddd' dddd 'ddd- C'dood
Mass. %.
Fig, 1
3 50
1,5%H6
Q5%Cu
1%Sn=
Q5%Fe
o,1~ni~
Q1%Fe
1% Nb-
q5%Fe=
1%Sn
Fig, 1, Effects, of trace components on the corrosion resistance of N-1 alloy in water at 350?C
for 7000 h,
Fig, 2, Corrosion resistance of zirconium alloys in water and steam-water mixtures in anauto-
clave and in a reactor loop (280?C, 3500 h): tests outside reactors: 1) autoclave, water under
pressure,, 0,3-0,6 mg/kg of 02; reactor tests: 2) boiling water containing 0,5-0,6 mg/kg of 02,
outside core; 3) the same, but with 12-17 mg/kg of 02; 4) boiling water with 12-17 mg/kg of 02,
irradiated by a fluence of 3,7.1020 neutrons/cm2 (E > 1 MeV),
t 2 3
penetration depth, ?
Fig, 3. Distributions of 180 in oxide films on: 1) zirconium,
which indicates diffusion over short paths; and 2) on N-1,
which is characteric of diffusion by a bulk-vacancy mech-
anism,
N-2,5 Alloy and Improvement of Zirconium-Niobium Alloys, .This alloy is used widely throughout the
world because of its good mechanical parameters and good corrosion resistance in reactors, Canada has
been particularly prominent in research on this alloy,
Figure 2 shows the weight changes for six alloys under various conditions, including ones that cause
the highest corrosion rates; the gain in weight of an alloy of zircalloy type (1% Sn-0,5% Fe) under irradiation
is three times that found in tests outside reactors, while an experimental alloy (0,7% Fe-0,7% Ni) showed
catastrophic .corrosion on account of rapid uptake of hydrogen consequent on the high nickel content, The
best results were obtained with TsZhNo alloy (1% Nb~-0,5% Fe-1% Sn), whose weight gain was minimal in
all cases, whether in an autoclave outside a reactor or in boiling water emerging from the core and contain-
ing 12-].7 mg/kg of oxygen, as well as on exposure to a fluence of 3.7.1020 neutrons/cm2 in boiling water with
the same oxygen content,
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 Declassified and Approved For Release 2013/03/22 :CIA-RDP10-021968000700120001-4
Fig. 4. Corrosion (a) and hydrogen absorp-
tion (b) during tests on zirconium alloys in
a steam-water mixture at 350?C or in super-
heated steam at 400-500?C for 3000 h forthe
following alloys; ^) TsZhNO; ^) Zr.- 1% Sn -
0,5% Fe; ?) N-2,5; 0) N-1; ?) Zr-(Fe, Cr,
W, Mo).
X300 350 400 450 500 550
Temperature. 'C
Oxygen Diffusion and Phase Composition for Oxide Films on Zirconium Alloys, Researches have been
performed in the Soviet Union in recent years on the corrosion of zirconium alloys in order to provide alloys
with good corrosion resistance in superheated steam.
Mossbauer techniques were used to determine the phase compositions of oxide films of alloys co~ain-
ing iron and tin when these elements enter the films during corrosion; this defines a relationship between
the phase composition of the film and details of the corrosion in media containing oxygen [7], The main re-
sult from the Mossbauer studies was that there are various phases in an oxide film, viz,: a-Fe, ~-Sn, FeO,
Fe203, Fe304, SnO, Sn02, as well as solid solutions of various ions in Zr02 and ZrH2, while the proportions
of these phases vary during the corrosion. Therefore, the oxide film is substantially heterogeneous, The
plastic metal phases (a-Fe, ~- Sn) probably are advantageous to the protection provided by the film, whereas
the oxide and hydride phases make the film brittle. The volume ratio between the metallic and nonmetallic
phases is clearly one of the major factors governing the corrosion resistance.
The mode of oxygen transport in the oxide films was examined by treating the oxidized alloys with H2O18
followed by evaluation of the oxygen concentration gradient by microanalysis [8]. The oxygen distribution
(Fig, 3) shows that unalloyediodide zirconium produces one type of diffusion mechanism, whereas N-1 alloy
under the same conditions (500?C and about 100 kPa in steam) has another: in the first, the diffusion is along
short paths (grain boundaries, cracks, etc.), whereas the diffusion occurs throughout the lattice (via vacancies)
for N-1. However, the overall results from the two processes are fairly similar and the corrosion rates are
roughly the same. Also, marked differences in corrosion resistance under these conditions were observed
for most of the alloys of zirconium containing 1 at. % Al, Cu, Fe, Mo, or W, but the oxygen diffusion mech-
anism in the films was always the same: preferentially along short paths, This is a difference from alloys
containing niobium and tin, where the vacancy mechanism predominates.
Therefore, the oxygen transport by the bulk-vacancy mechanism characteristic of zirconium alloys con-
taining tin and niobium indicates that these alloys have comparatively little corrosion resistance in steam at
400-500?C; on the other hand, alloys containing iron, copper, tungsten and probably also chromium have good
corrosion resistance in superheated steam.
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 Declassified and Approved For Release 2013/03/22 :CIA-RDP10-021968000700120001-4
Alloys for Reactors with Nuclear Steam Superheating. The Soviet Union has been a pioneer in reactors
with nuclear steam superheating [9]. Replacement of steel tubes by zirconium ones not only increases the
efficiency of the system considerably but also provides a reduction in the temperature of the superheated.
steam, A zirconium tube must be of bimetallic type: The side facing the superheated steam must be made
of acorrosion-resistant alloy containing no tin or niobium, while the other side must be made of a strong and
heat-resistant alloy containing tin, niobium, and other strengthening elements,
Figure 4 shows co-rrosion-resistance and hydrogenation curves for 350-500?C for-the above alloys con-
taining tin and niobium on the one hand and for alloys containing Cr, Fe, Cu, W, and Mo in various combina-
tions on the. other. The oxidation and hydrogen uptake. are most pronounced for alloys containing tin, TsZhNO,
and Zr-1,5% Sn-Q,5% Fe, Somewhat lower weight gains due to corrosion occur for N-1 and N-2,5, The
exceptionally large weight gain for TsZhNO is due to the simultaneous presence of tin and niobium, The
largest hydrogen.uptake occurs in Zr-1,5% Sn-0,5% Fe, whose composition is similar to that of zircalioy,
and this is due to the lack of niobium. The hydrogenation of TsZhNO alloy, -which contains 1%_ Nb, was less
by a factor oft.5,whileevenhigherhydrogenationresistance.occurredinthe pure niobium alloys. N-1 and N-2;5,
Alloys with several alioying elements show corrosion.factors smaller by 5-10" times, and these are
clearly promising for fuel-rod sheaths and tubes for reactors with nuclear steam superheating:
1. O, S, Ivanov (editor), Physical Chemistry of Zirconium Alloys [in-Russian], Nauka; Moscow (1968);
Structure and Properties of Alloys for Nuclear Power [in Russian], Nauka, Moscow (1973),_
2, R. S, Ambartsumyan et al,, "Mechanical parameters and corrosion resistance of zirconium alloys in.
water; steam, and gases at elevated temperatures," Proceedings of the Second Geneva Conference, 1958,.
USSR" Paper No, 2044,
3, A, A, Kiselev et al? "A study of the corrosion of zirconium alloys in water and steam at high tempera-
tures and pressures," Paper at the IAEA Conference on Corrosion of Reactor Materials, Salzburg
(1962),
4, A., D, Amaev et al., "Principles for the choice of zirconium all"oys for fuel-rod sheaths in routine-
production VV~R-440 power reactors," Proceedings of-the Conference on Nuclear Power, Fuel Cycles,
and Radiation Materials Science, Ulyanovsk, 1970 [in Russian], Izd, S~V, Moscow, No, 3(1971), p,.503.
5, A. D. Amaev et al? "Corrosion of zirconium alloys in boiling water under irradiation," Proceedings. of
the Fourth Geneva Conference, 1971, USSR Paper: No, 428.
6, A, D. Amaev, E, G, Ivanov, and G, P. Saenko, "Mechanical and corrosion parameters of zirconium.alloys
for fuel-rod sheaths and tubes in RBMK reactors," Paper at the British-Soviet. Seminar on Reactor.
Channels and Fuel Rods [in Russian], Moscow (1977),
7. Yu, F, Babikov, A: A. Khaikovskii, et a1? "MiSssbauer studies of the effects of oxidation. on the distribu-
tion of Fe-57 and tin-119 in zirconium alloys," in: Metallurgy and Metallography of Pure Metals [in
Russian], No,' 12, Atomizdat, Moscow (1976), p, 16.
8. V. N, Abramtsev and A. A. Khaikovskii, Diffusion of Oxygen in Oxide Films Formed on Zirconium and=
Zirconium Alloy Containing 1% Niobium [in Russian], Preprint VN[INM, Moscow (1973).
9. N. A. 1)ollezhal' et al? "Design of uranium-graphite channel reactors with tubular fuel rods and nuclear
steam superheating," Proceedings of the Conference on Nuclear Power, Fuel Cycles, and. Radiation
Materials Science, Ulyanovsk, 1970 [in Russian], Izd. S~V, Moscow, No. 1 (1971), p. 368.
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 Declassified and Approved For Release 2013/03/22 :CIA-RDP10-021968000700120001-4
SWELLING OF STEELS AND ALLOYS IRRADIATED
IN THE BOR-60 REACTOR TO A FLUENCE
OF 1,1 1023 NEUTRONS/CM2*
N,
P.
Agapova, V. S, Ageev,
I,
N.
Afrikanov, N. T, Budylkin,
V.
A.
Krasnoselov,: E, G, Mironova,
V.
D,
Onufriev, Z, E, Ostrovskii,
V,
I,
Prokhorov, and Yu. N. Sokurskii
A serious difficulty in the construction of fast power reactors is the radiation-induced swelling of the
materials' of the fuel-element jackets and the fuel-assembly (FA) sheaths, At a fluence of (2-3) ? 1023 neutrons/
em2, the expected swelling of the stainless steels and alloys now in use considerably exceeds 10%, which is
clearly unacceptable for the fast reactors now being designed and makes it necessary to find ways of reduc-
i~? the swelling of jacket and sheath materials to reasonable limits,
Because the chemical composition and the prior thermomechanical working is a major factor in the
radiation-induced swelling of stainless steels and alloys, this swelling must be evaluated for each variant
of chemical composition and working.
The investigation was conducted on specimens of OKh16N15M3B austenitic stainless steel, after various
types of working, on the nickel alloys OKh17N40B, OKh20N60B, and KhN77TYu, on specimens of some alloys
of Kh16N15 steel, and also on specimens of nickel and iron with purity values of 99,99% and 99,98%, respec-
tively (Table 1),
In order to determine the swelling by hydrostatic weighing, we used cylinders 6 mm in diameter and
30 mm in height, Specimens of melts of Kh16N15 steel and OKh77N40B alloy were subjected to heat treat-
ment in a vacuum at 1050?C for 0,5 h, specimens of KhN77TYu alloy at 1050?C for 0,5 h and at 720?C for 2 h,.
and specimens of iron and nickel at 850?C for 0.5 h. Some specimens were deformed by tension by values of
5, 10, and 20% (0Kh16N15M3B steel) or 70% (iron, nickel). A number of specimens of OKh16N15M3B steel,
after austenizing annealing, were subjected to various types of working: a) aging at 650?C for 100 h; b) me-
chanical-thermal working (MTW) and cold deformation (CD) by 10%, with subsequent annealing at 800?C for
3 h; c) thermomechanical working (TMW), which was deformed by 10% at 650?C with subsequent annealing
at 650?C for 100 h; d) cold deformation by 80% with subsequent annealing at 800?C for 1 h in order to produce
a fine grain (3-5 ?). At least three specimens of each material were irradiated,
The specimens for investigation by the method of electron microscopy were cut from the appropriate
cylinders and were reduced to a thickness of 0.1-0.2 mm by mechanical grinding and electrolytic polishing,
From these blanks we cut out disks with a diameter of 3 mm, which were sealed into special vacuum cassettes
before irradiation.
The irradiation was carried out in materials-study ampuls in the BOR-60, The specimens for hydro-
static weighing were irradiated to a fluence of 1,1 ? 1023 neutrons/em2 at a calculated temperature of 450 to
550?C, and the cassettes with the specimens for electron microscopy were irradiated to a fluence of 7,5' 1022
neutrons/cm2 (E > 0) at a calculated temperature of ^? 520?C.
After irradiation, the specimens for electron microscopy were electrolytically thinned and investigated
on the EM-300 electron microscope, The hydrostatic weighing was carried out on aremote-controlled balance
in CC14 with an accuracy of 0.2% or better.
* Journalversionof areport delivered at the Conference on the Study of Reactor Materials (Alushta, 1978).
Translated from Atomnaya Energiya, Vol. 45, No, 6, pp. 433-439, December, 1978,
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TABLE 1. Chemical Composition of Steels and Alloys, %by Weight
?`:atPiial
I C 1
Mn
I Si I
Cr
I Ni I
Nb
I Mo
~ Ti
I Al I
Fe
OKh16N15M3B
0.042
0,72
0,40
16,05
15,20
0,45
3,00
-
-
Remainder
OKh16N15
0,052
0,45
0,25
14,90
15,96
-
-
-
-
Same
OKh16N15M3
0,055
0,43
0,25
15,34
15,56
-
2,99
-
-
?
OOKh16N15M3
0,005
0,52
0,32
15,75
15,34
-
2,91
-
-
?
OKh17N406
0.030
1,59
0,05
16,70
41,05
0,64
-
-
-
3g,g
OKh20N60B
0,020
1,85
0,11
20,30
57,94
0,57
-
-
19,1
KhN77TYu
0,050
0,25
0,3?
20,83
74,60
-
_ ~
2,73
0,78
0,39 .
Irradiation parameters
Change in
Material
Treatment
E, neutroiis/crri2
F,1022 neutrons/
Tcalc'
density,
(E> 0)
cm2(E> 0,1 McV)
I C
%
OKh16N151`136
Austenization (A) at 1050 `C,
1,1.1028
9,0
550
13
7
steel
30 min
,
A-I-iO?~ CD
1,1.1023
9,0
550
0,4-7,0
A~-20?o CD
1,1.1029
9,0
550
0
MTW (A 110?ro CD -800 ?C, 3 h)
1,12.1025
9,2
550
7,4-13,5
A-~-650 ?C, 10U h
1,92.1023
9,2
550
12,3
TMW (A190%CD at 650 ?C~-?
1,12.1028
9,2
550
3,8
Nickel
-{-650?C, 100h)
?Anneal. 850 ?C, 1h
1,1.1023
9,7.1022
9,2
8,5
550
450
3,3
4,0
Anneal. 850 ?C, 1h -I-10% CD
8,7.1022
7,5
450
3,8
Iron
nneal, 850 ?C, 1h
8,7.1022
7,5
460
1,5
nneal. 850?C, 1h 110% CD
8,7.1022
7,5
460
9,0
TABLE 2. Swelling of Materials According to the Hydrostatic Weighing Data
TABLE 3? Parameters of the Radiation
Porosity in Irradiated Materials
Material
Treatment
ev/v,
%
I d.
A
a, io14
cm-3
OKh16N15M3B
A at 1050 ?C
1,8
265
16,5
steel
30 min
A-{-5?,ro CD
0,2
160
5,4
A-}-10?ro CD
0,2
100
29,4
A+20?,o.CD
0,9
75
34,0
MTW (A-{-10%
1,5
350
4,0
CD +g00 ?C, 3h )
TMW(A190%CD
1,5
230
14,0
at 650 ?C-~
=650 `C, 100 h)
A-}-80?,-o CD -{-
5,4
365
17,0
1800 ?C 30 min
OKh16N15
A at 1050 ?C,
2,1
340
6,6
OKh16N15M3
30 min
Same
2
0
320
6
6
OOKh16N15M3
"
,
1,3
250
,
4
6
OICh17N406
"
1,3
385
,
17,0
OICh20N60B
?
1,6
425
4,0
KhN77TYu
A at 1050 ?C,
0.01
i5
5,5
30 min=720 ?C,
2h
Experimental Results and Their Evaluation
Hydrostatic Weighing. The results of the density measurements on the materials (Table 2) show that
the maximum swelling of the OKh16N15M3B austenized steel at a fluence of 7.,1.1023 neutrons/cm2 (E > 0)
and a temperature of ^'550?C was 13,7%; at 10% CD the swelling was reduced to 0.4-7.0% (dispersion of the
results for four specimens), and at 20% CD it was practically absent. MTW and prior aging at a temperature
of 650?C had little effect on the swelling of the steel, while TMW reduced the swelling considerably (see Table 2).
The swelling of the nickel after irradiation to a fluence of 1.1.1023 neutrons/cm2 (E > 0) at a calculated
temperature of ^' 450? and 550?C did not exceed 4%, and the swelling at ^' 460?C and a fluence of 8.7. 1022 neu=
trons/cm2 was only 1.5%. A 10% cold deformation reduced the swelling of the nickel and iron to some extent,
but its effect was considerably less than the effect on OKh16N15M3B steel.
Electron Microscopy. Table 3 shows the values of the average concentration p, the pore dimension d,
and the swelling OV V, determined by the electron-microscopy method for materials irradiated to a fluence
of 7,5.1022 neutrons/cm2 (E > 0) or 5,9 ? 1022 neutrons/cm2 (E > 0.1 MeV) at a calculated temperature of X520?C,
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2,a
_,l--------------
.~
s ~a zo
?/? cold deformation
Fig, 1, Swelling, pore dimension and pore concentration (Q O, ^ ), and of lamellar precipita-
tions (~ , ?, ^) as functions of the equivalent cold deformation at A +20% CD (O, ?), MTW (~,
?), and TMW (O, ? ).
Fig. 2, Structure of OKh16N15M3B steel subjected to various kinds of treatment and irradiated
to a fluence of 7,5.1022 neutrons/em2 at 520?C {x 130,000): a) austenization; b) MTW; c) TMW;
d) fine grain,
OKh16N15M3B Steel in Various Structural States, The results of the electron-microscopic investigation
of the irradiated OKh16N15M3B steel in various structural states are shown in Fig? l., In plotting the experi-
mental points, we took account of the fact that the original density of the dislocations in these specimens cor-
responded to the density of dislocations in specimens with 5 and 10% cold deformation,
0
In the irradiated austenized steel we observed pores, prismatic dislocation loops measuring 200-?00 A,
dispersed precipitations (to 100 A) of Nb (C, N) and lamellar precipitations (Fig, 2a). The swelling amounted
to 1,8%. The pores were essentially unrelated to the precipitations,
In steel with 5% deformation, the pore dimension and concentration and the swelling were reduced (see
0
Fig. 1 and Table 3). In addition to the pores, we observed prismatic dislocation loops measuring ~ 500 A,
individual dislocations, and large lamellar precipitations, whose dimension and concentration increased in
comparison with the austenized steel (Fig, 3a), The pores are essentially connected with the lamellar pre-
cipitations.
The lamellar precipitations in the irradiated specimens of OKh16N15M3B steel deformed by 5 and 10%,
as well as in the specimens subjected to MTW and TMW, were investigated by the microfraction method. The
results obtained give reason to assume that the lamellar precipitations are Laves phases of Fee (Mo, Nb) with
a basis plane parallel to the plane of the plates, In 1.0% deformed steel the dimensions of the pores and the
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Fig, 3 Fig, 4
Fig, 3. Effect of prior cold deformation on structure of OKh16N15M3B steel after irradi-
ation to a fluence of 7,5 ? 1022 neutrons/cm2 at ~ 520?C: a, b, c) 5, 10, and 20% CD, respec-
tively: d) 20% CD, same segment, but with more clearly marked dislocations (x. 130,000).
Fig. 4, Structure of Kh16N15 type steel with various kinds of alloying after irradiation to
a fluence of 7,5 ? 1022 neutrons/cm2 at ~ 520?C: a) OKh16N1.5; b) OKh16N].5M3; c) OOKh16N15M3
(x 130,000),
connected lamellar deformations (presumably Laves phases) were reduced, while the concentration increased
(see Fig, 1 and Fig, 3b), it was impossible to estimate the dislocation structure in this case, apparently be-
cause of the high specific volume of the lamellar precipitations,
In 20% deformed steel we observed a further reduction of the dimensions of the lamellar precipitations
and the connected pores and an increase in their concentration (see Fig. 1 and Fig. 3c), We observed a high
density of linear dislocations and loops measuring up to 300 A (see Fig, 3d),
Before irradiation, in the specimens subjected to MTW we observed small-angle boundaries and dis-
location lattices decorated with dispersed precipitations (30-100 A) of Nb (C, N) whose concentration was
2,5 ? 1015 cm-3, After irradiation we observed large lamellar precipitations and large pores connected with
them (see Fig, 2b), Along the grain boundaries, as in the unirradiated specimens, we observed M23C6 carbides
measuring ^~ 0.3 ?.
In the specimens which had been thermomechanically worked, before the irradiation we observed in the
grain bodies dislocation lattices decorated with dispersed precipitations (up to 100 A) of Nb (C, N), and along
the grain boundaries M23C6 carbides measuring ~ 0,3 ?. After irradiation, we also observed M23C6 carbides
along the grain boundaries, while in the grain bodies we observed lamellar precipitations and pores connected
with them (see Fig, 2c), the concentration of which was larger than in the irradiated specimens subjected to
MTW (Fig, 2b), The swelling of the irradiated specimens after MTW and TMW was the same, and in both
cases we were unable to detect any precipitations of NB (C, N) or any dislocation lattice,
In fine-grained (3-5 ?) OKh16N15M3B steel we observed pores whose dimension was considerably larger
than in the irradiated specimens with ordinary-sized grains (30-50 ?) and many precipitations of excess phases,
which, for the most part, were there before the irradiation as well (see Fig. 2d), in Fig. 2d we can see apore-
deficient zone ~ 0.1 ? in width along the grain boundaries,
Steels with a Kh16N15 Base and Various Kinds of Alloying, After irradiation, OKh16N15 and OKh16N15M3
steels revealed large precipitations (0.2-0.5 ?) of M23C6 carbides along the grain boundaries. In the grain
bodies in OKh16N15 steel there were individual precipitations of M23C6 carbides measuring up to 500 A and
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 Declassified and Approved For Release 2013/03/22 :CIA-RDP10-021968000700120001-4
Fig, 5, Structure of high-nickel alloys irradiated to a
fluence of 7,5.1022 neutrons/cm2 at X520?C: a)OKh17N40B;
b) OKh20N60B; c) KhN77TYu (x 130,000).
pores (Fig, 4a) whose concentration was considerably higher than that of the precipitations; in contrast, the
OKh16N15M3 steel reveals only two systems of pores: large pores (450 f 50 ~i) and small pores (100 f 20 l~)
Fig, 4b),
In irradiated steel with reduced carbon content we observed a large number (4,5 ? 1014 cm-3) of large
(~ 20,000 A) lamellar precipitations (presumably Laves phases), to which were connected a number of large
pores (450 t 50 A), while the small pores (100 f 20 A) were distributed homogeneously throughout the body of
the grains (Fig, 4c).
High-Nickel Alloys, In the OKh17N40B alloy, the irradiation caused the formation in the body of the
?
grains of some comparatively large pores (see Table 3), dispersed particles (50-100 A) of IVb (C, N), and
large (^900 A) lamellar precipitations, and along the grain boundaries there was an elongated form of MzsCs
carbide measuring ^~ 1 ? (Fig, 5a). In addition to the precipitations and the pores, we observed prismatic
dislocation loops whose dimension and concentration were 270 A and 1,8.1015 cm-3, respectively, We did
not identify the lamellar precipitations in the irradiated OKh17N40B alloy and the OKh20N60B alloy,
The phase composition of the irradiated OKh20N60B alloy is the same as in the OKh17N40B alloy, but
the concentration of the lamellar precipitations is considerably higher (5.1014 cm-3), while the dimension
is somewhat smaller (~ 560 A) (Fig, 5b), The pores, whose average dimension and concentration were 425 A
and 4,0.1014 cm-3, respectively, are essentially connected with the lamellar p oecipitations, The concentra-
tion and dimension of the dislocation loops are equal to 8,3.1014 cm-3 and 350 A, respectively,
After irradiation, in the dispersion-hardened KhN77TYu alloy, along the grain boundaries, we observed
solid walls of M23C6 carbides, pores with small dimension and low concentration, and a large number of dis-
location loops with p = 6.4.1015 cm-3, a =180 A (Fig, 5c).
We were unable to detect the presence of pre-precipitations of a y' phase; probably, owing to the high
concentration of loops, to judge by the electronograms, there were no y'-phase precipitations,
An analysis of the results of the hydrostatic weighing of cylindrical specimens of OKh16N15M3B steel
in the austenized, cold-deformed state and after MTW and TMW showed that under irradiation to a fluence
of 1,1 ? 1023 neutrons/cm2 (E > 0) at a temperature of ^? 550?C the swelling is reduced proportionately to the
degree of cold deformation, i,e? the dislocation density in the specimens before irradiation, MTW led to the
appearance of dislocations decorated with Nb (C, N) particles in the specimens, with a dislocation density
equivalent to 5% cold deformation, but it did not bring any substantial reduction in swelling. After TMW the density
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 Declassified and Approved For Release 2013/03/22 :CIA-RDP10-021968000700120001-4
of dislocations decorated with NB (C, N) particles was equivalent to 10% cold deformation, and here we ob-
served asubstantial positive effect: The swelling was reduced by a factor of about 3.5 in comparison with
steel in the austenized state.
Electron-microscopic investigation of the specimens irradiated to a fluence of 7.5.1022 neutrons/cmZ
(E > 0) at a temperature of ^-520?G showed that under the indicated conditions even 5 and 1~ cold deformation
is sufficient to reduce the swellingby a factor of about 10. However, account must be taken of the fact that
such cold deformation does not yield a uniform distribution of the density of dislocations over the grains of
the specimen, a fact confirmed by the dispersion in the swelling values for the cylindrical specimens with
10% cold deformation (see'Table 2), The results of the electron-microscopic investigation relate to deformed
grains, Since the total density of dislocations in the irradiated specimens increases with increased degree of
prior deformation, the reduction in swelling of the deformed specimens may be connected with the increase in
the migration of point defects to the dislocations.
In austenized OKh16N15M3B steel, pores are generated and grow under irradiation without any visible
connection with precipitations of excess phases, whereas in cold-deformed steel and after MTW and TMW
we observed a connection between the pores and the lamellar precipitations (presumably Laves phases) . The
nature of the variation of the dimension and concentration of pores coincides with the- nature of the variation
in the dimension and concentration of lamellar precipitations, i.e? the refinement of lamellar precipitations
that is observed in the deformed specimens corresponded to the refinement of the pores connected with them
and the reduction in swelling:
in the specimens subjected to MTW and TMW, before irradiation there were Nb (C, N) precipitations
connecting the dislocations, thus preventing the migration of point defects, Earlier [1] it was shown that such
precipitations serve as centers for the generation of lamellar precipitations, Therefore, after MTW and TMW
we should expect the generation of vacancy pores and lamellar precipitations and their growth during the ir-
radiation process to be more rapid than in deformed specimens. We did in fact find that after irradiation to
a fluence of 7,5.1022 neutrons/cm2 (E > 0) the swelling in OKh16N15M3B steel in the austenized state and after
MTW and TMW was about the same (see Table 3), and after irradiation to a fluence of 1.1.1023 neutrons/cm2
(E > 0) the specimens were found after TMW to swell much less than the austenized specimens (see Table 2),
as a result of the high initial density of dislocations.
The swelling of OKh16N15M3B steel with fine (3-5 ?) grains is almost 3 times as high as the swelling in
steel with ordinary (30-50 ?) grains. The dimension of the depleted zones (see Fig, 2d) was ~ 0,1 ?, i,e? not
commensurable with the dimension ofthegrain, At such dimensions (3-5 ?), apparently, the conditions for the
annihilation of vacancies and displaced atoms are violated, it may be assumed that the displaced atoms, pos-
sessing high mobility, move out to the grain boundaries and create a supersaturation with respect to vacancies,
thereby increasing the swelling.
We should also take account of the depletion of the solid solution in alloying elements and interstitial
impurities as a result of the precipitation of a large number of excess phases after treatment on the fine
grains, which can also increase the swelling [2],
In the OKh16N15 and OKh16N15M3 steels we observed relatively few precipitations in the body of the
grains, and the pores were not connected with them. This again confirmed the viewpoint that in unstabilized
steels the generation of lamellar precipitations is difficult [3], The concentration of pores in these is less
by a factor of about 3, and their dimension is larger, -than in the OKh16N15M3B steel, The reduction in the
carbon content of the OOKh16M15M3 steel led to the appearance of large lamellar precipitations (presumably
Laves phases) and the generation and growth of pores on them, The large pores are connected with the lamellar
particles, while the system of small pores is generated homogeneously,
An analysis of the effect of the chemical composition of steels of the KhN16N15 type showed that slower
or faster precipitation of lamellar phases does not always determine the nature of the generation and growth
of vacancy pores and the amount of swelling, However, since a change in the chemical composition of the
steels may shift the maximum-swelling temperature, it is impossible to draw any firm conclusions about the
effect of the lamellar precipitations on the swelling of steels,
The model metals -nickel, and especially iron -swelled much less than steels of the 16-15 type under
irradiation to a fluence of 1,1.1023 neutrons/cm2 (E > 0), There are practically no published data concerning
the swelling of these metals under neutron irradiation to high fluences, After irradiation with 5-MeV nickel
ions in an accelerator at 625?C with a dose of 140 displacements per atom, the swelling of nickel was 12%, and
the swelling of 316 steel was ~ 30% [4], If we compare these data with those obtained in our investigations, we
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 Declassified and Approved For Release 2013/03/22 :CIA-RDP10-021968000700120001-4
are led to suppose that nickel and a iron, as well as alloys in which these metals predominate, have less of a
tendency to swell than austenitic steels of the 16-15 type. Unlike the case of steels, cold deformation did not
have a marked effect on the swelling of nickel and a iron, possibly because of the more rapid annealing of the
dislocations under irradiation.
The swelling of alloys with 40 and 60% nickel (OKh17N40B and OKh20N60B) after irradiation to a fluence
of 7,5 ? lOZZ neutrons/cm2 (E > 0) at 520?C is about the same (1,3% and 1,6%), despite the fact that in the OKh20N60B
alloy the concentration of lamellar precipitations is 10 times as high and there is a heterogeneous generation
of pores, The addition of titanium and aluminum to the KhN77TYu alloy led to the appearance of y'-phase pre-
precipitations after heat treatment. The dimension of the pores in this alloy after irradiation was less by a
factor of 5-6, and the swelling was less by a factor of 100, then in alloys with 40% and 60%. nickel. Comparing
these results with the data for austenitic steels, which swell slightly more than alloys with 40 and 60% nickel,
we can assume that for such alloys one of the most important factors determining the tendency to swelling is
the phase. composition, especially the presence of y'-phase pre-precipitations.
The favorable effect of finely dispersed particles of y'-phase on the swelling of alloys of the PE-16 type
has been noted in the past [5, 6]. However, some data- [7, 8] indicate that the separation of a y'-phase either
during the irradiation or after aging leads to an increase in the swelling because of the depletion of titanium
and aluminum atoms in the solid solution.
C ONC LUSiO NS
The structure created by preliminary treatment has a strong influence on the swelling, the nature of
pore generation, the dislocation density, and the phase composition of OKh16N15M3B steel, -The greatest re-
duction in swelling was produced by 20% cold deformation, which at a high dislocation density led to a sharp
refinement of the pores. and lamellar precipitations.
The lamellar phases precipitated in steels of the Kh16N15 type during the irradiation after special alloy-
ing or preliminary treatment do not always determine the nature of the generation and growth of vacancy pores,
At a temperature of 460 and 550?C the swelling of nickel is less by a factor of about 3, and the swelling
of a iron is less by a factor of about 10, than the swelling of austenized OKh16N15M3B steel.
The least swelling was found in the Kh77TYu alloy (with 77~c nickel), which before irradiation contained
y'-phase pre-precipitations, whereas ;the swelling of alloys with 40 and 60% nickel is 100 times as great, From
this it is clear that the nickel content is not the only factor determining the swelling.
LITERATURE CITED
1, N. P. Agapova et al? izv, Akad. Nauk SSSR, Ser, Fiz? 38, No, 11, 2351 (1974),
2. J. Leitnaker, E, Blom, and. J, Stiegler, J. Nucl. Mater, 49, 57 (1973/74),
3, B, Wiess and R, Stickler, Metall. Trans., 3, 851 (1972),
4, W. Johnston et al? J. Nucl, Mater, 54, 24 (1974).
5. R, Bullough and R. Perrin, in: Proc. Reading Conf. on Void Formed by Irradiation of Reactor Materials,
Harwell, BNES, (19?1), p. 79,
6. J. Buswell et al? in: Proc. Intern. Conf. "Physical Metallurgy of Reactor Fuel Elements," E. Sykes
.(editor), London (].975), p. 170.
7, J. Gittus, J, Watkin, and I, Stand ring, in: Progr, with Abstracts of intern. Conf. "Radiation Effects in
Breeder Reactor Structural-Materials," Scottsdale (1974), p, 22.
8. W. Johnston and J. Bates, .ibid., p. 35,
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 Declassified and Approved For Release 2013/03/22 :CIA-RDP10-021968000700120001-4
MEASUREMENT OF THE 237Np/23sPu AND z41Am/z3sPu
FISSION CROSS SECTION RATIOS
FOR 0.1.3-7.0 'MeV NEUTRONS
V. M, Kupriyanov, B, I. Fur.sov, UDC 539,173.84
V, i. Ivanov, and G, N, Smirenkin
The method of measuring fission cross section ratios is described in detail in [1-8]. The measurements
were performed at electrostatic accelerators, Neutrons from the 7Li(p, n)7Be, T(p, n)3He, and D(d, n)31ie re-
actions bombarded solid targets consisting of layers of 239Pu, 237Np, and z4iAm oxides of practically 100% iso-
topic purity 0.2-0,4 mg/cm2 thick deposited on thin aluminum or platinum backings,
Tlie energy dependence of the fission cross section ratios was studied in detail by the ionization method,
and the components of the neutron background were measured as described in [2, 3]. Then, using the glass
detector method [3], which permits. an accurate determination of the ratios of the efficiencies of countingfis-
sion fragments, the absolute values of the fission cross section ratios were measured- for En= 2,0, 2,5, and 3,0
MeV, The -final results were obtained by normalizing the data on the energy dependence to the absolute values,
The 237~/23sPu and z4tAm/2sspu fission cross section ratios were measured simultaneously by using
two back-to-back fission chambers [2]. The ratios of the numbers of fissionable nuclei were determined by
comparing the alpha activities of the layers, The use of 239Pu instead of 23~U, whose cross section is thestan-
dard, was dictated by the shortcomings of 2~U as an alpha emitter: The low specific activity, particularly in
comparison with z4iAm, the complex ~-particle spectrum, and the considerable partial activity of the admix-
ture of 234U, whose content in highly enriched z3sU is difficult to determine with acceptable accuracy. At the
same time the 237Np/239Pu and 291Am/239Pu fission cross section ratios can be converted to z37 z35 241-
Np/ U and Am/
zssU fission cross section ratios without appreciable loss of accuracy by multiplying by the 239Pu/235U fission
cross section ratio measured by the authors with an error of ~ 1,5% [3] by using the same procedure,
The c2 activities of the fissionable layers were compared to good geometry by using a semiconductor
detector, The measurements performed at the beginning and end of the whole cycle of experiments agreed
within the limits of error, and were averaged. The total error in determining the number of fissionable nuclei
was 0,9% for 237Np/298pu and 1,3% for 241Am/23spu, taking account. of the uncertainties of the half-lives (2.41 f
0.01) ? 104 yr for 239Pu [4]; (2,14 f 0.1) ? lOs yr for 237Np [5], and 432 f 4 yr for z4lAm [4], -
The ratio of efficiencies of counting fragments was easily found by using glasses, since the geometric
factors were kept very closely the same, and the small correction to take account of the dependence of the
efficiency on the angular anisotropy of fission was calculated by using the values of the anisotropy measured
in the same experiment.
Since the ionization detector was used only to study the-energy dependence of the fission cross section
rations, it was not necessary to determine the efficiencies of the fission chambers. The dependence of the
efficiency ratio on neutron energy was taken into account by a small correction (< 2.3%) calculated by allow-
ing for the momentum introduced by a neutron, and the angular distributions of the fragments, _
The correction to take account of neutron scattering by the target structure was greatest near the fis-
sion threshold, and at En = 0,13 MeV it reached 12.0 and 4.2% for af2~Np/Qf239pu and Qf24iAm/Qf239pu, respec-
tively, in the 1,5-7.0-MeV range this value varied from 2,2 to 0,2%,
The background of accompanying (d, n) reactions reached 30% at En=7.0 MeV, The correction to the
fission cross section ratios resulting from this background was 0.2-8.1% for En=3.6-7,0 MeV, The analogous
correction for (p, n) reactions did not exceed 1,2% for Ens 3 MeV, The correction for the background in the
laboratory did not exceed 0.3-0,4% over the whole energy range studied.
In measuring absolute values, the glass detectors were- irradiated alternately from both sides to aver-
age the neutron flux and eliminate the necessity of introducing a correction for the elastic scattering of neu-
Translated from Atomnaya ~nergiya, Vol,. 45, No, 6, pp: 440-442, December, 1978. Original article
submitted April 17, 1978,
1176 0038-531X/78%4506-1176$07.50 ?-1979 Plenum Publishing Corporation
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 Declassified and Approved For Release 2013/03/22 :CIA-RDP10-021968000700120001-4
TABLE 1, zsr~/zsaPu and ""Am/`""Pu Fis-
sion Cross Section Ratios Measured by Glass
Method .
n.
AEn
~
~ I
of/a1
I
~(aflaf). % I
aft/ajI
Daft/aj),
MeV
teV
2,0
2
5
60
72
0,832
0,860
1,4
1,5
0,935
0,970
2,0
2,0
,
3,0.
78
0,873
1,5
.0,997?
2,1
TABLE 2, Corrections and Experimental
Errors of Measurements of z~Np/z~Pu and
zafp~/zsspu Fission Cross Section Ratios
(Glass Method, En=3 MeV), %
ss~Np/zsapu
zaiAm/zsaPu
Sources of correctians
cor-
cor-
and errors
rec- l
ector
rec- t
error
lion
lion
Ratio of Nos. of fissionable
nuclei
-
0,9
-
1,3
glass
Scannin"g of
,
detectors
-
0,5
-
0,5
Statistical
-
9
0
-
1,4
error
Angular anisouopy of
,
fission
0,5
0,3
0,8
0,3
Fission of minority
2
0
1
0
0
2
0,1
isotopes
,
,
,
.
Difference of neutron
-
0 2
-
0,2
fluxes through la er
Inelastic scattering of
-
0,4
-
0,4
neuuons
Neutron background in
2
G 0
2
0
G 0,2
0,2~
laboratory
,
,
Bkgd. of neuuons scattered
from target structure
0,6
0,3
1,4
0,3
Bkgd, of neutrons from ac-
companying(p,n)reac-
0,8
0,2
1,2
0,3
lions
Total error
, -
~ 1,5
-
~ 2,1
trons by the backings, No correction was introduced for the inelastic scattering of neutrons, but an upper
estimate of its possible effect (0,3-0,4%) was included in the 'value of the experimental error.
The absolute values of the 237Np/z~Pu and za1Am/zssPu fission cross section ratios measured by the
glass method are listed in Table 1, The value listed for the. total experimental error is the rms sum of the
uncertainties, Table 2 shows the structure of the characteristic corrections and errors for En=3 MeV, The
results of the ionization chamber measurements are shown in Table 3 for the whole range of energies studied,
The total error listed in Table 3 is the rms sum of the uncertainty of the energy dependence data obtained
by the ionization method (statistical error and the errors related to the, experimental and calculational cor-
rections), the average error of the absolute values of the ratios of Qfzs7Np/QfzzsPu (1,5%), QfzaiAm/vf239Pu (2%),
and the uncertainty in the normalization of the energy dependence data to the absolute values (0.25%).
For En?6 MeV the increase in the total error results from the increase in the background of accompany-
ing (d, n) reactions, and for Ens 1 MeV from the decrease in statistical accuracy and the increase in the.error
related to the correction for the background of neutrons scattered by the target structure.
To compare our data with the results of other experiments, the Qfz37Np/Qf23sPu and Qfz4iAm/trfz~Pu ratios
were converted to Qfz37Np/QfzssU-and Qfz4fAm/QfzssU ratios by multiplying by the z~Pu/z35U fission cross sec-
tion ratio [3]. This procedure increases the total error of the results by 0.3-0.5%.
There is good agreement with the data of [6, 7] for the 237Np/z35U fission cross section ratio. For En=
1,5-6 MeV there is a systematic deviation from the preliminary data of [8], but below 1.5 MeV the two results
are in general agreement when possible small differences in the calibration of the energy scale are taken ittto
account, For En> 6 MeV the energy dependences of the results are different.
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 Declassified and Approved For Release 2013/03/22 :CIA-RDP10-021968000700120001-4
TABLE 3. 2s~~p/23sPu and z41Am/Z~Pu Fission Cross Section Ratios
En MeV I
keV ~
I ?flO1
I e (Q olaj), I
Q~1/Q~
n (af/lof).I
I En: McV
I keVn
Q f/Qt I
e (a~/uj). I
Q~ /Q~ 1
I
e (oj alof).
0;130
21
0,015
4,8
0,012
4,3
1?,90
64
0,830
1,7
0,936
2,3
0,180
20
0,022
4,3
0,016
4,4
2,00
66
0,836
1,8
0,937
2,2
0,230
19
U,025
4,1
0,019
4,2
2,10
68
0,841
1,7
0;948
2,3
0,280
18
0,034
4,0
0,024
4,5
2,20
69
0,841
1,8
0,946
2,2
0,350
38
0-,068
3,5
'0;030
3,6
2,30
70
0,844
1,8
0,956
2,2
0,400
37
0,107
2,9
0,037
3,2.
2,40
71
0,854
1,8
0,960
2,2
0,450
36
0,17~e
2,7
0,050
3,3
2,50
73
0,862-
1,7
0,970
2,2
0,500.
34
0,265
2,6
0,064.-
- 3,1
2,60
75 _
_0,861_
?1,7
0;971
2,3
0,5.
33
0,356
2,4
0,091
3,3
2,70
77
0,863
1,7
0,977
2,2
0,600
32
0,441
2,3
0,112
3,3
2,80
78
0,868
1,7
0,984
2,2
0,650
32
0',539
2,6 ~
0,171
~" 3,2
2,90
80 .
0,868.'
1,7
0,984
2,2
0,700
32
0;613
2,3
0,232
3,3
.3,00,
82
0,866
~1,7
0,994
2,2
0,750
32
0,650
2,3
0,314
3,4'
3,60
190
0,858
1,8
1,027?
2,4
0,800
31
0,686
2,3
0,388
~ 3,2
3,80
180
0,864
1,8
1,027
2,4
0,850
31
0,731
2,1
0,494
3,1
4,00
!45
0,859
1,8
1,038
2,4
0,900
31
0,763
2,1
0,565
3,2
4,20 ~
140
0,865
1,8
1;049
2,4
0,950
31
0,781
2,0 -
0,658
3,1
4,40
130
0,861
1,7
1,055
2,4
1,00
31
0,810
1,9
0,734
2,7
4,60-
128
0,861
1,7
1,061
2,3
1,05
34
0,814
1,9
0,786
2,8
4,80
124
0,860
1,7
1,066
2,4
1,10
37
0,812
1,9
0,807
2,7
5,00
125
0,854
1,7
1,062
2,3
1,15
40?
0,812
1,9
0;857
2,8
5,20
128
0,859
1,7
1,066
2,3
1,20
42
0,811
1,9
2,876
2,6
5,40
130
0,852
1,8
1,072
2,3
1,25
44
0,807
1;9
0,902
2,6
5,60
133
0,857
1,9
1,077
2,5
1,30
45
0,804
1,8
0,907
2,6
5;80
136
0,874
1,9
1,072
2;5
1,35
46
0,805
1,8
0,911
2,6
6,00
140
0,905
2,0
0,059
2,6
1,40
47
0,809
1,7
0,921
2,6
6,20
144
0,939
: 2,2
1,051
2;8
1,45
48
0,811
1,7
0,924.
2,5
6,40
-
150
0,974
2,2
1,070
2,7
l,50
49
0,814
1,7
0,928
2,3
6,60
157
0,986,.
2,5
1,081-
2,7
1,60
58
0,820
1;7
0,924
2;4
6,80
165
0,993
2,3
1,056
2,7
1,70
59
0,828
1,7
0,928
2,3
7,00
172
0,983
2,5
1,064
.
2,9
1,80
6i
0,835
1;7
0,931
2,4
The data on the Qfz41Am/vg235U ratio [9], reduced to the half-life of z41Am (432 f4 years), vary more
rapidly with energy, particularly for En> 2 MeV, Preliminary results in [10] in the 1-6 MeV range are sys-
tematically higher than our values by 3-7%. Below 1 MeV the agreement ?of the results is very good, but for
En > 6 MeV there is a difference in the shape of the energy dependence, as fs the case for Qf237Np/Qf2ssU. This
leads to the conclusion that the difference results from the shape of the zssPu/zssU cross section ratio for
En > 6 MeV,
The authors thank S, E, Lavrov, F, B. Samylin, and M, K, Golubevaya for participating in the work,
LITERATURE CITED
i, B, i, Fursov, V. M, Kupriyanov, and G. N, Smirenkin, in: Proceedings of a Conference on Neutron
Physics [in Russian], Vol, 3, TsNliAtominform, Moscow_ (1977), p, 144,. ?
2, B, i, Fursov et al? At, Energ? 43, 181 (1977),
3, B, I: Fursov et al? At, Energ? 43, 261 (1977),
4, S, A, Baranov, A, G, Zelenkov, and V. M, Kul~kov, At, Energ? 41, 342 (1976),
5, F. Brauer et al,, J, inorg, Nucl, Chem., 12, 234 (1960), ?
6. W, Stein, R. Smith, and H, Smith, in: Proceedings of the Conference on Neutron Cross Sections and
Technology, Vol, 1, Washington (1968), p, 627,
7, P, White, J, Hodgkinson, and G, Wall, in: Proceedings of the Symposium on Physics and Chemistry
of Fission, Vol, 1, Salzburg,-May 22-26, (1965), p, 219; P. White and G. Warner, J, Nucl, Energy, 21, .
671 (1967),
8.. J, Behrens, J, Magana, and J. Browne, Report UCID-17370 (1977).
9, ? ~, F, Fomushkin and E, K, Gutnfkova, Yad. Fiz? 10, 971 (1969),
10, T, Behrens and J, .Browne, Report UCID-17324 (1976).
Declassified and Approved For Release 2013/03/22 :CIA-RDP10-021968000700120001-4
 Declassified and Approved For Release 2013/03/22 :CIA-RDP10-021968000700120001-4
EXPERIMENTAL RESEARCH ON THE PRODUCTION
AND STORAGE OF ULTRACOLD NEUTRONS
V. i, Morozov UDC 539,125,5
Ultracold neutrons (UCN) have energies of 10-~ eV or less [.1], Since 1968, when the first report [2] on
the production of UCN was published, the number of experimental and theoretical studies of their properties
has grown continuously, The state and prospects of research with UCN are analyzed in detail in [3-6], Dur-
ing the last few years, a great many experimental papers on UCN have been published, A majority of them
deal with techniques for producing intense sources of UCN and storing them in traps of various kinds, in the
present paper we give a brief summary of recent experimental research in the USSR on the production and
storage of UCN.
The commonest method of obtaining UCN is to extract them directly from a moderator by guide tubes
with reflecting walls, In the horizontal method of extraction the experimental room is connected to a reactor
core by an evacuated bent horizontal guide tube with walls having a high limiting energy Elim,n=h2~coh/
2mn in the range of UCN, where N is the number of nuclei per unit volume, bcoh is the coherent scattering
length, m is the mass of a neutron, and h is Planck's constant, An auxiliary moderator (converter) is located
in the guide tube close to the core. When the converter is irradiated with thermal neutrons. part of them lose
energy as a result of inelastic scattering and enter the range of UCN, The UCN are extracted from the con-
verter as a result of their total reflection from the vacuum-medium. interface for arty angles of incidence if
E < Elim of the medium. The UCN emerging from the converter reach the guide-tube exit after a number of
reflections from the walls of the tube. if the guide tube is long and has sharp bends the spectrum of the ex-
tracted neutrons has an upper bound E =Elim,n? The flux of UCN at the guide-tube exit is determined by the
flux of UCN from the converter and the transmission of the guide tube. The neutron flux from a unit area of
the converter in the energy range from E to E +dE is
dd? - LTG (Tc . Tn i E dE. (1)
4
where ~r is the thermal neutron flux density, G(Tc, Tn)= cool(~cap+oheat)-'Tn zE-1' ocool is the cross sec-
tion for the production of neutrons of energy E, averaged over the thermal neutron spectrum, Qcap is the
cross section for the capture of neutrons of energy E,heat is the heating cross section equal to oinel -the
cross section for the inelastic scattering of neutrons of energy E, and Tn and Tc are, respectively, the tem-
peratures of the neutron spectrum and the converter [3]. For Tc =Tn and o'cap? ?~heat the neutron flux from
the converter corresponds to the flux in the equilibrium Maxwellian spectrum:
dcU = cA,.E dEl4Tn . (2)
Equations (1) and (2) are valid when the limiting energy of the converter material is zero, For Elim,c ~ 0 the
flux will be smaller because of the reflection of UCN from the converter-vacuum interface, and its spectrum
has the lower bound E =Elim,c~ since neutrons in a vacuum acquire an additional energy Elim,c? in view of
this the converter material should have a small value of Elim,c in comparison with Elim,n in order to obtain
a sufficiently broad spectrum of UCN.
Technically it is simplest to extract UCN from a converter when its temperature is close to that of the
neutron spectrum (not cooled to low temperatures), Techniques for extracting UCN from uncooled converters
are described in [7-15] where tests of practically all the converter materials suitable for horizontal guide
tubes (water, aluminum, magnesium, zirconium hydride) are reported. These experiments showed that a
zirconium hydride converter is optimum from the point of view of yield of UCN, radiation stability, and the
possibility of using it uncovered fn a vacuum, Stainless steel with its large limiting energy (~1,9 ? 10-~ eV)
and high corrosion resistance in strong y fields turned out to be the most suitable material for guide tubes,
Figure 1 shows a schematic diagram of one of the best horizontal guide tubes, constructed at the SM-2
reactor [14]. The guide tube is 70-90 mm in diameter, 5,5 m long, and is made of electropolished stainless
Translated from Atomnaya ~nergiya, Vol, 45, No, 6, pp. 442-449, December, 1978. Original article
submitted May 22, 1978,
0038-531X/78/4506-117907.50 ?1979 Plenum Publishing Corporation 1179
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Fig, 1 Fig, 2
Fig, 1, Schematic diagram of arrangement for obtaining UCN from the SM-2 reactor: ]) core;
2) converter, 3) entrance region of guide tube; 4) reactor shield; 5) additional shielding; 6) vac-
uum valves; 7) exit pipes; 8) UCN flux commutator.
Fig. 2. Schematic diagram of vertical guide tube for extracting UC N from the V VR-M reactor:
1) core; 2) beryllium converter; 3) reactor shield; 4) mirror guide; 5) screen for determining
background; 6) detector.
steel, The converter is a thin plate of zirconium hydride with Tc =350?K, irradiated with a thermal.neutron
flux density of (2-4) ? 1014 neutrons/cm2 ? sec. The total flux of UC N with velocities from 3.2 to 5.7 m/sec at
the guide-tube exit where the cross-sectional area is 60 cm2 is 2 ? 103 neutrons/sec.. The flux~of extracted
neutrons with velocities above 5.7 m/sec does not exceed 50 neutrons/sec.
In using a vertical variation of the extractor [5, 9, 16], a wider range of converter materials is possible.
Since neutrons.from.the converter can be slowed down in the gravitational field to energies.in the UCNrange,
it is possible to use converter materials with a high limiting energy (n rising 1 em a neutron loses ^~ 10-9
eV of kinetic energy), One such material is beryllium (Slim=2,4.10-7 eV, Qcap? oinel) which has a higher
radiation resistance than zirconium. Beryllium was used as a converter in the vertical guide tube of the
VVR-M reactor [9] (Fig, 2), -The guide tube consists of a vertical portion 5,5 m long, a smooth bend, and a
horizontal portion 3 m long. The entrance section of the-guide tube is 60x 70 mm in c"ross section and is
made of stainless steel; the horizontal part of the tube is 60 x 60 mm in cross. section, and the bend is made
of glass mirrors (3Ni dusted- on glass) . For a thermal .neutron flux density of (3-5) ? 1013 neutrons/cm2 -
sec in the converter region there were 1500 UCN/sec at the. exit of the .horizontal part of the guide. tube with
velocities between 3.2 and 6,8 m/sec. There were 8500 neutrons/sec with velocities between 6.8 and 20 m/
sec,
The g,~ide tubes described in [9, 14] illustrate the possibilities of the method of extracting UC N by using
uncooled converters of the most suitable materials (beryllium, zirconium hydride), it is shown in [9j that by
raising the quality of the guide-tube surface and improving its transmission, the flux density of UCN extracted
with velocities Between 3,2 and 6,8 m/sec can in practice be brought up to 100 neutrons/cm2 ? sec for ~ =1014
neutrons/cm2 ? sec. However, the suitability of a guide tube for experiments with UC N is determined not only
by the magnitude of the extracted neutron flux but also by its spectral-composition. Guide tubes which extract
a flux which is sharply limited at E =Elim.n are more suitable for experiments on the storage of UCN, spec-
tral measurements, and the formation of monoenergetic lines,- Slightly curved mirror guides having a high
transmission for UCN but furnishing a large number of neutrons with E >. Elim.n are suitable only for experi-
ments in which the presence of a background of very cold neutrons is unimportant, An increase in the ex-
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tracted flux of UCN resulting from an improvement of the transmission of guide tubes will be accompanied
by an increase in the ratio of background to effect, and this makes sense only in certain cases,
Low-temperature converters offer much more promise for increasing the flux of UCN [17, 18]. As
Tc is decreased relative to Tn, both ?heat and Qcool decrease, but Qcool falls more slowly and G(Tc, Tn)
increases,
According to calculations in [17, 18] the increase in the flux of UCN should be particularly large if
cap remains appreciably smaller than cheat during cooling, and the converter material has a high Debye
temperature. These requirements are satisfied by beryllium for which G(Tc, Tn) should increase by afactor
of 14 for cooling from 300 to 50?K [18].
A low-temperature beryllium converter was used for the first time in an arrangement for extracting
UCN from the VVR-M reactor [19]. When the converter was cooled from 300 to 30?K by gaseous helium the
flux of UCN was increased by a factor of 8, The flux of UCN was (1,5-2) ? 104 per sec with velocity components
in the following ranges: vz (along the axis of the horizontal portion) from 3.2 to 6.8 m/sec, vx and vy (per-
pendicular to the axis) from -6,2 to +6.2 m/sec. The count of background neutrons with vz > 6.8 m/sec was
increased by a factor of 6-7, This arrangement with aloes-temperature beryllium converter is the most in-
tense source of UCN in existence at the present time,
An appropriate material for slow-temperature converter in a horizontal guide tube was sought in [17,
20, 21], According to data in [17] the cooling of zirconium hydride was ineffective, Cooled gases in an alu-
minum ampul and hydrogenous materials (water and alcohol) frozen onto a cooled backing were investigated
as converters in the horizontal guide tube of the VVR-K reactor [20, 21], Water and hydrogen were the best
of the converters tested, having identical yields at 300?K, After cooling hydrogen in the ampul to 80?K and
freezing water onto bare backing at 80?K the yield of UCN from hydrogen increased by a factor of 3.2, and
that from frozen water by a factor of 17 in comparison with water in an ampul, Taking account of the reflec-
tion of UCN from the window of the ampul the observed increase in the yield from water was almost twice as
large as the value 5.8 calculated in [17],
These results show that the optimum material for aloes=temperature converter in a horizontal guide
tube is water frozen onto a backing. A hydrogen converter is less efficient because of the reflection of UCN
from the aluminum window which is required to separate the gas from the evacuated guide tube, A parahydro-
gen converter appears to be rather promising, According to preliminary data [21] its yield at 80?K is 3 times
as high as that of ordinary hydrogen.
The results in [19-21] show that by using low-temperature converters of beryllium and water frozen
onto a backing at stationary reactors with ~`=1014 neutrons/emz ? sec the density of the extracted fluxes of
UC N with velocities from 3.2 to 6,8 em/sec can be increased to about 103 neutrons/cm2 ? sec. Further tech-
nically realizable ways of improving the method of extracting UC N by using guide tubes are not evident, For
this reason there is great interest in the development of methods of obtaining UC N by using pulsed reactors,
A recently developed method for obtaining UC N by storing them in portable traps [22-24] is promising,
Experiments have been performed [22] on the storage of UCN in quartz and beryllium bottles which were in-
serted into the core of a stationary reactor and then together with their contents of UC N withdrawn from the
core to the surface of the shield, This method is particularly promising with a pulsed reactor, if the neu-
tron bottle is open to the entrance of UCN at the beginning of the reactor pulse and closed at the end, the den-
sity of the stored UC N will be determined by the peak neutron density in the moderator, and this is appreciably
greater than the average.
This method was first realized in practice at the IIN reactor [24], The UCN were stored in a copper
bottle 40 cm in diameter and 40 cm high (Fig, 3). The lid of the neutron bottle is afast-acting shutter with
an .operating time of 20 msec, Over the shutter there is a polyethylene moderator and a cooled polyethylene
converter, After a reactor pulse of 2 msec, during which 1017 neutrons are emitted, the neutron bottle is con-
veyed to the measuring room where measurements are begun after 70-80 sec, At the beginning of the mea-
surements 103 neutrons remained in the bottle (density of 20 neutrons/liter), while immediately after the re-
actor pulse the bottle contained 5.103 neutrons, This density is of the same order of magnitude as that
achieved in filling neutron bottles from the guide tube of the SM-2 reactor [14] using an uncooled converter
(15-25 neutrons/liter), However, with further improvement of the trap there is the possibility of increasing
the number of stored UCN by another factor of 30, and by using the GIDRA reactor the density may be in-
creased by more than 3 orders of magnitude [25],
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0 7D 10 30 40 50
.Fig, 4
Fig, 3, Arrangement for storing UCN at the IIN reactor: 1) reactor vessel; 2) converter
cooling system; 3) converter; 4) neutron bottle; 5) moderator; 6, 7) fast shutter; 8) slow
shutter; 9) shaft; 10) sylphon assembly; 11) solenoid; 12) device for moving filters; 13) de-
tectors, "
Fig. 4. Schematic diagram and results of an experiment on the anomalous heating of UCN:
a: 1) entrance shutter; 2) entrance pipe; 3) UCN bottle; 4) external proportional counter; 5)
cadmium; 6) boron carbide; 7) exit shutter; 8) borated polyethylene; 9) UCN detector; b: time
dependence of UCN stored in bottle and neutron counts of external detector.
While there has been considerable success in producing a dense gas of UCN, the problem of storing it
for a long time is still unsolved. Study of the storage process, stimulated by the possibility of using U,CN in
experiments to determine the basic characteristics of the neutron (dipole moment, charge, lifetime), has
acquired independent interest, The main reason for the increased attention to the problem of storing UCN
is the anomalously short storage times found experimentally,
Theoretically [3, 26] the escape of UCN from a bottle can occur as the result of capture and inelastic
scattering in collisions with the bottle walls. The effective absorption coefficient of UC N per collision, aver-.
aged over the angles offncidence,is
k (acapf tea')
~, - arc sin y -
4nbcoh
where y2 = ~/Elim and k is the wave number.
However, in almost all the experiments performed on the storage of UCN [7-9, 13, 14, 24, 27=29] the
storage times actually attained were considerably shorter than those calculated by using. data on the cross -
secttons vcap and oinel extrapolated into the UCN region. The minimum experimental value of the .cross sec-
tion for the "removal" of UCN from neutron bottles is ~ 2.10~b for v =10 m/sec, including materials with
very small vcap and oinel cross sections -glass, beryllium, graphite,. etc. In view of this, the attainable
storage time of UCN does not exceed 300 sec even for neutrons with E? Elim stored in a bottle of optimum
geometry [29], So short a storage time does not permit the use of UCN to solve one of the fundamental prob- _
lems -the precision measurement of the neutron lifetime from the decrease. of density of a neutron gas stored
in a bottle [3].
The observed anomalous leakage is difficult to account for by the presence of a~ strorgly absorbing
surface impurity, since the experiments were performed with bottles of various materials processed in differ-
ent ways. The, assumption of a hydrogenous film or an admixture of ~ 10?lo hydrogen 100 A thick on the bottle
walls seems more probable. In this case a strong dependence of the storage time on the temperature of the
bottle walls should be observed, but experiments on the storage of UCN in bottles whose wall temperatures
were varied from 120 to 700?K did not show such a dependence.
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Fig. 5, Schematic diagram of trap: 1, 2) horizontal
and vertical mirrors; 3) magnetic plug.
The results obtained lead to the conclusion that heating and capture are improbable causes ~of the anom-
alous leakage of UGN from bottles, However, a direct experiment reported in [30] showed that the anom-
alously short storage time, at least for copper bottles, is due to the heating up of the UCN to thermal energy.
A schematic diagram of the experimental arrangement is shown in Fig. 4a. A copper bottle for storing UCN
was located inside asix-chamber proportional helium detector whose efficiency for recording thermal neu-
trons was X100%, One of the bottle shutters admitted UCN, and the second released the remaining UCNto
the gas proportional detector, The purpose of the experiment was to observe the heating up and leakage of
the UCN by means of the external detector.
Figure 4b shows the storage curve of UCN in a bottle with a storage time of 23 sec. This storage time
is close to the maximum attainable for such bottles, The figure also shows the time dependence of the neu-
troncounting rate of the external detector, The two curves have the same slope, and' the .number of neutrons
recorded by the external detector, according to the calculations in [30], is from 90 to:110010 of the total num-
ber of UC N stored in the bottle. This shows that the neutrons recorded by the external detector are former
UCN.
The physical nature of the observed heating of the UCN is still not clear. The most realistic cause of
the heating appears to be the presence of hydrogen on the bottle walls, although this contradicts the tempera-
ture independence of the storage time. This contradiction would be lessened somewhat if the assumption [31,
32] that hydrogen is strongly bound in the direction of the normal to a container surface but free to move
along the surface is confirmed. In this case when a neutron is inelastically scattered by a proton it could
gain an energy of the order of thermal, The cross section for this process v^-fT, where T is the wall tem-
perature, and therefore the experiments might show a weak dependence of the storage time on temperature.
The testing of the hydrogenhypothesis is ofgreat interest. If it is confirmed, the problem of a long storage time
can obviously be solved by using clean containers as proposed in [30]. Otherwise the problem of extended
containment of UCN in bottles remains open until the physical nature of the anomalous heating is explained.
In principle, the reduction of the storage time of UCN because of anomalous heating can be avoided by
using magnetic traps for storage [33]. A neutron introduced into a closed magnetic cavity in which the field
increases from the center to the periphery should be retained in it so long as its spin is directed along the
field E