SOVIET ATOMIC ENERGY VOL. 55, NO. 6

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• Publication Date: December 1, 1983
• Content Type: REPORT

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Russian Original Vol. 55, No. 6, December, 1983
June, 1984
SATEAZ 55(6) 797-904 (1983)
SOVIET
ATOMIC
ENERGY
ATOMHAR 3HEP('I1H
(ATOMNAYA ENERGIYA)
n
TRANSLATED FROM RUSSIAN
CONSULTANTS BUREAU, NEW YORK
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is a translation of Atomnaya Energiya, a
SOVIET I Soviet Atomic Energy
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ENERGY
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SOVIET ATOMIC ENERGY
A translation of Atomnaya Energiya
June, 1984
CONTENTS
Engl./Russ.
Escape of Radioactive Fission Products from Leaking Fuel Elements
into an Organic Coolant - E. I. Shkokov, V. V. Konyashov,
Yu. C. ;Simonov, Yu.,V. Chechetkin, A. D. Yurchenko,
. .
and E..K. Yakshin. . . . . . . . . . . . . ?
Thermal Desorption of implanted Helium from Austenitic Steels
of 16-15 Type - V. S. Karasev and V. G. Kovyrshin. . . . .'. .
Synergic Effect in Irradiating Graphite with H+ Ions and Electrons
- M. I. Guseva, S. M. Ivanov, and A. N. Mansurova. . . . .
Reprocessing and the Solidification of Nuclear Power Station
Wastes - A. S. Nikiforov, A. S. Polyakov,
and K. P. Zakharova. . . . . . . . . . . . . . . . . . .
Treating Radioactive Waters with a Mixed Ion-Exchange Bed
in a Continuous-Operation Plant - B. E. Ryabchikov,
E. I. Zakharov, A. P. Darienko, A. V. Rakhchev,
and M. Ch. Murabuldaev . . . . . . . . . . . . . . . . . . . .
Experimental Investigations of u-Atomic and u-Molecular Processes
in Hydrogen on the JINR Synchrocyclotron - V. P. Dzhelepov
and V. V. Fil'chenkov . . . . . . . . . . . .
Version of a Hybrid Reactor Based on Muon Catalysis of the D-T
Reaction - V. V. Orlov, G. E. Shatalov, and K. B. Sherstnev. .
Correction to the Readings of a Thermoelectric Thermometer
in Reactor Conditions - A. A. Greshilov, V. S. Terekhov,
V. I. Nalivaev, S. V. Priimak, and I. I. Fedik . . . . . . . .
Equivalent Doses of Different Types of Radiations
- B. G. Dubovskii. . . . . . . .
855
399
LETTERS TO THE EDITOR
Breeding Characteristics of a Fast Neutron Reactor in a Transient
Fuel Cycle - V. A. Chirkov . . . . . . . . . . . . . . . . .
859
402
Effect of Intermolecular Interaction in Gaseous Nitrogen
on the Scattering Cross Section of Cold Neutrons
- S. B. Stepanov, V. E. Zhitarev, A. M. Motorin,
and Yu. V. Sharanin. . . . . . . . . . . . . . . . . . . .
862
403
Composition of the Gaseous Phase and the Behavior of Xenon
and Krypton in Irradiated Fuel Elements of a BOR-60 Reactor
- A. P. Kirillovich, Yu. I. Pimonov, Yu. G. Lavrinovich,
and 0. S. Boiko. . . . . . . . . . . . . . . . . . . .
Basics of Pulsed Neutron Logging When Strong Absorbers Are Established
- D. K. Galimbekov, I. T. Ilamanova, B. E. Lukhminskii,
and A. I. Pshenichnyuk . . . . . . . . . . . . . . . . . . . .
Nuclear-Petrophysical Basics of Neutron Measurements in Rocks
-.A. I. Pshenichnyuk . . . . . . . . . . . . . . . . . . . . .
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CONTENTS
(continued)
Engl./Rues.
Optimization of Rectification and Exchange Pilot Plants for Isotope
Production - V. A. Kaminskii, G. A. Tevzadze, V. M. Vetsko,
0. A. Devdariani, and G. A. Sulaberidze . . . . . . : . . . . .
Angular Distributions of Fluxes of Charged Particles from a Thick
Target Bombarded with Beams of Protons, a Particles, and 12C
Nuclei with Energies of 3.65 GeV/Nucleon - V. E. Aleinikov
and G. N. Timoshenko . . . . . . . . . . .
Experimental and Computation-Theoretical Studies of the Development
of the Neutron Spectrum in Subcritical Assemblies with Heterogeneous
Reactor-Enriched Fuel - A. V. Bushuev, S. A. Bychkov, V. M. Duvanov,
A. Yu. Davydov, and V. I. Naumov . . . . . . . . . . . . . . . . . .
INDEX
Author Index, Volumes 54-55, 1983o . . . . . . . . . . . . . . . . .
Tables of Contents, Volumes 54-55, 1983o . . . . . . . . . . . . . .
The Russian press date (podpisano k pechati) of this issue was 11/25/1983.
Publication therefore did not occur prior to this date, but must be assumed
to have taken place reasonably soon thereafter.
87
410
878
412
881
414
887
893
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ESCAPE OF RADIOACTIVE FISSION PRODUCTS FROM LEAKING
FUEL ELEMENTS INTO AN ORGANIC COOLANT
E.
I.
Shkokov, V. V. Konyashov, Yu. G. Simonov,
Yu.
V.
Chechetkin, A. D. Yurchenko, and
E.
K.
Yakshin
Investigations and experience in the operation of reactors show that the physicochemical
properties of the coolants used (water, boiling water, sodium) considerably affect the escape
of radioactive fission products from leaking fuel elements. This applies also to-an organic
coolant, the radiation-thermal decomposition of which with the formation of deposits on the
cladding and under the cladding of leaking fuel elements, suppresses the escape of fission
products. At the same time, data about the escape of fission products are particularly im-
portant for reactors with organic coolants, as the inherent activity of an organic coolant is
very low (lower by 3-4 orders than the activity, for example, of the water of water-cooled/
water-moderated reactors (VVER) [1]). Therefore, for a correct calculation of the biological
shielding of the pipelines and plant of the primary circuit, it is necessary to take account
of the contribution of fission products to the coolant activity.
The coolant is ditolylmethane (DTM); its temperature is 410-450?K and its pressure is
0.5-0.8 MPa. Fuel elements with a length of 1100 mm and with an outside diameter of 9.8 mm
were tested. The fuel pellets, of sintered UO2 with a density of 10.2 g/cm3 and with a 235U
enrichment of 3%, had an outside diameter of 7.6 mm and an axial opening with a diameter of
1.4 mm.
The escape of fission products from-fuel elements with damaged cans was studied for a
fuel burnup of 1017_1.7.1020 fissions/cm (4-7000 MW?days/ton U), a fission intensity of 6.109-
9.1018 fissions/cm3?sec (specific power 0.18-28 W/g) and a maximum fuel temperature of 450-
1300?K .
The escape of fission products was monitored by the activity of the radionuclides in the
coolant. The value of the escape of the i-th radionuclide from the leaking fuel elements
(Ki), equal to the ratio of the number of atoms formed in the fuel element (Ri), was deter-
mined by the total activity of the radionuclide in the loop circuit (B1). Samples of the coolant
and gas from the volume compensator were taken periodically, and samples of the deposits
from the inner surfaces of the plant and loop were taken at the end of each experiment. The
activity of the radionuclides in the samples was measured on a y-spectrometer with a Ge(Li)-
detector; the relative statistical error of the measurements was up to 10%, and the total
error of the determination of the escape was about 30%.
In all the experiments, in the initial period of irradiation of the fuel elements, a
reduction of the escape of fission products with increase of the fuel burnup occurred (Fig. 1),
well known from experiments with oxide fuel [2-5] at a temperature below 1300?K. The escape
of gaseous fission products (GFP) depended on the intensity of the fissions in the fuel (Fig.
2).
The dependence of the escape of GFP from fuel or leaking fuel elements on the radio-
active decay Ai is assumed to be characterized by the power index n in the relation Ki ain.
In the range of fission intensities from 6.1010 to 9.1012 fissions/cm3?sec, the value of n
in the investigations carried out varied from 0 to 0.3. With a fission intensity in the
fuel of 6.1010 fissions/cm3?sec, the yield of GFP from the leaking elements was almost inde-
pendent of ai. Consequently, the migration time of the GFP beneath the cladding of the fuel
elements investigated was significantly less than the half-life of the short-lived radio-
Translated from Atomnaya Energiya, Vol. 55, No. 6, pp. 359-362, December, 1983. Original
article submitted February 7, 1983.
0038-531X/83/5506-0797$07.50 ? 1984 Plenum Publishing Corporation
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Fig. 1. Dependence of the escape of 133Xe
from leaking fuel elements on the fuel burn-
up (fission intensity 9.1012 fissions /CM3.
1
IN1
1NT
1,920P, fiuiono/cm3
sec) .
10'2
10~~0 .I I I'y 4 .,
sec-
Fig. 2. Escape of gaseous fission products
from leaking fuel elements versus the decay
constant (burnup 5.1017 fissions/cm3, fission
intensity, fissions /cm3-sect 0): 6.1010; x)
1.1012, ?) 9.1012; for 135Xe, the burnup in
the neutron flux is taken into account).
nuclide 138Xe (17 min). The escape of GFP also did not vary with increase of the opening in
the fuel element cladding from 0.01 to 1.2 mm2.
Before the coolant flows in beneath the cladding, the change of the GFP escape within
the measurement limits studied of the fission intensity, was due mainly to the nature and
parameters of the processes taking place in the oxide fuel. The escape of GFP, defined by
the value of n = 0 and obtained at a.low fission intensity in the fuel (6.1010 fissions/
cm3?sec), corresponds either to a recoil mechanism of the fission products from the fuel or
to a "self-expulsion" mechanism [6]. An estimate showed that satisfactory coincidence with
experiments is given by a calculation based on the recoil mechanism, in which it was assumed
that those fission fragments escape from the fuel which had completely lost their energy in
the gas void between the fuel and the cladding, and also in the gas space of open pores. The
calculated value'of the GFP escape from the fuel, found by the self-expulsion mechanism, is
a factor of 5 lower than the measured value.
With increase of the fission intensity in the fuel from 6.1010 to (1-9)?1012 fissions/
cm3?sec, the value of n was increased up to 0.2-0.3. The GFP escape, characterizing n =
0.2-0.3, is typical for uranium dioxide at a low temperature [2, 3, 5-7] and confirms the
significant contribution to the total escape of the GFP from the fuel at least of one further
mechanism of release, besides the recoil mechanism. Subtracting from the yields measured for
a fission intensity of (1-9).1012 fissions/cm'-sec the yields measured for 6.1010 fissions/
cm3?sec (contribution of the recoil mechanism), the following dependence of the escape of
GFP on the radioactive decay and fission intensity f is obtained:
Kt,. k'
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Fig. 3. Variation of the escape from a
leaking fuel element versus time, in the
case of the formation of deposits on its
cladding.
toe
'S f0? ulI ~
X41 ios
0 2 4 8 8 10 12 14 18 18 202224 28ffi3012 14 38 38 40
Time, days
Fig. 4. Change of bulk activity of the
fission products in the coolant during
irradiation of a leaking.fuel element
(Nfuel is the power of the fuel element
during irradiation).
It is characteristic for the mechanism-- of radiation-stimulated diffusion, well known from
[8-10]. The results of the calculation of the GFP escape from the fuel according to this
mechanism, coincide with the experimental data obtained with an error of +30%. The radiation-
stimulated diffusion coefficient is taken from [10].
We note that the calculation by the thermal diffusion mechanism gives underestimated
values by comparison with experiment. According to the data of [8, 11], at a temperature of
the oxide fuel below 1300?K, radiation-stimulated diffusion predominates over thermal diffu-
sion.
The escape of iodine radioisotopes from the fuel elements investigated into the organic
coolant was a factor of 5 less than the escape of GFP at the start of irradiation, and a
factor of 300 less than for a fuel burnup of (0.5-1.7)?1020 fissions/cm'-sec. The escape
of iodine radioisotopes, less by a factor of 10 by comparison with the escape of GFP, is
obtained also in boiling water reactors; in this case, the marked difference is explained by
the sorption of iodine on the inside surface of the fuel-element claddings [12-14]. It can
be postulated that the same mechanism is valid also for the leaking fuel elements with alu-
minum cladding operating in the organic coolant. This supposition is based on the results
of investigations of the sorption of iodine on the surface of metals from a gaseous medium
In the operating conditions of the fuel elements in the organic coolant, the formation
on the claddings of deposits of radiation-thermal cracking products (RTCP) of the coolant is
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characteristic [1]. An increase of the RTCP of DTM, taking place during irradiation of the
fuel element and in. regions of; damage, leads to a reduction of the escape of GFP from the
fuel. element (Fig. 3). The rate of reduction of the escape of radionuclides is approximately
proportional to the radioactive decay constant. It is probable that this is due to an in-
crease of the migration time of the radionuclides :before their escape into the coolant, in
particular by diffusion through the layer of deposits. The rate of reduction of the escape
of iodine radionuclides was higher than'for the GFP, which may be caused by the difference
in the diffusion coefficients of the radionuclides in the RTCP. At the end of irradiation
of the fuel elements, the yield of 131I from them,.amounted in all to 2.10-'.
During material-behavior investigations, radiation-thermal cracking products of DTM were
detected not only outside but also beneath the cladding of leaking fuel elements. This is
the consequence of the inflow of DTM under the cladding which probably occurs after shut-
down of the reactor because of a reduction of the gas pressure under the cladding. Resump-
tion of operation of the reactor leads to decomposition of the DTM which has flowed in, and
to the formation of cokelike RTCP. The effect of these processes on the release of GFP is
illustrated in Fig. 4. After the first reactor shutdown, the escape of GFP decreased by a
factor of 5-7, with conservation of the escape of iodine radioisotopes. The phenomenon men-
tioned can be explained by the fact that the cokelike RTCP blocked the outflow of fission
products from the major part of the fuel element. This led to a significant reduction of
the escape of GFP but, at the same time, the escape of iodine radioisotopes was unchanged as,
because of sorption on the inside surface of the fuel-element cladding, they entered the
coolant only from the region,of the fuel element close to the defect. After the second
reactor shutdown (see Fig. 4), the escape of iodine radioisotopes was also reduced (scaled
to nominal reactor power), which confirms the coking-up of the inside spaces of the fuel
element in the vicinity of the defect.
In conclusion, it should be noted that the escape of fission products from defective
oxide fuel elements into an organic coolant is significantly lower than the escape of fission
products from leaking fuel elements in water-cooled/water-moderated reactors. According to
the data of [13], with a linear loading of 11 kW/m, the escape of 133Xe from leaking fuel
elements of light-water reactors is equal to approximately 10-2. This is almost a factor of
10 higher than the escape of 133Xe from the fuel elements investigated into the organic
coolant with the same linear loading and with a burnup of more than 1019 fissions/cm3. The
reason for this difference, as proposed in [13, 16, 17], lies in the increase of the diffusion
coefficients of the fission products in oxide fuel by the action of water and steam pene-
trating under the cladding of the defective fuel element and increasing the stoichiometric
ratio in the uranium dioxide.
LITERATURE CITED
1. V. A. Tsykanov, et al., At. Energ., 50, No. 6, 376 (1982).
2. R. Carroll.et al., Nucl. Sci. Eng., 38, 143 (1969).
3. J. Findlay, J. Nucl. Mater., 35, 24 (1970).
4. P. Chenebault and R. Delmas, in: Behavior and Chemical State of Irradiated Ceramic
Fuels, IAEA, Vienna, p. 337 (1974).
5. K. Shiba, et al., J. Nucl. Mater., 48, 253 (1973).
6. S. Yamagishi and T. Tanifuji, J. Nucl. Mater., 59, 243 (1976).
7. R. Soulhier, Nucl. Appl., 2, 138 (1966).
8., A. Hoh and Hj.. Matzke, J. Nucl. Mater., 48,.157 (1973).
9. 0. Gautsch, J. Nucl. Mater., 35, 109 (1970).
10. N. Beatham, J. Nucl. Mater., 98, No. 2 (1981).
11. D. M. Skorov, Yu. F. Bychkov, and A. I. Dashkovskii, Reactor Material Behavior [in
Russian], Atomizdat, Moscow (1979), p. 80.
12. G. Eigewilling.and R. Hock, Trans. Am. Nucl. Soc., 23, 258 (1976).
13. E. Shuster et al., Nucl. Eng. Design, 64, 81 (1981).
14. A. V. Vasilishchuk et al., Preprint Scientific Research Institute of Nuclear Reactors,
NIIAR-1 (26) [in Russian], Dimitrovgrad (1982).
15. M. Osborne, R. Briggs, and R. Wicher, Trans. Am. Nucl. Soc., 38, 463 (1981).
16. D. Cubieciotti, Nucl. Technol., 53, No. 1, 5 (1981).
17: -B. Lastman, Radiation Phenomena in Uranium Dioxide [in Russian], Atomizdat, Moscow
(1964).
800
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THERMAL DESORPTION OF IMPLANTED HELIUM FROM
AUSTENITIC STEELS OF 16-15 TYPE
The accumulation of helium when a material is irradiated to a high fluence with fast
neutrons and a particles has a substantial effect on embrittlement, swelling, the formation
of surface flaws, and other processes that restrict the working life of the material in the
first wall of a fusion reactor [1].
Austenitic stainless steels are considered [1] as the material for the first wall.
The release of implanted helium has been examined for relatively low ion energies (up to
some tens of keV) [2-5] and low concentrations (10-6-10-2 at.%). However, the theoretical
spectrum of a fusion reactor contains a high-energy helium-ion component (>3.5 MeV) [1]. One
needs data on the thermal desorption of helium at high temperatures after irradiation with
ions of energy about 1 MeV, and also on the correlation of these processes with the change
in properties in the material at high concentrations of implanted helium (up to 50 at.%).
Specimens of the austenitic stainless steels OKh16N15M3B and OKh16N15M3BR, which are
similar in composition, were irradiated with the ESU-2 electrostatic accelerator at the
Kharkov Technical Physics Institute of the Academy of Sciences of the Ukrainian SSR using
helium ions of energy 0.5-1.6 MeV at doses in the range 5.1016-1.6 x l018.ion/cm2 at about
50?C. Descriptions have previously been given [6-8] of the methods of irradiating the speci-
mens and those used to examine the helium release and the surface changes during heating at
a constant rate to 1300?C.
Effects of Radiation Dose. Two series of experiments were performed on irradiating
OKh16N15M3B steel [6] in the dose range 5.1016-1.2 x 1018 ion/cm2 (energy of helium ions
0.8 MeV) and the same for OKh16N15M3BR in the range 5.1016-1.6 x 1018 ion/cm2 (helium-ion
energies 0.5 and 0.8 MeV). The thermal desorption spectra for the OK h16N15M3BR specimens
after irradiation at 0.5 and 0.8 MeV were identical and differed only in that the peaks were
displaced by 20-30?C. The specimens were irradiated with normal incident monoenergetic beams.
Figures 1 and 2 show the temperature dependence of the relative release rate for three
doses for each steel together with the background curves obtained under equivalent condi-
tions without the specimens. I n all cases, the spectra are for a heating rate of 0.25?C/sec.
The general trends are the same for the two steels. As the dose increases, the number
of peaks rises from one to five, and the strengths of the peaks increase together with the
amount of..helium released in the first (low-temperature) peaks, while the temperature at
which the release begins is lower, as is the formation of visible surface damage. The pres-
ence of several peaks indicates that there are various stages, each of which is associated
with a particular migration mechanism. The values of the activation energy Ea (eV) as deter-
mined from the shifts in the peak temperatures Tmax [9] with heating rates of about 0.1 to
0.25?C/sec are given in Figs. 1 and 2 at the tops of the corresponding peaks. At this heating
rate, all the observable stages in the annealing were satisfactorily observed from the dis-
placement of T , which was sufficient to give E . Some of the values of E have been re-
max a a
vised in relation to earlier ones [6, 7]. The desorption characteristics and the surface-
structure changes provide some conclusions on the mechanism for each stage.
The first stage, which is characteristic by the least release, is observed only after
irradiation to a dose of about 1018 ion/cm2. Here the activation energy is 0.4 + 0.1 eV,
which indicates the release of helium from surface layers. On heating, a transition structure
is formed: swellings of irregular shape of size 100-300 pm (Fig. 3a); in the range 300-800?C
in the first stage, these crack open (characteristic brief high-intensity pulses on curves
I in Figs. 1 and 2). The resulting increase in the free surface favors the release of helium.
Translated from Atomnaya Energiya, Vol. 55, No. 6, pp. 362-366, December, 1983. Original
article submitted August 30, 1982.
0038-531X/83/5506- 0801$07.50 ? 1984 Plenum Publishing Corporation
801
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'x J00 $00 700 900 1100 9300
..Annealing tenitierature, ?C
Fig. 1. Kinetic curves-for the release of
helium from OKh16N15M3B steel specimens
irradiated by helium ions of energy 0.8
MeV: I) dose 1.2 x 1018 ion/cm'; II) 4 x
1017 ion/ cm'; III) 1 x 1017 ion/ cm'; IV)
background curve.
20 ~ I I i i . / ' f
I 'F
T 10 1 ~ ~ ~ ~ I ~ ~ I I ~ I ~
S00 700 900 1100
Annealing temperature, t
Fig. 2. Kinetic curves for the release of
helium from OKh16N15M3BR steel specimens
irradiated with helium ions of energy 0.5
MeV: I) dose 1.6 x 1018 ion/cm's II) 5 x
1017 ion/cm'; III) 5 x 1016 ion/cm'; IV)
background curve.
Also, the desorption from the surface of OKh16N15M3B steel after reactor irradiation under
conditions producing saturation with helium from the medium also occurs in the first stage
[10].
The value Ea = (1.5 + 0.2) eV for the next stage somewhat exceeds the activation energy
for vacancy migration in austenitic steels (E = 1.3 eV [11]) and may include the energy of
the binding of'helium to vacancies, as well as the free-vacancy formation energy in the de-
composition of point-defect clumps. No changes were observed in the surface structure in
the temperature range.for this stage.
The stage with Ea = (4 + 0.4) eV controls a certain proportion of the released helium
dependent on the type of steel and the dose. Desorption with an activation energy of about
4 eV is associated [12] with the formation,. migration, coalescence, and emergence of helium
bubbles at sinks, particularly at grain boundaries. The evidence on the surface-morphology
change agrees with this. Amounts of about 1017 ion/ cm' are insufficient to produce gas
cavities of substantial size. Figure 1 shows that the desorption spectrum for OKh16N15M3B
steel (dose 5.1016-1017 ion/ cm') is characterized by the release of helium only in one peak
with E a = 4 eV. On average, the grain size of the irradiated specimens remains at the
initial level after heating to 1300?C (Figs. 3b and c, in the first of which the boundary
between the irradiated and initial zones is seen). This indicates that helium emerges at
the grain boundaries during heating, fixes them, and retards recrystallization [13], whereas
in the unirradiated zone and on the outer side of the specimen (thickness 0.2 mm) there are
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Fig. 3. Microstructures of the surfaces
of irradiated specimens of OKh16N15M3B
steel (a-c) and OKh16N15M3BR steel (d)
after heating to 600?C (a) and 1300?C
(b-d): a) dose 9 x 1017 ion/cm2, x300;
b) 5 x 1016 ion/cm2, x450; c) 4 x 1011
ion/cm2, x450; d) 5 x 1017 ion/cm2, x450.
appreciable increases in grain size. We examined specimens that had received doses of 5?
1036-1017 ion/cm2 after annealing and additional shallow etching, and this showed more
clearly than does Fig. 3b that there are chains of pores of size 0.5-1 pm lying mainly at
the grain boundaries but partly in the bodies.
At doses of over 2 x 1017 ion/cm2, blistering occurs: a stage where bubbles arise at
the surface (Fig. 3c). In the light of current concepts on the blistering mechanism [14],
this is due to the fusion of gas bubbles and the growth of the blisters to critical sizes in
the bulk of the grains. Helium is released from the blisters in a stage with Ea = (4.5 +
0.5) eV, and also along the channels in the porous structure (pore diameter 0.5-1.5 um, Fig.
3d) in the last stage with Ea = (5 + 0.5) eV. The energy. parameters for the last three
stages are fairly close (4-5 eV) and characterize the formation and emergence of large helium
accumulations.
Each of these stages involves a shift in the peak as. the heating rate or dose is varied,
which indicates thermal activation for the processes controlling the release. The charac-
teristic changes in surface structure (Fig. 3b-d) and the high values of the activation energy
show that the release is associated with microplastic deformation and failure. Such pro-
cesses are also thermally activated [15] and are characterized by the corresponding values
for the current activation energy.
The spectra show that the peaks shift to lower temperatures as the dose-increases with-
out change in the energy characteristics in a given stage. This applies for all peaks in
both steels (Fig. 4). The dependence of Tmax on the logarithm of the dose is linear. There
is an appreciable difference in slope between curves 1 (1.5 eV) and 2-4 (4-5 eV), which may
be associated with different migration mechanisms (single atoms of groups).
803
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1300
1200
1100
1000
900
800
;' 700
F-! 1200
1100
NO
800
800
700
800
10? 10,9
Dose, ion/cm2
Fig. 4
ol
,900 1000 1100 1200
Annealing temperature, t
Fig. 5
Fig. 4. Dependence of Tmax on dose for specimens of OKhl6N15M3B
steel at an ion energy of 0.8 MeV (a) and for OKh16N15M3BR steel
at 0.5 MeV (b) and activation energies in eV of 5 (1); 4.5 (2);
4 (3); and 1.5 (4).
Fig. 5. Comparative desorption spectra for specimens of
OKh16N15M3B steel irradiated to doses of 5 x 1017 ion/cm2 at the
following energies in MeV: 0.5 (1); 1 (2); 1.6 (3); and peak
depths of 1.3, 2.1, and 3 um correspondingly [16].
Effects of Ion Energy. As the energy increases, the surface damage changes substantially
in character [16]: The blister diameter increases, while the density of them decreases, and
the diameter distribution broadens. Results are available for ion energies of 0.5, 1, and
1.6 MeV as regards the effects on the peak positions on heating (dose 5 x 1017 ion/ CM2) for
OKh16N15M3B specimens, which show that the temperature for the start of release increases
with the energy (from 650 to 850?C), and also Fig. 5 shows that the peak temperature relates
to the stage in which most of the helium is released. The other stages are now shown in Fig.
5, as substantially smaller proportions are released in these.
A similar shift in peak temperature with energy occurred in the release of helium from
tungsten [17]. It has been determined analytically [18] that the peak positions are dependent
on the radiation-defect concentration, which affects the migration conditions. According to
these concepts, the shift in peak temperature is related to change in the depth of the helium
at a constant activation energy (the effects of the energy are shown in Fig. 5), with further
effects from the radiation-defect concentration (Fig. 4 shows the effect of dose), which
agrees with our results.
Effects of Implantation Conditions. After irradiation to a given maximum concentration,
the nature of the surface damage on subsequent heating is dependent on the conditions used
(normal monoenergetic beam or a depth-scanning one) [8]. The implantation profiles are
broadened and are similar to those expected for high-energy helium ions in a fusion reactor.
The spectra were determined after irradiation with a scanning beam in which the angle of in-
cidence of the ions varied from 0 to 74?. The maximum dose at the center of the target was
2.8 x 1018 ion/cm2 and fell by about a factor 10 at a distance of 0.8 cm. The helium implan-
tation profile varied correspondingly, as determined by Rutherford proton backscattering
(Fig. 6a). There are reductions in the peak intensity (Fig. 6b) and in the area under the
desorption curve (proportion of the least helium). Throughout the irradiation zone there
was a single type of surface damage: pores of size 0.5-1.5 um, while the desorption spec-
trum showed a single stage with Ea = 5 eV.
Figure 6b shows the shift in Tmax produced by the implantation conditions, which is due
to effects from factors acting in opposite senses. One the one hand, the implantation depth
alters by about 0.6 um from curve 3 to curve 1 (Fig. 6a), which should cause the peak to
shift by about 40?C to higher temperatures (on change in the ion energy from 0.5 to 1.6 MeV,
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Helium implantation depth, pm
1100 1000 1200
Annealing temperature, ?C
Fig. 6. Implantation profiles (a)
and desorption spectra (b) for speci-
mens of OKh16N15M3B steel irradiated
by a depth-scanning beam of energy 1
MeV: 1) central zone;.2) at 0.4 cm
from center; 3) at 0.8 cm from center;
4) background curve.
which alters the implantation depth by about 1.6 um [16], the peak in Fig. 5 shifts by almost
100?C). On the other hand, an increase in the overall dose by a factor 10-15 has the reverse
effect of shifting the peak by 150-200?C (Fig. 4a). The resultant shift from curve 3 to
curve 1 (Fig. 6b) agrees with these estimates and shows that the dose (i.e., the radiation-
defect concentration) has a substantial effect on the release during the formation of the
porous structure.
Composition Effects. Although there are similarities between the desorption-kinetic
curves for the two steels, there are certain differences, which are due to the presence of
0.005 at.% boron in OKh16N15M3BR steel designed to improve the physicomechanical properties,
in particular the plasticity [19]. In this material, the main release process is the one with
E = 5 eV at 5 x 1016 ion/cm2 (Fig. 2, curve III), which is accompanied by the formation of a
porous structure at the surface after annealing. The peaks with Ea = 5 eV also have the
highest intensity for OKh16N15M3BR steel at higher doses, and the surface damage is charac-
terized by a porous structure (Fig. 3d). The desorption and surface topography change for
these steels under identical conditions of irradiation and annealing shows, [7] that helium is
retained in OKh16N15M3BR steel at higher temperatures than in OKh16N15M3B. Therefore, the
release of helium through channels in the porous structure makes a.contribution in OKh16N15M3BR
steel. This may be due to the presence of boron compounds at the grain boundaries that hinder
the emergence of helium bubbles, and therefore the release is reduced in the stage with ac-
tivation energy 4 eV.
Conclusions. A study has been made of the effects of irradiation conditions on the ther-
mal desorption of implanted helium from austenitic steels of 16-15 type. The energy parameters
of the various release stages have been determined. There is a correlation between the desorp-
tion and the growth of the surface damage on heating. The peak positions and the temperature
ranges for the corresponding stages are substantially dependent on the.dose and ion energy.
The results provide further insight into the behavior of helium in materials for the first wall
of a fusion reactor under nonstationary conditions. The differences in release kinetics for
two steels similar in composition indicate that it may be possible to make a suitable material
by varying the additives.
We are indebted to G. D. Tolstolutskaya for providing the irradiated specimens and the
data on the implantation profiles, and also for participating in the discussion of the results..
LITERATURE CITED
1. V. V. Orlov and I. V. Al'tovskii, Nuclear Science and Engineering, Series Physics of
Radiation Damage and Radiation Materials Science [in Russian], Issue 1(15) (1981), p. 9.
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 Declassified and Approved For Release 2013/02/21: CIA-RDP10-02196R000300030004-5
2. D. Whitmel and R. Nelson, Rad. Effects, 14, 249 (1972).
3. W. Bauer and W. Wilson, "Radiation-induced voids in metals," AEC Symposium Series
CONF-710601, Albany (1971), p. 230.
4. D. M. Skorov et al., Nuclear Science and Engineering, Series Physics of Radiation
Damage and Radiation Materials Science [in Russian], Issue 1(1) (1974), p. 58.
5. D. M. Skorov et al., ibid., Issue 1(6) (.1978), p. 46.
6. V. S. Karasev et al., ibid., Issue 2(13) (1980), p. 82.
7. V. S. Karasev et al., ibid., Issue 1(15) (1981), p. 52.
8. V. V. Gann et al., At. Energ., 48, No. 4, 266 (1980).
9. V. S. Karasev et al., ibid., 34, No. 4, 251 (1973).
10. V. S. Karasev et al., Nuclear Science and Engineering, Series Physics of Radiation
Damage and Radiation Materials Science [in Russian], Issue 2(10) (1979), p. 48.
11. S. Harkness and Che-Yu-Li, in: Radiation Damage in Reactor Materials, Proc. Symposium
IAEA, Vol. 2, Vienna (1969), p. 289.
12. D. Reed et al., Vacuum, 24, No. 4, 179 (1974).
13. A. S. Nikiforov et al., At. Energ., 53, No. 1, 3 (1982).
14. M. I. Guseva, Yu. V. Martynenko, and V. F. Rybalko, Nuclear Science and Engineering,
Series Physics of Radiation Damage and Radiation Materials Science [in Russian],
Issue 4(18) (1981), p. 35.
15. F. Garofalo, Creep and Long-Term Strength Laws for Metals [Russian translation],
Metallurgiya, Moscow (1968).
16. G. D. Tolstolutskaya et al., Fiz. Khim. Obrab. Mater., No. 6, 29 (1979).
17. J. Pierre and D. Paulmier, C. R., 280, No. 2, B275 (1975).
18. S. Donnely and D. Ingram, Vacuum, 28, No. 2, 269 (1978).
19. N. P. Agapova et al., in: Peaceful Uses of Atomic Energy, Vol. 10, IAEA (1972), p. 3.
SYNERGIC EFFECT IN IRRADIATING GRAPHITE WITH H+
IONS AND ELECTRONS
M. I. Guseva, S. M. Ivanov, and A. N. Mansurova UDC 621.039.531:621.039.532.21
In some designs for fusion reactors, it is envisaged that there will be diaphragms
made of graphite and sometimes also screens that are subject to radiation loads (with the
exception of neutrons) and thermal loads. In particular, the diaphragm and screen will be
irradiated by hydrogen ions and electrons. The screen material should be characterized
by a small value of SZ2 (S is the sputtering coefficient and Z is atomic number). The
maximum sputtering coefficient for graphite produced by hydrogen ions is determined by the
interaction that produces hydrocarbons at 300-700?C [1-3]. When the graphite is acted on
simultaneously by atomic hydrogen from the gas phase and electrons of energy 100-600 eV,
the chemical sputtering coefficient is much higher than that in the absence of electrons
[4].
Here we examine the chemical sputtering of sintered carbon and MPG-8 graphite on
parallel irradiation by H+ ions of energy 10 keV and electrons or by hydrogen ions alone.
The irradiation was performed on the ILU-2 accelerator [5] in an irradiation device,
in which the main elements were a planar graphite straight-channel oven with a system of
screens and an electron gun with an indirectly heated cathode made of lanthanum hexaboride.
The targets in the form of wafers of size 20 x 10 x 1 mm were mounted in the oven in special
holders. The angles of incidence of the electrons and the H+ ions on the target were
correspondingly 30 and 0?. The MPG-8 graphite specimens were irradiated simultaneously by
hydrogen ions and electrons at 100?C (Ee = 50-1500 eV) and 470?C (Ee = 400 eV), or by
hydrogen ions alone at temperatures from 100 to 1600?C. The carbon-sinter specimens were
Translated from Atomnaya Energiya, Vol. 55, No. 6, pp. 366-368, December, 1983. Original
article submitted December 14, 1982.
806 0038-531X/83/5506- 0806$07.50 C)1984 Plenum Publishing Corporation
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 Declassified and Approved For Release 2013/02/21: CIA-RDP10-02196R000300030004-5
m 10~
8
l0'~ I I i I h
0 (?00 800 1200 1600 apt
Fig. 1 Fig. 2
Fig. 1. Temperature dependence of the sputter-
ing coefficient for MPG-8 graphite by H+ ions
(0) at E = 10 keV and H+ + e(x) at EH+ = 10
keV, Ee = 400 eV.
Fig. 2. Dependence of the sputtering coeffi-
cient on electron energy for MPG-8 graphite
to H+ ions of energy 10 keV and also electrons
(Tirr = 100?C).
irradiated simultaneously by H+ ions and electrons at temperatures from 100 to 800?C (Ee =
400 eV) and by hydrogen ions alone at 100-1400?C. Before and after irradiation, the speci-
mens were annealed at 1100?C for 1 h to outgas them. This heat treatment was performed for
specimens irradiated at temperatures below 1100?C; for Tirr > 1100?C the specimens received
only the preirradiation annealing at Tann Tirr' The target temperatures during irradia=
tion and annealing were monitored with a tungsten-rhenium thermocouple. In most of the ex-
periments, the H+ dose was 3.6 x 1019 ion/cm2, current density about 300 pA/cm2, and ratio of
the It flux density to that of the electrons 1:20. The residual-gas pressure in the chamber
was about 10-4 Pa, while the working pressure was about 10-3 Pa. Nitrogen traps were used to
freeze out the oil vapor. The sputtering coefficient S was determined by weighing.
Figure 1 shows how the temperature affects. the sputtering coefficient for MPG-8 graphite
by H+ ions. When the temperature is raised outside the limits stated in [1-31, there is a new
sharp rise in the sputtering coefficient. To avoid any systematic error, check experiments
were performed with the target heated to 1600?C in the absence of irradiation but at the same
partial pressure of hydrogen in the chamber. The mass did not alter, which means that the
mass reduction produced by H+ bombardment is associated with sputtering.
Figure 1 shows that at 1600?C the sputtering coefficient is approximately two times
Schem in the region of the maximum for molecule formation at about 500?C. When the composi-
tion of the sputtered particles were examined, hydrocarbons were not detected, which indicates
that there is no chemical sputtering at Tirr > 800?C. Further research is needed to inter-
pret the accelerated sputtering of graphite at high temperatures.
Simultaneous bombardment by H+ and electrons increases the sputtering coefficient by
more than an order of magnitude even at 100?C, i.e., where there is no chemical sputtering
in the absence of electron bombardment. In checks with the graphite bombarded by electrons,
it was found that there was no change in the target mass, which means that a combination of
H+ and electrons clearly produces a synergic effect, whereas each form of bombardment
separately does; not produce chemical sputtering at this temperature.
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0 900 800 9200
r ?c
Fig. 3. Temperature dependence
of the sputtering coefficient
for USB-15 sintered carbon pro-
duced by H+ ions (0) at E = 10
keV and by H+ + e(x) or EH+ _
10 keV, Ee = 400. eV.
The electron energies in our experiments were varied from 50 to 1500 eV. The sputtering
coefficient varied most in the range 200-300 eV. Figure 2 shows that the electron energy
has an appreciable effect on the sputtering rate produced by the hydrogen ions. The elec-
tron bombardment breaks the bonds between the carbon atoms and thus increases the probability
that CH4 , molecules will be formed. Electrons of energy about 1500 'eV penetrate a consider-
able distance into the graphite, while hydrocarbon molecules are produced by the H+ only
near the surface, so high-energy electrons have less effect on the sputtering coefficient.
Also, the ionization cross section in electron impact has a maximum at about 300reV. The
synergic effect with MPG-8 graphite occurs also at.about 500?C, which corresponds to the
peak chemical sputtering by hydrogen ions. Parallel bombardment of MPG-8 graphite by hydro-
gen and electrons (Ee = 400 eV) increases Schem by almost a factor 2.5, virtually up to the
maximum possible value Smax = 0.25, which is realized if all the hydrogen atoms striking the
target form escaping methane molecules.
Figure 3 shows how the temperature affects the sputtering coefficient for sintered car-
bon bombarded by Hl" and electrons together and for comparison by H& alone. There is a
difference from fine-grained MPG-8 graphite in that at 100 and 200?C there is no synergic
effect within the errors of experiment. There is no chemical sputtering of sintered carbon
at temperatures up to 200?C, which indicates the exceptional chemical inertness of this
variety of graphitized material. On the other hand, the synergic effect is prominent at
Tirr = 400-800?C; at Tirr = 460?C, HF and electrons together increase the sputtering coeffi-
cient by about.a factor of five. In that case, the chemical sputtering coefficients become
comparable with those from MPG-8 graphite exposed to hydrogen ions alone. Therefore, the
sputtering rate of sintered, carbon in parallel irradiation at temperatures up to 200?C gives
a value about 25 times less than that for ordinary fine-grained graphite, while the figure
for Tirr 500?C is about a factor of two.
Sintered carbon at Tirr > 1000?C (Fig. 3) resembles MPG-8 graphite in showing an in-
crease in the sputtering coefficient with temperature. However, throughout the range, the
value for the sintered carbon is much less than that for MPG-8 graphite.
These results show that it is necessary to consider synergic effects in choosing
materials for the first wall of a fusion reactor. In particular, they show that it is
hardly desirable to use graphite in a fusion reactor in devices that accumulate heat on
account of thermal capacity, i.e.,-that operate by radiative cooling. On the other hand,
our results show that sintered carbon can be used as the diaphragm if external cooling is
provided, as-in the PLT. Martynenko in 1977 [6) predicted a synergic effect from the
combination of ions and electrons on the basis of concepts on the breakage of bonds for
the surface atoms and thus-reduction in the binding energy.
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 Declassified and Approved For Release 2013/02/21: CIA-RDP10-02196R000300030004-5
1. N. P. Busharov et al., Fiz. Plazmy, 2, No. 4, 588 (1976).
2. M. I. Guseva, E. S. Ionova, and Yu. V. Martynenko, Preprint IAE-3225, Moscow (1979).
3. M. I. Guseva, et al., "Chemical sputtering of graphite by H+ ions," Paper at the
Soviet-American Seminar on Surface Effects in Fusion Equipment [in Russian], Moscow
(1975).
4. C. Ashby and R. Rye, J. Nucl. Mater., 92, 141 (1980).
5. V. M. Gusev et al., Prib. Tekh. Eksp., No. 4, 19 (1969).
6. Yu. V. Martynenko, Preprint IAE-2815, Moscow (1977).
REPROCESSING AND THE SOLIDIFICATION OF NUCLEAR
POWER STATION WASTES
A. S. Nikiforov, A. S. Polyakov, and
K. P. Zakharova UDC 621.039.7
Nuclear power is making an increasing contribution to the production of electrical
energy throughout the world [1] (Table 1), which makes more acute the question of power
station safety, particularly a reliable system of rendering all forms of radioactive waste
harmless. Power station operation leads to the accumulation of spent fuel, which contains
99.9% of the radionuclides formed by fission. The spent rods are held temporarily for 3-10
years at the stations in special basins and are then transported for reprocessing.
About 0.1% of the radionuclides formed in the reactor will pass to special equipment
for processing before storage in the form of liquid, solid, and gaseous wastes. The safe
disposal of liquid radioactive wastes LRW involves treating effluents to acceptable standards
to enable them to be reused or discarded into open waterways, with the accumulation of the
pollutants in a minimal volume isolated from the environment.
Two types of reactors form the basis of nuclear power in the USSR: the VVER and RBMK.
There are some composition differences between the LRW from stations with VVER and RBMK
reactors, but the processing schemes are essentially identical. All the homogeneous wastes
are collected in receiving vessels, where they are mixed and averaged. The mixture is
separated from the coarse suspensate and the pH is adjusted to about 10, after which the
mixture is sent for evaporation, which is the main operation in purifying and concentrating
LRW.
The evaporation is performed in two stages. The.crude residue (concentrate) will con-
tain 200-300 g/liter of salts, and this along with the slimes from spent filter materials
(ion-exchange resins, filter pearlite containing precipitate, and activated charcoal) is sent
to liquid-waste stores (concrete vessels of volume 400-3000 m3 lined with stainless steel and
designed to work for 20 years). After the lapse of 20 years, or in some cases earlier, the'
filled vessel must be emptied. The method of storing radioactive solutions and slimes in
such vessels is temporary, since there is a potential hazard of leakage (due to corrosion of
the materials) and radioactive contamination of the environment.
Major problems in LRW reprocessing are volume reduction and the salinity of the resulting
wastes. Here improvements can come from advances in water-treatment technology and methods
of decontaminating the loops, equipment, and buildings, together with the provision of sealed
equipment during station operation and servicing.
Considerable importance attaches to differentiated collection of effluents differing in
composition and radioactivity, which means that some of the water can be processed in the
Translated from Atomnaya Energiya, Vol. 55, No. 6, pp. 368-372, December, 1983. Original
article submitted June 23, 1983. -
0038-531X/83/5506-0809$07.50 ? 1984 Plenum Publishing Corporation
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 Declassified and Approved For Release 2013/02/21: CIA-RDP10-02196R000300030004-5
TABLE 1. Increase in the Overall Installed
Power of All Stations and Nuclear Power
Stations in the World in GW(el)
1978
1830
108 (5,8)
1979
1900
122 (8,4)
1980
2030
148(7 3)
1
1985
2450-2850
)
)
290-350 (,
2000
5230-8200
1030-1650 (20-27)
*In parentheses is the proportion of nuclear stations, %
most economical way or, if possible, discarded directly to external bodies.
Shower waters contain surfactants and have very low levels of radioactivity, and these
should probably be collected separately from laundry waters and not processed if the level of
radioactivity allows them to be discarded to the ordinary drainage. The scope for discarding
various groups of laundry waters without processing will be determined not only by the radio-
element contents but also by the concentrations of resistant surfactants type OP-7, SF-3K,
and so on. At present, researches are in .hand on treating laundry effluents by ultrafiltra-
tion, foam flotation, radiation-chemical destruction, etc. It is necessary to improve laun-
dering processes to reduce the volumes of effluent.and the contents of surfactants and salts
in them. The progressive and more economical approach is to collect low-salt unorganized
flows in special vessels and process them separately from salt solutions by ion-exchange
treatment.
Filter materials (pearlites, ion-exchange resins, and activated charcoal) together with
LRW concentrates will also be solidified for subsequent storage. The amount of spent pearl-
ites sent every year to liquid-waste stores (per GW of installed electrical power) is 25-35
tons. The pearlites along with ion-exchange.resins are transported hydraulically to the
stores, while the decantation water along with trap water passes to reprocessing. The filter
material sludges in the store consolidate, which complicates extracting them from the vessels
for subsequent transfer to solidification.
One way of improving mechanical filtration of LRW in order to reduce the amount of pearl-
ite in power station wastes is to use cermet filter elements that are regenerated by reverse
flow, which are very widely used in various branches of industry in the. USSR and are also
employed at foreign nuclear power stations.
The volume of LWR can also be reduced by optimizing and automating the control systems
for various processes, particularly coolant treatment, loop decontamination, and the cleaning
of buildings and equipment, the regeneration of ion-exchange materials, and the processing of
the resulting wastes, etc. A substantial reduction in the volume of wastes can be attained
by improving methods of monitoring LRW and gaseous discharges, as well as sampling systems,
together with automatic instrumental operation of analytical monitoring at power stations.
In the operation of a nuclear power station, there is the problem of eliminating ammonia.
The usual alkali mode of evaporation for trap water leads to the distillation of the ammonia
from the cubed residue and the accumulation of it in the ion-exchange filter, whence it
passes with the regenerated material to evaporation again, i.e., there is no quantitative
removal of ammonia in existing schemes for handling LRW. This leads to it accumulating in
the system, which reduces the exchange capacity of the ion-exchange materials, shortens the
filter cycle in the condensate treatment plant, and increases the amount of regeneration
solution. To localize the ammonia one can best distill it, which utilizes its volatility in
an alkaline medium and also the rectification effect.
It is also necessary to improve methods of removing traces of oil from effluents and
reworking them. Removing oil from the distillate prevents the oil from entering the reactor
loop and contaminating the ion-exchange filters, which thereby increases the working life of
these. If the LRW contain oil, this should be distilled off with the steam. Development is
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 Declassified and Approved For Release 2013/02/21: CIA-RDP10-02196R000300030004-5
unfinished on methods of mechanical oil separation and sorption of residues on activated
charcoal. It is necessary to increase the working life of carbon filters and the perfor-
mance of the entire oil-purification system.
The final stage in LRW processing is to convert the concentrates to solid form for
storage. Bitumening and cementing are the main methods of solidifying the LRW used abroad.
Until recently, cementing was more widely -sed at lower stations, with bitumening used in
nuclear centers. Since 1975, a cementing plant has been in operation at a nuclear power
station in Sweden. In 1978, the equipment was reconstructed to improve the throughput and
adapted for solidifying mixtures of ion-exchange materials containing boric acid [2].
Cementing is used also as a method of solidifying LRW in France (constructional
hydraulic cements are employed), as well as in the Federal German Republic, Japan, and the
USA [3-7]. In Britain, research is in hand on incorporating ion-exchange materials into a
cement-vermiculite matrix [6], and in the USA studies are being performed on matrices based
on cements with additives [8]. Although cementing technology is well developed, further
efforts are being devoted. in various countries to simplify and improve it to give stronger
materials in which the radioactive nuclides have less tendency to leach out; much attention
is also being given to reducing the volume of the products for storage. In particular, a
method has been developed in the Federal German Republic for cementing previously dewatered
wastes [5].
Studies have been made on the effects of sodium silicates, clay materials, and shales
on the water resistance of materials formed by incorporating radioactive wastes into Port-
land cement, and this is being done not only in the USSR but also elsewhere. For example,
studies in the Federal German Republic [8, 9] have shown that the addition of 20 mass % of
Ca bentonite reduces the diffusion coefficient of 137Cs by a factor of 6000-7000, which re-
duces the leaching rate by about afactor of 80. The leaching of strontium is reduced by
increasing the density of the cement material.
Radionuclides are divided into three classes in accordance with the transformations that
occur in cements [7]:
1) thosethat do not hydrolyze in the unsolidifed cement (pH about 12), for example
137Cs., 51Cr(VI), 1311 , and 3H;
2) those that do not hydrolyze but form insoluble compounds, in particular carbonates,
such as 9 OSr, 14 OBa; and
3) those that hydrolyze in the unsolidified cement with the formation of insoluble
hydroxides such as 6OCo, 54Mn, 114Ce.
Improved properties are attained in polymer-cement concretes or in concretes impregnated
with polymers [10]. In the USSR [11] and in the Federal German Republic [12], studies have
been made on the scope for solidifying radioactive wastes using cementing substances derived
from metallurgical slags.
Bitumening of LRW is used in virtually all countries along. with cementing, and in some
cases this has been implemented. The wastes from the CANDU heavy-water reactors are bitu-
menized in Canada, while extruder bitumenizers have been introduced at nuclear power stations
in the Federal German Republic, and at one power station in Sweden a thin-film bitumenizer has
been in operation since 1976 for solidifying slimes and ion-exchange materials [13], while
in Japan the extruder bitumenizers operating at nuclear power stations such as at Tokai are
being supplemented with bitumenizers having horizontal rotor mixers of throughput up to 140
kg/h [14].
Most of the. equipments currently used to solidify power-station wastes. involve partial
or complete: preliminary dewatering. Examples are provided by extruders, thin-film rotor
equipment, horizontal rotating dryers, fluidized-bed equipment,, and spraying dryers. In
the use of this equipment, the wastes may be incorporated into thermoplastic materials or
into cements, polymers, etc.
Ion-exchange materials are dewatered by vacuum filtration followed by hot-air drying.
In that method, the water content of the ion exchanger is reduced to 25 mass %, and in that
state the resin becomes a material that can be poured. In Japan, a method has been devised
for incorporating dewatered.ion-exchange materials, slimes, and concentrates into thermo-
plastic resins: polyethylene and chlorinated polyethylene [15]. In the USA, a system has
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been built for reducing the volume of power station. wastes by the use of a fluidized bed,
which combines the processes of calcination and combustion, which reduces the volume of the
wastes by a factor of 10. A full-scale plant has operated for two years [16].
In all countries, studies are being made on the specific properties of bitumen com-
pounds, which largely determine the conditions for safe storage: Studies are being made on
the fire safety and biocorrosion of bitumen compounds, as well as the corrosion of the steel
drums holding the solidifed wastes.
In no foreign country with a developed nuclear power industry is there an agreed view
on the methods of solidifying medium-level nuclear power station wastes. Virtually all
countries have devised and used cementing and bitumening techniques, which have been brought
to the stage of industrial use. Studies on waste processing have not only not ceased but
have actually acquired larger scales.
Bitumening is. being introduced for solidification at Soviet power stations. Conditions
have been defined for producing the bitumen compound that ensure that the radioactive wastes
are reliably localized. It has been found that for most LRW the optimum salt content in the
compound is about 50 mass %, while the content of ion-exchange materials should not exceed
40 mass %, while that of filtered pearlite should not exceed 10 mass % to avoid marked in-
creases in viscosity. Concentrates from heterogeneous LRW (filter pearlites and ion-exchange
materials) are solidified periodically.
The optimum bitumening temperature is 160-180?C, at which the water distills off virtually
completely and the hazard of the compound igniting is eliminated. The unloading temperature
for the compound should be such as to prevent separation: In accordance with the composition
of the wastes and the grade of bitumen, it can vary from 80 to 130?C.
A rotor bitumenizer can operate continuously and has high throughput together with certain
other advantages, so this appears the most promising for Soviet nuclear power stations. A
prototype apparatus with a rotor bitumenizer has been based on a standard rotor dryer and has
been operating successfully since 1978 at Moscow Radon production cooperative [17]. Another
type of rotor bitumenizer devised specially for processing LRW has passed marker's tests and
has been sent for prototype testing on real wastes from power stations containing RMBK reac-
tors [18]. This apparatus is intended to process salt concentrates in continuous mode, while
slimes from.filter materials are handled in batch mode. The bitumen compound is to be kept
in a special store equipped with a monitor system and fire-fighting facilities.
Real wastes from power stations containing VVER reactors are being used in testing the
BUR bitumenizer,in which the solution is dewatered (to a water content of about 20 mass %) in
a concentrator, while the bitumening proper (incorporation of the salts into the bitumen) is
performed in a pug mill [19]. The DB-100 semicontinuous bitumening plant with a tubular
bitumenizer has been introduced at Leningrad waste-storage station, which has provided comple-
tion to the LRW processing cycle giving solidified products and reliable radionuclide locali-
zation [20].
Along with the introduction of bitumening systems, studies are continuing designed to
resolve the following problems:
1) improving and cheapening the apparatus;
2) improving the properties of the bitumen compounds to provide reliable isolation of
the radionuclides for hundreds of years; and
3) reducing the volume of the final materials to be stored.
Much attention is now being given to the properties of solidified wastes in order to
ensure safe storage, and in particular estimates have been made of the fire and explosion
hazards in the storage of bitumen compounds, along with the biological and radiation
stabilities [21, 22]. In parallel with the work on bitumening, searches are being made for
alternative materials, including those among the wastes from large-scale chemical processes [23],
which can be used instead of bitumen in the same equipment.
An advantage of bitumening is that the water stability of the product is good (leaching
rate 10-`'-10_5 g/cm2?day), along with the high content of radioactive wastes (up to 50 mass %).
The main disadvantage is the fire hazard in the production and storage of the bitumenized
wastes. Cementing is free from this deficiency, but the water resistance is poor (the
leaching rate is high at about 10-2 g/cm2?day), while the waste contents are low (up to 15
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mass %), which firstly requires the creation of a reliable protection barrier having good
water-isolating properties and secondly improvement of the solid-product composition [11].
Drying produces the largest reduction in the amount of waste going for storage, but the
presence of water-soluble salts in the products complicates the reliable long-term storage.
At present there is no reason to restrict development to one method, since the desir-
ability of any given solidification technology will be dependent on the detailed conditions,
and in particular on the storage method. In the USSR and elsewhere, little is known about
topics related to the final disposal of solidified wastes; there has been no final decision
on the location of the graveyards for bitumen compounds, and at existing power stations and
ones under construction it is proposed to locate the stores directly on the station area or
at a short distance away. Two forms of storage have been developed for the bitumen compound:
in drums or in concrete vessels.
Economic studies are in hand on the best method of disposal in relation to the properties
of the various solid materials. Here one has to bear in mind all aspects relating to environ-
mental protection as well as the economy of the method.
Storage on the nuclear power station site in the form of concrete vessels appears pre-
ferable because this is more compact and does not require comparatively expensive loading
and transportation of the solidified wastes. However, disadvantages are the substantial fire
hazard and the complexity of removing the compound from the stores if necessary. Therefore,
one cannot rule out the possibility of loading the compound into drums for temporary storage
and subsequent shipping to graveyards, which is necessary for example if the hydrogeological
conditions in the region of the power station are unfavorable or for combined heat and power
stations in cities.
At present, much attention is. being given in many countries to underground stores and
graveyards for low-level and medium-level solid and solidified wastes [24], since these are
considerably cheaper than other stores. The creation of stores is necessary not only to
provide reliable containment of the radionuclides but also to provide reliable disposal of
the solid radioactive wastes SRW formed at nuclear power stations, which are divided by
activity. level into low, medium, and high types, and in composition consist of com-
bustible and noncombustible types. The amount of SRW produced at a nuclear power sta-
tion of power 1 GW(el) is indicated by various sources as ranging from 200 to 400 m9/yr.
The solid wastes include combustible ones (clothing, shoes, paper, rags, polyethylene, and
other plastics) and noncombustible wastes (metallic items and broken glass). The processing
of SRW includes collection, sorting, batching, combustion, separation, pressing, and solidifi-
cation.
At all Soviet power stations, the SRW is collected and in the main is stored in special
concrete structures. The filling factors for such stores in some cases constitute 50-60%.
However, the power stations lack systems for sorting, separating, and pressing the material,
which leads to unsound filling of the stores and requires the construction of new costly
structures. Pressing can reduce the volume of SRW for storage by factors of 3-5. Much
attention is being given in the USSR and elsewhere to the development of reliable equipment
for pressing SRW.
When power station wastes are burned, corrosive gases are released (chlorine and so on),
which hinders the creation of reliable gas-cleaning systems and ensuring the necessary puri-
fication coefficients, while there are also adverse effects on the working life of the puri-
fying plant and pipelines, and there may also be the formation of secondary wastes, etc.
There are difficulties in selecting the materials for the furnaces and in automatic control
of the combustion control of the combustion equipment.
The equipment devised and operated at power stations for processing liquid and gaseous
wastes provide for the discharge of radioactive aerosols and gases and effluents containing
radionuclides at levels below the health and safety norms. However, it is necessary to im-
prove the waste-processing systems at power stations to make them more economical and effec-
tive, and. on this aspect there is no final decision, and much attention should be given to
research in this area.
LITERATURE CITED
1. N. S. Babaev et al., in: Nuclear Power, Man, and the Environment [in Russian],
Energoizdat, Moscow (1981), p. 9.
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2. H. Christensen, in: Proceedings of the Symposium on Dealing with Nuclear Power Station
Wastes within Their Areas., Zurich, 26-30 March (1979), p. 333.
3. D. Roy, et al., "The role of additives in the preparation of dense cements for isolating
radioactive wastes," in: Proceedings of the International Symposium on the Scientific
Basis for Nuclear Waste Management, Boston, Nov. 28-Dec. 1 (1978).
4. H. Stuenkel, H. Faust, and A. Pusavala, [2, p. 465].
5. M. Lazer, H. Mallek, and W. Jablonski, [2, p. 373].
6. D. Ferret and W. Simons, [2, p. 169].
7. Hideo Matsuzuru and Noboru Moriyama, Nuci. Sci. Eng., 80, 14 (1982).
8. R. Koster and G. Randolph, Waste Management Research Abstract, No. 12, 28 (1978).
9. E. Zange, W. Schienter, and K. Trumper, ibid., 33.
10. A. Donato, ibid., 37.
11. K. P. Zakharova et al., Inventor's Certificate No. 880149, Byull. Izobret., No. 16, 293
(1982).
12. Federal German Republic Patent. No. 2945007.
13. Z. Harfore, [2, p. 449].
14. Teruo Tokubuchi et al., in: Scientific Basis for Nuclear Waste Management, Vol. 3,
Boston (1980), p. 219.
15. Seiichi Tozawa et al., Nucl. Sci. Technol., 18, No. 2, 162 (1981).
16. R. Vance et al., in: Decontamination and Decomissioning of Nuclear Facilities, Plenum
Press (1980), p. 249.
17. I. A. Sobolev et al., in: Proceedings of the Fifth Comecon Symposium on Researches in
the Reprocessing of Irradiated Fuel, Marianske Lazne (1981), p. 246.
18. V. I. Davydov et al., in: Processings of the Fourth Comecon Conf. on Researches in
Disposing of Liquid, Solid, and Gaseous Radioactive Wastes and Decontaminating Polluted
Surfaces [in Russian], Atomizdat, Moscow (1978), p. 74.
19. A. S. Nikiforov et al., At. Energ., 50, No. 2, 128 (1981).
20. V. I. Davydov et al., [17, p. 240]. -
21. K. P. Zakharov et al., At. Energ., 44, No. 5, 436 (1978). .
22. S. V. Zhukova at al., ibid., 52, No. 5, 326 (1982).
23. K. P. Zakharova et al., ibid., 49, No. 4, 258 (1980).
24. Storage of Radioactive Wastes near the Surface. IAEA Recommendations. Safety Series No.
53, Vienna (1981).
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 Declassified and Approved For Release 2013/02/21: CIA-RDP10-02196R000300030004-5
TREATING RADIOACTIVE WATERS WITH A MIXED ION-EXCHANGE
BED IN A CONTINUOUS-OPERATION PLANT
B. E. Ryabchikov, E. I. Zakharov, A. P. Darienko,
A. V. Rakhchev, and M. Ch. Murabuldaev UDC 621.039.73
There are some technological and economic advantages in using continuous ion exchange
to purify radioactive waters by comparison with the periodic process operated in ion-ex-
change filters [1, 2].
When water is purified by demineralization either continuously or periodically, it is
necessary to pass the solution in sequence through 1-3 pairs of columns containing cation-
exchange and anion-exchange materials. In that case it is preferable to use a mixed bed of
cation-exchange and anion-exchange materials, which enables one to perform a high level of
purification on passage through a single apparatus, which substantially reduces the equipment
size. Installations of this type for preparing feedwater for thermal power stations are
operated abroad [3, 4]. The process does..not differ essentially from the treatment of liquid
radioactive wastes, so the performance of such systems should be high. Difficulties in the
design of these systems are associated in the main with organizing the continuous motion of
the ion-exchange materials, the dispensing of these, and the separation of the spent mixture
into cation-exchange and anion-exchange materials before separate regeneration.
A system of pulsation-type ion-exchange columns has been devised to carry out all the
purification operations: sorption, regeneration, and sorbent washing, which enables one to
conduct the process effectively at low cost [2, 5]. To separate the ion-exchange materials
we have devised an apparatus of novel type, which can separate a mixture of industrial ion-
exchange materials.
Here we describe the proposed semiindustrial system for continuous ion exchange with a
mixed ion-exchange bed of throughput up to 1 ms/h and we give test results in the treatment
of radioactive waters.
The apparatus (Fig. 1) includes the sorption column 1 containing a hydraulically com-
pressed bed of ion-exchanger [6] of diameter 0.1 m and height 2 m, the column 4 for separating
the mixture of saturated resins, and two identical regeneration systems for the cation ex-
changer and anion exchanger, which each consist of the regeneration columns 3 and 6 and the
wash columns 2 and 3 of PSK type (pulsating sorption columns), all of which have fluidized
sorbent beds [5, 7]. There are also additional units: the cation and anion exchanger mixer,
the separator 7, receiving and supply vessels, and monitoring and automatic systems.
The solution to be treated passes into the bottom of the sorption column containing a
mixed layer of ion exchangers, where it is purified from.salts and radionuclides. The
solution speed during the sorption may be 100-150 m/h. Periodically (every 5-15 min) the
mixture of resins in the apparatus moves downward, with fresh resin added at the top, while
saturated resin is removed at the bottom. The exchangers are moved by pulsation, which is
provided by the pulsation chamber 11, which is slowly filled with air through the valve 10,
which is then rapidly discharged through the valve 9, whereupon .the liquid in the process
zone along with the lower layer of ion exchanger is transferred to the pulsation chamber and
the entire layer moves downwards. At the same time, the equivalent amount of the exchanger
mixture enters at the top from the buffer tank.
The saturated exchangers entering the lower zone under the pressure of the liquid (0.1-
0.2 MPa) pass to the resin-separation column 4, which consists of two zones with distributing
plates, with the upper zone having a cross section larger by a factor 1.5-2. The resin mix-
ture is introduced into the middle of the column and water is supplied at the bottom. The
separation is produced hydrodynamically because of the density difference between the cation
Translated from Atomnaya Energiya, Vol. 55, No. 6, pp. 373-376, December, 1983. Original
article submitted January 31, 1983.
0038-531X/83/5506-0815$07.50 ? 1984 Plenum Publishing Corporation 815
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 Declassified and Approved For Release 2013/02/21: CIA-RDP10-02196R000300030004-5
AV-17
Hv2
X
Initial
solution Y
HNO,
rl~
Fig. 1. Apparatus for continuous ion exchange with a mixed bed for de-
mineralizing effluents: 1) sorption column; 3, 6) regeneration columns;
2, 5) wash columns; 4) separation column; 7) separator; 8) mixer tank;
9, 10) valves; 11, 13) pulsation chambers; 12) air lifts; 14) air supply;
SD) special drain.
and anion exchangers. The cation exchanger is the denser and collects in the lower zone,
from which it is pumped by an air lift system to the regenerator, while the anion exchanger
fills the upper zone and is transferred along with the solution into the separator 7, from
which it also passes to regeneration.
To facilitate management, the demonstration separation column has been made transparent,
while the cation-exchange material has been colored with methylene blue dye. The cation ex-
changer is partly decolorized on acid treatment, but a color is still readily visible even
after 50 sorption-regeneration cycles.
The saturated cation-exchange material passes to the top of the regeneration column 3
containing a fluidized bed ofsorbent, and this is of PSK type, where there is motion from
the top downwards, with the resin freely falling in a flow of rising regeneration solution.
The pulsation and the distributing plates serve to improve the contact between the phases.
The regenerated cation exchanger is collected in the lower part of the column, from which it
is transferred by an air lift for washing in the column 2 of analogous design. The washed
cation exchanger is transferred by air lift into the mixer tank 8.
The bottom of the wash column 2 is supplied with water, which washes the remainder of
the regeneration solution out of the resin particles. The resulting solution contains up
to 0.5-1.0 g-eq/liter of HNO3, and this is taken from the top and flows spontaneously because
of the cascade arrangement of the columns to the bottom of column 3. HNO3 is constantly
supplied to the transfer line with a concentration of 12 g-eq/liter. The regeneration solu-
tion passes through column 3 from the bottom upwards in the opposite direction to the sorbent
and regenerates it. The resulting.regenerate is taken from the top of the column and sent
for processing. The chain of columns 5 and 6 for regenerating the anion exchanger works
similarly.
The sorbents.are transferred by air lift, and therefore they are separated from the
transporting solution by the use of drum-worm separators. The regenerated resins are collected
in the mixer tank, where they are mixed by means of air and are transferred automatically to
the sorption column as the resin mixture is used up.
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TABLE 1. Average Data on the Treatment of Liquid Wastes with a Mixed Exchanger Bed
Electrical
Concn., mg-eq/liter
RActivity in nCi/liter (1 Ci = 3.700 X 10 Bq )
conduc-
tivity,
Na I
Mgl
ca
Cl I
3O4 I
?ry
?Sm/cm
,
Experiment
Initial-solution
7
6
350
2,3
3,4
1,2
3,4
200
68
110
4,7
1,5
24
8,7
Effluent
,
6
1
8,8
0,08
0,13
0,1
0,05
3
0,33
0,58
0,41
0,003
1,6
(1,005
Purification coefficient
,
29
26
12
68
66
206
190
11,4
470
15
1720
Experiment 2t
Initial solution
6
7
230
5
2
3,5
1,2
3,4
200
68
120
17
1,5
17
8,7
Effluent
,
6
1
4,8
,
0,08
0,05
0,1
0,05
1
0,18
0,18
0,35
0,003
0,31
0,005
Purification coefficient
,
31
70
12
68
200
380
665
485
470
55
1720
P#
22
V a n , = t Vcat :Van=1 :2; Qs =0,8 m /h
# PCB is the permissible radionuclide concentration in water (see NRB-76, OSP-12/80, Energoizdat (1981), p. 35).
TABLE 2. Parameters of Equipment Using
Pulsation Columns and Ion-Exchange Filters
of Throughput 1 M3 /h with Identical Degrees
of Purification
Number Volume, liter
water
fl
Apparatus
Concn.
ow
for in-
type
quip.
vessels equip.
vessels
ion ex
coeff.
ternal
chan
use, %
Filters
6
6 900
1200
600
50
5-
Columns
6
2 250
80
39
70
1
,5-2
The.treatment is applied to a solution (Table 1) that has previously been averaged,
coagulated, and mechanically filtered. Preliminary experiments on purifying this solution
with laboratory filters led to the selection of the strongly acid cation exchanger KU-2-8 and
the strongly basic anion exchanger AV-17-8. The exchange materials are regenerated in solu-
tions of BM03 and NaOH of concentration 2 g-eq/liter. Tap water is used to separate the
resins and wash them.
The purpose of the tests was to determine the performance in purifying the solution
from radionuclides using a mixed ion-exchange bed, together with the scope for subsequent
separation of this into its components, and also the scope for producing a uniform mixture of
the regenerated exchangers. Most of the radionuclides in the solution that make the main
contribution to the overall a activity are usually in cationic form (134'137Cs, '90Sr),,so we
initially checked the state in which the cation-exchange material is regenerated with a large
excess (200%), while the anion exchanger, which in the main only corrects the pH, was
regenerated with a small excess (50%). The volume ratio of the exchangers in the mixture was
lel, and the flow rate was 5-6 liter/h. The experiments were performed continuously for 100
h, and the results are given in Table 1.
As would be expected, there was complete extraction of the salt cations and of the radio-
nuclides in cation form. The purification factor for the salts was about 30. The concentra-
tions of all the radionuclides were substantially less than PCB. However, the extraction of
124Sb and 6 ?Co in anion or complex form was incomplete. The relatively high electrical
conductivity was'due to considerable amounts of unextracted carbonic acid (up to 50%.of the
sum of the anions in the initial solution).
More extensive purification from radionuclides in anion form requires an increase in the
degree of purification from salt anions. For this purpose, a second series of experiments
was performed in which the regeneration of the anion exchanger was greater (excess alkali
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180%) and the amount of anion exchanger in the mixture was increased to make the ratio of
cation and anion exchanger volumes 1:2. The consumption of acid in regenerating the cation
exchanger was reduced to 50-80% excess (the flows of cation exchanger and anion exchanger
were then 3.3-4.0 and 6.6-8.0 liter/h correspondingly).
In the second series, the fuller regeneration of the anion exchanger led to better
extraction of the weak-acid anions and radionuclides in anion form, and there was also
improved performance in the cation exchanger, which is characteristic of a mixed bed. The
contents of all the radionuclides were at the limit of detection of the radiometers, while
the salt content was reduced to an extent such that the conductivity was 2-5 x 10-6 Sm/cm.
Water of such'quality can not only be discharged to any body of water but can also be used
to supply steam boilers.
The separation of the saturated resins (flows of 10-12 liter/h) was performed with tap
water with a flow rate of 20-22 liter/h. The resulting partially demineralized solution can
be used to wash the regeneration solutions out of the resins. The separation boundary be-
tween the cation and anion exchangers was very sharp. The position was controlled by the
output of the air lift handling the cation exchanger. Ina check on the resin separation
performance it was found that the content of cation exchanger in the anion exchanger was not
more than 3-4%, while that of the anion one in the cation one was 2%. This degree of
separation is more than sufficient to provide demineralization to a conductivity of 1 x 10-6
Sm/cm.
The resins were washed free of the regeneration solutions with a flow ratio of 1:1 or
1:2, and all the washed solution was then treated with acid or alkali and used in the
regeneration. The overall flow rate of the wash water was about 15 liter/h. Despite this
low flow rate, the washing performance was high, as is evident from the low conductivity of
the purified solution.
It was not the purpose of this study to-minimize the flow rates of the regeneration
solutions, since it had previously been shown [2] that it is possible to reduce the volumes
by factors of 1.5-3 relative to the traditional levels in the continuous process. However,
in that case also we obtained good results in regenerating both resins.
The flow rates of the regeneration solutions were less than 2.5% of that of the initial
solution. The mean content of sodium salts in the acid regeneration solution was 260 eq/m9,
hardness 330 eq/m9. The concentration coefficients for salts and radionuclides is about 100,
which is better by afactor of 1.5 than in the traditional equipment (two-stage ion exchange
in filters). Analysis of the saturated and regenerated resins showed that the degrees of
regeneration for the various radionuclides and ions vary from 50 to 90%, which was sufficient
to provide complete purification by the mixed exchanger bed.
Therefore, this continuous-operation plant with a mixed exchanger bed can provide a
high degree of purification for liquid radioactive wastes with initial salt contents up to
0.5 g/liter at high solution speeds (70-100 m/h) and small flows of regeneration solutions
(up to 2.5% of the volumes of the input solutions) with the absence of wash water returning
to the start. The equipment dimensions and the load of exchangers are also much less than
in the traditional apparatus (Table 2).
LITERATURE CITED
1. F. V. Rauzen et al., At. Energ., 36, No. 1, 27 (1974)..
2. B. E. Ryabchikov et al., At. Energ., 38, No. 4, 222 (1975).
3. C. Thorborg, Power, 113, 76 (1969).
4. C. Dallmann, in: Proc. 32nd Int. Water Conf., Pittsburgh (1971), p. 113.
5. S. M. Karpacheva and E. I. Zakharov, Principles of the Theory and Calculation of
Pulsation-Type Column Reactors [in Russian], Atomizdat, Moscow (1980).
6. B. E. Ryabchikov, E. I. Zakharov, and V. S. D'yachkov, Inventor's Certificate No.
712118, Byull. Izobret., No. 4, 16 (1980).
7. Pulsation Columns: Catalog, Part 2 [in Russian], Atomizdat, Moscow (1981).
818
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 Declassified and Approved For Release 2013/02/21: CIA-RDP10-02196R000300030004-5
EXPERIMENTAL INVESTIGATIONS OF p-ATOMIC AND p-MOLECULAR
PROCESSES IN HYDROGEN ON THE JINR SYNCHROCYCLOTRON
V. P. Dzhelepov and V. V. Fil'chenkov UDC 539.1.18:539.1.19:621.039.633
Two main stages can be identified in the development of experimental investigations on
P-atomic and P-molecular processes in hydrogen. The first stage began in 1957 when the
Alvarez group first recorded [1] the p-catalysis of fusion reactions in p + d and d + d
nuclei, which had been predicted theoretically earlier [2, 3]. By 1960 the foundations had
been laid for the theory of p-atom proceses, the main results of which were expounded in a
review by Zel'dovich and Gershtein [4].
In 1957-1965 processes involving p atoms were investigated intensively in a number of
scientific centers such as Columbia University, CERN, the Joint Institute for Nuclear Research
(JINR), etc. These processes were studied most extensively in the Laboratory of Nuclear
Problems. at JINR on a 680-MeV synchrocyclotron using a high-pressure diffusion chamber placed
in a magnetic field [5]. The use of a track detector during a stage of the investigations
when many of the experiments were in the nature of a search ensured the success of this work.
The main results were published in [6]. As for other experiments carried out during this
period., we should point out the work done by the group of Lederman [6]. The results they
obtained for the rate of formation of ppp mesic molecules and the rate of stripping of a muon
from a proton by a deuteron are consistent with our data [5].
By 1965 many laboratories had experimentally investigated the characteristics of many
processes initiated by negative muons in hydrogen (H2, D2, and H2.+ D2 mixture) and in most
cases the results were in good agreement with the predictions of the theory [8]. To a
considerable degree these investigations also fulfilled the "applied" objective, viz.,
experimentally obtaining information about p-atomic processes so as to permit correct inter-
pretation of the results of measurements of the rate of the fundamental reaction of weak
interaction, i.e., nuclear absorption of a muon by a proton.
A special place in the physics of p-molecular processes is occupied by p-catalysis of
nuclear fusion reactions synthesizing hydrogen isotopes. The essence of p-catalysis is well
known. Since in muonic hydrogen molecules the distance between nuclei is small (5.10-11 cm),
the probability of nuclei penetrating the Coulomb repulsion barrier becomes high enough so
that the nuclear fusion reaction would occur with a probability of nearly one during the muon
lifetime,(-up = 2.2.10-6 sec). Under conditions when the rate of formation of a muonic mole-
cule and the rate of the nuclear reaction in it are much higher than the muon decay rate
(ao = 1/TU = 4.55.105 sec-'), one muon can successively initiate many fusion reactions, with
a release of energy in each of them. It is of the utmost interest to determine the maximum
possible number of successive muonic-catalysis events. According to the concepts of the
theory which existed in the early 1960s, an efficient (multiple) muonic catalysis process
was impossible for two reasons [4]: first, because of insufficiently high expected values of
the rate of formation of.muonic molecules and, second, because of the finite probability of a
muon "sticking" to one of the fusion reaction products, the helium nucleus, which breaks the
chain of successive catalysis.
The results of the first experiments on muonic catalysis of p + d and d + d reactions,
it seemed, supported this pessimistic conclusion of the theory. The process of formation of
dtp molecules with.a subsequent reaction d + t,_* `'He + n + 17.6 MeV, it is true, had not been
investigated at all experimentally. In this case, however, the theory predicted a small
value (-104-105 sec.') of the dtp formation rate. All of this considerably dampened interest
in further experimental investigation of p-catalysis. It is significant that from 1963 to
1977 foreign laboratories did not carry out a single experiment in this area.
Translated from Atomnaya Energiya, Vol. 55, No. 6, pp. 376-391, December, 1983. Original
article submitted April 11, 1983.
0038-531X/83/5506-0819$07.50 ? 1984 Plenum Publishing Corporation
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Meanwhile, precisely the study of the process of muonic catalysis revealed an interesting
new effect which did not fit within the framework of the theoretical concepts at the time.
The essence of this effect was that the value of the rate of formation of ddu molecules
(Addu), which we measured [5] in gaseous deuterium at 240?K, was roughly an order of magnitude
higher than that obtained in liquid deuterium [9] at 30?K. In order to explain this anomaly,
we hypothesized [5] the existence of a previously unknown mechanism of ddp molecule formation,
a mechanism.that would be resonant in the energy of the du atom (in the deuterium temperature).
It must be pointed out that back in 1954 Zel'dovich [3] pointed out that the existence of a
weakly bound state in a muonic molecule could result in a resonant enhancement of the u-cataly-
sis yield.
Later, in 1967 E. A. Vesman considered a concrete mechanism [10] of formation of ddu
molecules such that the hypothesized resonance dependence Xddu(T) could be realized. At the
same time, however, the hypothesis that a resonance existed in the cross section for ddu for-
mation remained insufficiently substantiated, both theoretically and experimentally. On the
one hand, it had not been established reliably that ddu had a weakly bound state leading to
resonance while, on the other hand, the body of experimental data (in actual fact, only two
points) was not sufficient to establish the character of Xddu(T).
Another important problem, which was not solved in the initial stage of the research,
became particularly' clear upon completion of measurements [11] of the rate of muon trapping
by a deuteron in gaseous hydrogen (H2 + 5% D2, 0.69 MPa) and was associated with the deter-
mination of the spin states of du atoms. Such atoms, formed initially in a random mixture
of states with a total spin Fdu = 3/2 and Fdu = 1/2, can then fall to the lower state with
Fdu = 1/2. Until recently it was assumed [4, 8] that these transitions take place only in
du + d spin-exchange collisions. The initial estimate [12] of the rates Ad of these tran-
sitions, it seemed, was confirmed in the experiment in [7]. To reconcile the results of the
experiment in [11] with those expected from the weak-interaction theory it-was necessary to
assume for Ad a value several orders of magnitude higher than the result of [12] and the data
of later calculations [13, 14] or to assume the existence of some other (not spin-exchange)
mechanism [14] leading a high rate of 3/2 -> 1/2 transitions in a gaseous hydrogen-deuterium
mixture.
Mainly in view of these two problems, viz., the question of the existence and the nature
of the resonance dependence Addu(T) and the problem of determining the spin states of du atoms,
the Laboratory of Nuclear Problems at JINR decided to carry out a new cycle of investigations
of u-atomic and u-molecular processes; this cycle began in 1974-1975.
It must be pointed out that while in the early 1970s such investigations were conducted
practically nowhere (the sole exception was the CERN group), at the present time experiments
along these lines are being carried out on many accelerators (B. P. Konstantinov Institute
of Nuclear Physics of the Academy of Sciences of the USSR, Leningrad; Swiss Nuclear Research
Institute; meson factory in Los Alamos). Thus, one can speak of a new stage of intensive
research on p-atomic and p-molecular processes in hydrogen. This stage is characterized
primarily by the fact that in recent years significant progress has been made in attaining a
more profound understanding of u-atomic effects and in determining their quantitative charac-
teristics at a new level of accuracy.
We point out only two of the most important achievements of the theory: weakly bound
states with an energy of approximately 2 eV and 1 eV, respectively, have been shown to exist
in ddp and dtp systems; higher resonance values of the formation rates for these systems have
been predicted.
Experiments have obtained substantially improved capabilities: Work has started with
meson factories at Los Alamos (USA) and Villingen (Switzerland) making it possible to obtain
high-intensity meson beams. All of our measurements pertaining to the new stage of investi-
gations have been carried out on the same 680-MeV synchrocyclotron, but with improved charac-
teristics of the muon beam.
Investigation of the Processes of Formation of Muonic ppp and pdp Molecules in Gaseous
Hydrogen. Measurement of the rate of formation of muonic ppu and pdp molecules (appu and
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TABLE 1. Formation Rate of ppp and pdp
Molecules, 106 sec-'*
APP! ;pdr
Experiment
Columbia Univ., liquid H2 [7]
1,89?0,2
5, 8?3
CERN, liquid H2 1151
2,55:?:0,18
6,82?0,25
JINR, gaseous H2 [161
2,74?0,25
JINR, liquid H2 [11)
2,34?0,17
5,53?0,16
Calculation
Ref. [14]
5,9
7The -values of ],0,f(~ and X?Odp are given for a liquid-hydrogen den-
sity p0 = 4.25.101` nuclei/c.3 and can be converted for gaseous
hydrogen with a density p: app? = XPP1+V and X Pd? = X~?~,
where v = p/p0 is the relative density.
Apdp) not. only is important from the point of view of comparing experimental results and ex-
plaining their agreement with present-day calculations but also is entirely necessary in order
to choose the optimal conditions for the formulation of experiments on muon capture by a
proton and a deuteron and for the correct interpretation of the data obtained in such experi-
-ments. It was also important for us to know Apdp for gaseous hydrogen since this enabled us
to obtain information about the spin states of du atoms on the basis of measurements of the
yield of the nuclear fusion reaction in a pdu molecule.
The data from previous measurements of A and A are given in Table 1. As can be
PPp Pdp
seen, Apdp had been measured earlier only in liquid hydrogen. The value of appp for gaseous
hydrogen was measured only in [18]; the data processing method employed there did not exclude
a certain systematic error in the result.
In order to determine appp and apdp we employed the method of [15, 181, in which the
target consisted of hydrogen with a small impurity of gas with Z >> 1 (in our case we used
xenon with an atomic concentration of 10-5). The sequence of processes occurring in an H2 +
D2 + Xe mixture after a negative muon has halted in it is shown schematically in Fig. 1.. In
theexperiments conducted by our group [17] H2 + Xe and H2 + D2 + Xe mixtures were used; we
detected electrons from muon decay and y rays from mesic x radiation, arising as a result of
the process of stripping of a muon from a pp or dp atom to the excited state of an Xe p atom.
The time distributions of electrons and y rays can be written in one way for H2 + Xe and
H2 + D2 + Xe:
dn,,Idt = Xxe exp (- XBt);
dne
dt
XAXe ) exp (-1.8t) -}- exp (- x0t).
1 " 7v8-~o-7vjXe
In these expressions for the H2 + Xe mixture as = apu is the rate of decay.of a pp atom,
ax = appp, and AXe = A is the rate of stripping of a muon from a proton by xenon; for the
pXe
H2 + D2 + Xe mixture as = adp is the decay rate of a dp atom, Ax = apdu, and AXe = AdXe is
the rate of stripping of a muon from a deuteron by xenon. The remainder of the notation is:
ao = 4.55.105 sec t is the rate of decay of a free muon, ao is the rate of p-decay in an Xe
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3He~u+~(S,S MeV)
Qu
I '(Xa a0 t'pk1 I
I I
Fig. 1. Scheme of p-atomic and p-molecular
processes in a hydrogen-deuterium mixture.
u-atom, and X%e is the rate of nuclear capture of a muon in an Xe p-atom.
The decay rate of p atoms is Xpu = Xo + (Appu (1 - CD) + (pCXpM,,xe + cCDapd; ad? - 10 +
nnvu in gaseous hydrogen (H2 + 5% D2, 0.69 MPa) were completed in 1973 [11].
It must be said that it is very important to investigate this process both in order to
measure the constants of p capture in the simplest Gamow-Teller transition and in order to
determine the parameters of the nn interaction. As we have already pointed out, however,
the result obtained in [11] proved very difficult to explain within the framework of the
existing theory since it was necessary to assume the existence of fast irreversible 3/2 -}
1/2 transitions under the conditions of the experiment of [11].
The aim of the experiments conducted by our group [19] was to measure the absolute
yield of process (3a) and the time distribution of y rays from the reaction pdu } 3Heu+ y in
order to determine the rate of this reaction information about the spin states of dp atoms
in gaseous hydrogen. The measurements were carried out with a H2 + 7% D2 mixture at a
pressure of 4.12 MPa, i.e., under conditions similar to those of the experiments of [11].
In order to investigate process (3a) we used the same apparatus that had been used to
measure Appu and Apdu(see Fig. 2). In the experiment we carried out two main exposures:
with an H2 + D2 mixtures and an experiment with helium (background). The logic of the
experiment consisted in selecting events recorded by the y- and e-detectors in 10 usec after
stopping of a muon in the target.
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The measured time distributions of y rays from reaction (3a) were analyzed using Eq. (4),
the values found in these experiments for the electron yield from the decay, and the experi-
mentally measured y-ray counting efficiency (see Table 3). As can be seen, the value ob-
tained for xPd is in good agreement with the measurement of [7]. It would be interesting to
compare the cross section for the reaction pdu + 9Hep + y with the cross section for the
reverse process y + 9He + p + d as well as with the cross" section for the reaction n + d -ot
+ y at zero energy, but in order to do this it is necessary first to calculate the Coulomb
barrier factor in pdu.
The data we obtained on the yield and time distribution of y rays from process (3a) are
in agreement with the theoretically expected values for the statistical population of the
HFS states of the molecule. I t thus follows that contrary to the conclusion of Bertin et al.
[11] the character of the population of the spin states of du atoms in gaseous hydrogen is
nearly statistical.
Analysis of the time distributions of the y rays with the aid of Eq. (4) made it possible
to obtain data on the rate of the 3/2 + 1/2 transitions in collisions of dp atoms with deu-
terons (at a confidence limit of 90%):
)d De. At 300?K the mean energy of
thermalized p atoms is edu 2.0.04 eV, i.e., is close to Ac. Moreover, because of the small
cross section for thermalization in du + p collisions and the low density, of the deuterium
the thermalization of du atoms under the conditions of our experiments was most likely in-
complete (according to estimates the average thermalization time is -1 usec, i.e., is
comparable with the average lifetime rj? = loo + 1'd,,(P (1 - CD)J'1 x 1.3 usec). This means that
besides 3/2 + 1/2 transitions the reverse 1/2 -r 3/2 transitions could also occur. When these
circumstances are taken into account the estimate given by Eq. (5) may be underestimated.
Conclusions as to the state of the problem of spin states of du atoms will be given at
the end of the next section.
Measurement of the Residual Polarization of Negative Muons in Hydrogen. Investigation
of the process of depolarization of negative muons in hydrogen is of interest from at least
two points of view. One the one hand, the residual polarization Pu is determined in great
measure by the features of the cascade de-excitation of the muonic atom [26, 27] and, hence,
measurement of Pp makes it possible to obtain information about the relatively little
investigated initial stage in the life of the muonic atom. Moreover, the process of muon
depolarization* in hydrogen. is affected appreciably by various u-atomic processes, the
principal processes being spin-exchange collisions [4, 8, 14] of pu atoms with protons and
of du atoms with deuterons in the ground state. Thus, a new approach becomes possible to the
study of these processes, including transitions between the HFS states of muonic hydrogen
atoms. At the same time, unlike the yields of the u-capture or p-catalysis reactions, a
decrease in the polarization of muons in spin-exchange collisions is observed regardless of
whether or not these transitions are. irreversible.
*Here and henceforth in speaking of negative muons we omit the word "negative."
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Cascade depolarization occurs because of the spin-orbit s-Z interaction (s is the
muon spin, Z is the orbital moment) at those levels whose width r is much smaller than the
fine splitting A, i.e., wherever the muon spin manages to complete many rotations around
the total moment j = s + Z. The radiation width is always smaller than the fine splitting
(by a factor a-' = 100). Another mechanism of de-excitation, Auger ionization, dominates
at the upper levels, where the corresponding width is rAug >> 0, but its rate falls with a
decrease in the principal quantum number n. Depolarization, obviously, occurs when rAu
g
A. Since A - Z4 (Z is the atomic number) and rAug depends weakly on Z, this condition is
satisfied at different values of the principal number n = no for light and heavy atoms.
As follows from [27], for the lightestmesic atoms no = 3-4.
The residual polarization Ps-Z, calculated with allowance for only the: spin-orbit
interaction, can be represented as a function of no [27];
Pe-t x 2c3 (no+s) (no-1).
From Eq. (4) we can see that for no = 3-4 we have
P,_ix0.30.
An additional loss of polarization occurs because of the hyperfine s -I interaction (I
is the . nuclear. spin). The residual polarization due to the hyperfine interaction for pu and
du atoms in a statistical mixture of HFS states is [28]
PS-i (p?) =1/2; P i (dp) =11/27. (8)
Thus, when a muon passes through an atomic cascade its polarization is expected to be
p?,= SP,_,P8_,= (0.10 - 0.12), (9)
where the values of Ps-Z and Ps-1 are found from Eqs. (7) and (8). and a = 0.85 [29] is the
initial polarization of the muon beam.
Being in the ground state, the hydrogen atoms undergo exchange collisions [4, 8, 12,
30] (this refers to the exchange of a muon between two nuclei) with protons or deuterons,
and this results in a further loss of polarization. An effective mechanism of depolarization
is that of exchange collisions with a change in the total spin of the atom. When the energy
of the atoms is lower than the energy of hyperfine splitting the only possible transitions
are
P1t(F=1)+P'-'P'?(F=0)+P;
(10a)
dp(F=3/2)+d'-+d'?(F=1/2)+d.
(10b)
In each collision (10a) the polarization is lost completely while in the collision (10b)
the polarization decreases by roughly an order of magnitude.
Muon polarization is also lost in exchange-collisions without a change in the total
spin of the atom but with a reorientation of the spin, i.e., with a change in the projection
on the direction of the initial polarization of the muon. In this case the polarization de-
creases on average by roughly a factor of two in each collision. As follows from [12, 30],
the cross sections for exchange collisions with and without a change in the spin are of the
same order of magnitude. Thus, as estimates of the expected values of the depolarization
rate yd in exchange collisions we can take yd(H2) - Xi-.o and Yd(D2) - As/2 -s 1/2; then, in
accordance, with the results of the calculations in [13, 14] we have
Yd (H2) - X1-o =1.3.1010sec-1;
1'd (D2) - X03/2--1/2 = 4.7 ?107 sec 1.
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IMMIMMEM
Fig. 3. Experimental setup for measuring
[24] the residual polarization of negative
muons in gaseous hydrogen: 1-3) muon detec-
tors (plastic scintillator); 4) moderator;
5) CsI(T1) detector of slow muons; 6) magnet;
7) CsI(Tl) detectors of muons and electrons;
8) target; El and E2 are electron detectors
(plastic scintillator).
From Eqs. (11). and (12) it follows that rapid total depolarization (in roughly 10-9 sec at
3.92 MPa) is expected in protium. In deuterium the depolarization rate is substantially
lower and should have a value yd b 2.5.106 sec-' at a pressure of 3.92 MPa. Before the ex-
periments of our group [24] the residual polarization of muons had been measured [31, 32] only
in liquid protium.
The experimental [31, 32] values of the residual polarization P11 of negative muons in
hydrogen and deuterium are presented below (in %):
Liquid H2 [32] ........................3 + 3
Liquid H2 [31] ................. .7 + 4
H2 at 3.92 MPa [24]............. ..0.3 + 0.9
D2 at 3.92 MPa [24] ................1.0 + 0.9
We see that the accuracy of the earlier measurements. of Pu had been 3-4% and, therefore,
definite conclusions about the depolarization cross section were difficult to make from a
comparison of the experimental results of those investigations with the expected theoretical
"initial" values (i.e., until the formation of a p atom in the ground state) Pu as 10%.
Our goal was to measure PV not only in protium but also in deuterium with an
accuracy several times better than in [31., 32] and, most important, under the conditions
of gaseous hydrogen, making possible a sharp increase in the sensitivity of the data to
the depolarization cross section.
In order to determine the residual polarization we employed the well-known VSR method
(precession of the muon spin), which records and analyzes the time distribution of electrons
from the decays of muons that have stopped in a certain volume placed in a magnetic field.
The.time distribution of electrons recorded in the plane of precession has the form
Ne (t) = B esp (- At) [1-- A exp (- yt) cos (wt + i[.)], (13)
where A is the rate of muon decay in the material (for hydrogen A = ao = 4.55.103 sec-'), Yd
is the rate of depolarization, A = Pu/3 and w, respectively, are the amplitude and angular
frequency of precession, i is the initial phase, and B is the normalization constant.
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2
Time, ?sec
Fig. 4. Time distributions of electrons,
measured in experiments with graphite for H =
70 Oe (a); with protium (b) and deuterium (c)
for H = 140 Oe. Each distribution has been
normalized to exp (-t/T), where T is the muon
lifetime in carbon or hydrogen. The curve in
Fig. 4a is a function of the form y = A cos x
(wt + ip) with optimal values of the parameters
A, w, and ~, found by computer.
For the chosen value H = 140 Oe (1 Oe = 79.57 A/m) the expected values of the precession
frequency in protium and deuterium were
w (FPIL = 1) = 0.34w?= 4.1 rad/?sec;
co (F(,? = 3/2) = 0,30w? = 3.6 rad/psec,
(14)
(15)
where w = 8.53.10" H (0e) is the precession frequency for a free muon (rad/sec). The values
of Eqs. (14) and (15) were obtained using the expressions found in [28] for the gryomagnetic
ratios in the upper and lower spin states of the u atom.
The experimental setup for measuring the residual polarization of muons in hydrogen is
given in Fig. 3. As in our previous work, in the measurements we employed a gas target with
internal CsI(T1) scintillators. In the experiments we used ultrapure hydrogen (containing no
more than 107 of impurities with Z > 1). The magnetic field was produced by Helmholtz coils
with R = 42 cm. The spatial inhomogeneity and the time instability of the field strength did
not exceed 1%. Electrons from decay were recorded by detector 7 [a CsI(T1) cup] and two
scintillation detectors El and E2 (350 x 250 x 60 mm).
During the measurements. we carried out several experiments. The main experiments
involved exposure with protium and deuterium (pressure 3.92 MPa) in a field H = 140 Oe. In
order to check possible sources of systematic errors, we also carried out control measurements
with graphite for which the amplitude and frequency of precession (Fig._ 4) had been known
earlier [29].
Upon analyzing the experimental data we established that the time distributions of elec-
trons, obtained in experiments with protium and deuterium, are described well by a simple .
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b
o
0 0
0 0 0 0
0,5
0
-0,s
0,5
0
-0,5
zz d
0 0 0 0
f o y o 0
O O O 0 O 0 0 0 n O
0 00
0 4 8 12 16
Precession frequency, rad/usec
Fig. 5. Time distributions of electrons ob-
tained in exposures with graphite with H =
140 Oe (a), H = 70 Oe (b), H = 0 (c); and
with protium (d) and deuterium (e) for H
140 Oe.
expression of the form y = B exp (-At) + C. Thus, these data are not at variance with the
assumption of zero residual polarization of muons in hydrogen. In the next stage of the
analysis we compared the experimental distributions with the function
y=Bexp(-?t)[1-{-ACos ((Ot-?-p)J+C, (16)
i.e., with an expression of the form (13) for yd = 0. During the analysis we either inde-
pendently varied all the parameters of Eq. (16) or we varied the parameters A, B, X, and
p for a number of successive fixed values of ,w - so-called frequency analysis (Table 4 and
Fig. 5). These data indicate that the values of the muon lifetime (T11 - X-1) in carbon and
hydrogen are in agreement with the data of measurements in [33] while the values of the
frequency of precession in graphite are in agreement. with the expected values w = wu for
H = 70 and 140 Oe. The accuracy of the determination of A, A, and w is also in agreement
with the expected accuracy. This indicates the absence of appreciable systematic errors in
the measurements.
From the data of Table 4 and Fig. 5 we can see that within the limits of experimental
errors we do not observe the amplitude of precession in hydrogen and deuterium to exceed
the average value A z 0. The values of the residual polarization P. of muons in hydrogen
for the frequencies (14) and (15) are given above.
Further analysis gave estimates of the depolarization rate of muons in hydrogen,
Yd(H2), Yd(D2) > 2.106 sec-1 (at a 90% confidence level), or for the density of liquid
nitrogen
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TABLE 4. Analysis of the Time Distributions
of Electrons Using the Expression y(t) _
B exp(-Xt) [1 + A cos(wt + iir)]
Conditions of
measurement
106 sec 1
w, rad/? sec
A. %
Graphite, H = 140Oe
0,492?0,002
11,8?0,2
4,9?0,2
Graphite, H = 700e
0,494?0,002
5,8?0,3
5,1?0,2
H,, t1=140 Oe
0,454?0,002
4,1
0,08?0,25
D,i H-140 Oe.
0,455?0,002
4,8
0,28?0,25
1'?d(H2)>4.107 sec1;
-ed(D2)>4.107sec1.
(17a)
(17b)
If, in'accordance with the theory of [26, 27], we assume that after an atomic cascade
polarization is conserved at a 10-15% level, then the estimates (17a) and (17b) should be
assumed to apply to the depolarization rate in the ground state of the u atom, i.e., in spin-
exchange collisions and, possibly, in collisions with hydrogen molecules with a change in
their rotational state [14].
In order to compare the estimates (17) with the theoretical predictions we must use
Eqs. (11) and (12). The-results of comparison show that the value given by Eq. (17a) is
consistent with the large calculated value (11) for the rate of spin-exchange collisions of
Pp atoms with protons.. Data on the depolarization of muons in deuterium (17b) are consistent
with the calculations of [13, 14] and the experiment of [23], which yielded the value
X3/2- 1 /2 = (4.3 h 0.1) .107 see
and are in poor agreement with earlier estimates [12].
Analyzing the whole body of data on the spin states of du atoms, we can make the
following conclusions.
1. The conclusion of Bertin et al. [11] that fast irreversible 3/2 -> 1/2 transitions
exist under the conditions of their experiments (H2 + 5% D2i 0.69 MPa) does not find explana-
tion in the theory [8, 14] and is not supported experimentally [19]. Accordingly, as before,
major difficulties are encountered in interpreting the result of measurements of the rate of
capture by a deuteron in [11].
2. The results of our measurements [24] of Pp and of the experiment in [23] are in
good agreement with the A3/2 -. 1/2 calculations in [13, 14], and our data [19] perhaps are not
at variance with them.
3. New investigations, both theoretical and experimental, are necessary. Apparently, it
makes sense once again to consider the reaction p + d -' 3He + y in pdu and the process of the
3/2 1/2 transition. Of great interest is the mechanism, proposed in [14], of the 3/2 -> 1/2.
transition as a result of the interaction of the spin of a du atom with the orbital moment of
H2, D2, or HD molecules, since this mechanism may turn out to be resonant in the energy of
-~.
the du atom. It is of interest experimentally to measure the yield of.the reaction pdp
3Heu +.Y and the rate of the p -d 3 nnvu process in an H2 + D2 mixture as a function of the
deuterium density.
Taking account of the concepts of the theory of muon depolarization in hydrogen, we can
hope to measure the nonzero residual polarization and its time dependence in gaseous deuterium
at a pressure 106 sec-'. The value of addu was found on the basis of measurements of the
neutron yield from reaction (22a) using the calculated value of the neutron detection effi-
ciency and turned out to be consistent with those measured earlier [5] in a diffusion chamber.
A distinctive feature of the measurements in [20] was the use of a gas target with
internal scintillators for investigations over a wide range of temperatures. The main diffi-
culties which had to be overcome involved ensuring the necessary parameters of the detectors
with internal scintillators as well as with maintaining the hermeticity of the target and the
necessary purity of the deuterium filling it.
In order to separate the neutrons and y rays we made use of the difference in the shape
of, the light pulse in stilbene for electrons and protons. For each neutron detector we
measured the amplitude of two signals, which were proportional to the total area Al of the
light pulse and its "fast" part Af, respectively. In the two-dimensional distribution (A1,
Af) the neutron and electron events are located in two-overlapping regions. The. events due
to the detection of neutrons from reaction (22a) should be projected onto the Af axis in the
form of a characteristic stepped spectrum with a limit corresponding to the energy En = 2.5
MeV. Figure 7 shows the distribution (A1, Af) obtained in an exposure with deuterium at T =
250?K.
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Amplitude of E signals (channels)
POO +00 600, 600
. ,,
0,2 0,4 0,6 0,6
Energy, MeV
Fig. 7. Two-dimensional distribution (Aft AE)
for one neutron detector'. The arrow indicates
the position corresponding to the limit of the
spectrum for a neutron energy En = 2.5 MeV from
reaction (22a).
The temperature range investigated wa divided into two overlapping subranges:-. in the
subrange from -160 to -20?C we used a plastic scintillator in counters 6 and 7 (see Fig. 2;
experiment 2); in the subrange from -60 to +110?C we used a CsI(Tl) scintillator (experi-
ment 2). In each experiment we carried out five main exposures with deuterium, differing
as to the target temperature. The temperature was changed in steps of 30-40?.
The neutron background was determined in exposures with a target filled with helium to
a pressure equivalent to the deuterium pressure in the main exposures as far as the number..
of muon stoppings was concerned. The background measurements were carried out at a different
temperature. The contribution of the neutron background was a maximum (20%) at the lowest
temperature.
For each exposure with deuterium the time distributions of neutrons was analyzed with
the aid of the expression
dN? Nli~n%dd?_Xdd 1ll
=t - n dd (oxp(- -(a'O+Xdd?(Q+?4 )t]/, (23)
2 (1dd?(P'+ ' J?' )
which was derived with allowance for the regeneration of muons in reaction (22).*, Here, Nu
*In the derivation of Eq. (23)-we assumed that the coefficient of "sticking" of a muon to
helium in reaction c)dd? -' 'He It + it (22) d d dd? -+ 3He + n + It is zero. In the general case the expression
for dNn/dt has the form A[exp(-y,t) - exp(-y2t)], where
1 dd dd _. o x o d Addu(Tm), where in accordance with [36]
Xdd? (T) _ X1+ 12T + aT' 312 exP (- ea?lkT ).
In this expression the terms al and A2 make allowance for the contribution of nonresonant
mechanisms to the formation of ddp molecules. For absolute normalization, the values of Addy
measured in [5, 9, 21] were incorporated into the analysis.
The results of analysis of the experimental data are presented in Fig. 8. It can be
seen that our results are in good agreement with predictions based on a consideration of the
resonant mechanism of the formation of ddy molecules and they confirm the existence of this
mechanism. For the position of the resonance we found the value
s$? = (0.050 f 0,003) ev ,
while for the extreme value of the rate of formation
Addu (83?) = (0.85 f 0.11) ? 100 sec 1
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As we see from Eq. (25) the experimental accuracy of the determination of cad was 3 May.
The other terms (Ev and Eo) of Eq. (20) can be calculated with roughly the same or better
accuracy. It thus follows that the investigation of the characteristics of the resonant
formation of ddu molecules makes it possible for the energy of the level with L = v = 1 to
be determined precisely. It is important to calculate echo with an error -1 MeV, which
would make it possible, by comparing the calculations with experiment, to find the contribu-
tions to that energy from interactions, in particular vacuum polarization, which are in addi-
tion to the Coulomb interaction. According to [40], the expected value of this contribution
is 10 MeV.
The results of measurements of Addu [5, 9, 20, 211 were analyzed by us in (20] taking
into account the existence of one resonance whose position was determined from Eq. (25). As
can be seen from Fig. 8, in this case it was possible.to get a fair description of all the
available experimental data for Addu? In actual fact, however, because of the presence of
hyperfine splitting [37] of the ground.state of the du atom and of the ddp state with L = v =
1 instead of one resonance.condition (20) it is necessary to account of several such rela-
tions which correspond to transitions. from states with Fdu = 3/2, 1/2 to states of the ddu
molecule with a total spin Sddu = 3/2, 1/2. In accordance with this, instead of one reso-
nant dependence Area (T) it is necessary to consider several functions Ares,i(T), each of
ddu ddu
which has the form of Eq. (21) and differs from others by the position of the resonance (To).
The contribution of each of these components to the total value of Addu is determined by the
population of the spin states of the du atoms at the time of the formation of the ddu mole-
cules and, hence, Addu should depend not only on the temperature but also on the deuterium
density. This effect was observed in the experiment of [23] which was carried out at T =
34?K and a deuterium pressure of 1.57, 3.14, and 6.28 MPa. Taking the time dependence of
the neutron yield from reaction (22a), Breunlichet al. [23] obtained the value (18) for the rate
of the transition Fd1j= 3/2 -> Fdu= 1/2. The experiment of [23] has not yet been completed
and Breunlich et al. propose also to measure the temperature dependence Xddu(T).
As for other work done recently, we should mention the experiment of [38]. The use of
an original procedure (pulsed ionization chamber) enabled Balin et al. to determine the
coefficient for sticking of a muon to helium in the reaction (22a) with good accuracy and to
obtain the relation
(Odd = dd? - 31je[t+ n _ 0.14 ? 0.01.
dd? -r 3He -}- n ~- ?
Analysis of all results now available.from investigations [5, 9, 20, 21, 23] of the process
by which the system is formed permits the conclusion that these data support the theory of
resonant formation of ddp molecules. For reliable determination of all the parameters that
specify the resonant formation of ddu molecules it is important to measure the temperature
dependence of Addu for different values of the deuterium density.
Investigation of u Catalysis of the Fusion of Deuterium and Tritium Nuclei. Experimen-
tal confirmation [20] of the conclusions of the theory [14, 35, 36] concerning the resonant
formation of ddp molecules was a serious argument in favor of the predictions [35, 36] that
an analogous, but much more intense (by 2-3 orders of magnitude) resonance exists in the
cross section for the formation of dtu molecules. The scheme of the processes initiated by
negative muons in a D2 + T2 mixture is given in Fig. 9. Once the dtp system is formed the
fusion reaction
Oft 2'F> (Me+n+it+17.bMeV
t 'He ? + n ?17.6 MeV
(27a)
(27b)
occurs in it rapidly (1012 sec-' [41]); in one channel of this reaction the muon is liberated
while in the other it is "stuck" to helium.
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-wq
A.a
Fig. 9. Scheme of processes initiated by
negative muons in D2 + T2.
2 J 4 S B '7 iv E B 9
Fig. 10. Experiment [48] to detect and inves-
tigate p catalysis of the d + t reaction: 1)
muon channel; 2, 3, 4, 7) scintillation muon
detectors; 8) gas target; 9) vacuum jacket;
1N-4N) neutron detectors (NE-213, volume 109
cm9); El-E8) scintillation electron counters.
Muonic catalysis of the reaction (27) and its possible use for practical purposes [42]
have.been discussed in detail in a number of reviews [43-46]. According to the theory it is
expected that one muon in a deuterium-tritium mixture can initiate -100 cycles of the reac-
tion (27) and thus release an energy of -2 GeV. In order to ascertain whether effective
(multiple) p catalysis of the d + t reaction is possible it is necessary experimentally to
determine the parameters of many p-atomic and u-molecular processes [45]. In this problem,
however, it is possible to isolate three main conditions: The 'formation rate of dtu mole-
cules (Adtu) and the rate of muon capture from deuterium by tritium (Adt) should be suffi-
ciently high (aatu, Xdt > 108 sec-') while the probability of a muon "sticking" to helium, on
the other hand, should be low (ws < 0.01). Theoretically, all of these conditions should be
satisfied [14, 35, 47].
Experimentally, p-atomic and p -molecular effects in a deuterium-tritium mixture (as
well as in pure tirtium) had not been studied at all until 1978 (the beginning of the experi-
ments of our group [48]). The aim of the measurements [48] was to determine Xdtu and Adt and
to try to find the parameters of Adtu(T). To this end we decided to use.a high-pressure gas
target (relative gas density W =0.1), filled with a D2 + T2 mixture at different relative
deuterium and tritium contents.
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The experimental arrangement for the investigation of catalysis of the d + t reaction
is shown in Fig. 10. Its main part consisted of a gas target and neutron detectors (1N-4N).
The process (27) was identified by'detecting neutrons with an-,.energy En m 14 MeV, correlated
with stoppings of muons in the target.
Two major problems had to be solved because a large quantity of radioactive tritium
(5000 Ci; 1 Ci - 3.700.1010 Bq) was used in the experiments. One of them involved the ob-
servance of special, more stringent safety measures to ensure that the target remained leak-
proof. The second problem was that of suppressing the intense background (neutrons from
capture in iron and random coincidences) owing to the large number of muon stoppings in the
target walls, this number being 2-3 orders of magnitude larger than the number of stoppings
in the gas. Since they can be "irradiated" by R rays from the tritium, scintillators in-
side the target to record stoppings in the gas cannot be used in order to suppress that back-
ground. The following measures were undertaken to discriminate the background:
a) A logic of ternary delayed u-n-e coincidences was used, i.e., it was required that
in the 10 psec following a muon stopping there at first be a signal of detection of a neutron
from the reaction (27), followed by detection of an electron from a V decay;
b) only those events associated with the detection of neutrons 1 usec after a muon
stopping [muon lifetime in iron T11 (Fe) = 0.2 usec] were selected;
c) neutrons and y rays (electrons) were separated according to the shape of the light
pulse in the scintillator of the neutron detector.
For the neutron detection we employed detectors with an original design with an NE-213
scintillator (volume 1 liter). For each detector we measured the recording time of an event
and the amplitude of two pulses, one of which (Af) was proportional to the intensity of the
"fast" part of the light pulse while the other (As) was proportional to the intensity of its
"slow" component. Joint analysis of the amplitudes Af and As permitted separation of the
neutrons from the y rays and electrons. Figure 11 shows a two-dimensional distribution (Aft
As) measured in an exposure with a'D2 + T2 mixture. An idea of the clearcut.identification
of the process (27) under investigation can be had from Fig. 11.
The electron detectors E1-E8 (see Fig. 10) consisted of four telescopes, used in "fast"
NE anticoincidences during detection of neutron events and in delayed p-n'--e coincidences during
detection of an electron from a p.decay, following a neutron from reaction (27).
During the experiment we carried out 14 exposures, which differed as to the temperature
of the D2 + T2 gas mixture or the deuterium (CD) contents in them. The relative tritium
content varied from 0.8 to 3%. The temperature range investigated was 93-613 *K. For control
and normalization we also carried out measurements with pure deuterium and with an evacuated
target.
The experimental data were analyzed using expressions. for the yield and time distribu-
tion of neutrons from the d + t reaction obtained in [49], according to which at a mixture
pressure of 0.49-9.81 MPa and a tritium concentration CT 5:0.5 the yield of reaction (27) is
Yn __ ?dtpCDCT (fin-f-? dt) (28)
21o ()o+XdtCT+7-dt1,Cn)
dd n T+CDCTCf [a'dt exp (- 11t) +CD (2'dt?CD--~,dt) exp
atCT+)1dt? 1J
X1 ' X0, %2 = a'0 + a'dtCT T %dt,,CD, %dtu = %dt?(P, 21dt = Xdtcp?
During the preliminary analysis we compared the values YeXp = N IN , where N and N are
n n e n e
the numbers of neutrons and electrons, respectively, which had been obtained for the experi-
mental yield of the reaction (27) with different exposures.
In this case we establish that:
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.M -7
192
499
o
64 s-z; a3
0 64 128 192 2S6
Amplitude of fast-component signal
Fig. 11. Two-dimensional distribution (A f, As)
for .a neutron detector measured in an exposure
with D2 + T2.
a) The values of fxp remain practically constant in the temperature range 93-613 ;
b) the relative neutron yield does not change when only the deuterium content in the
D2 + T2 mixture is varied and, on the other hand, varies roughly in proportion to the tri-
tium content (CT) in the mixture. As follows from Eq. (28), such laws in the variation of
the neutron yield could be explained by the fact that under our experimental conditions
AdtUcpCD >> Xo. The nature of the time distributions of neutrons is also consistent with
that. expected [49:] for a high value of AdtU. For each exposure such a distribution, in
accordance with Eq. (29), can be represented as the sum of two exponential functions with
exponents (-alt) and -alt, where Al ` Xo and A2 >> AO-
The complete analysis consisted of comparing the experimental time distributions of
neutrons with an expression of the form (29), using the method of least squares. For abso-
lute normalization we used the calculated value of the detection efficiency for 14-MeV neu-
trons. As a result of the analysis we found (at a 90% confidence level) the values of the
rate of stripping of a muon from deuterium by tritium
2Ldt = (2.9 f 0.4) ? 10a sec -I
(30)
and the formation rate of dtp molecules
~dt4 >108Secl. (31)
The value of aat given-in Eq. (30) is in good agreement with those calculated in [14]..
For 1dtp we obtained only the lower limit (31). This is explained, in accordance with Eq.
(28), by the insensitivity of the neutron yield from reaction (27) to Adta for higher values
A 0 C>> A o.
U
D The large value found for AdtU substantially exceeds the known formation rates of other
muonic molecules (ppp, pdp, and ddU) and is in agreement with calculations [36] carried out
on the basis of a consideration of the resonant mechanism of dtj formation. Thus, the exis-
tence of this mechanism, which we had earlier established for ddp molecules, is also confirmed
for dtp molecules.
On the basis of the resonant nature of.the process of formation of the dtp system, one
would expect a .considerable variation in the yield of neutrons from reaction (27) as a func-
tion of the temperature of the D2 + T2 mixture. The measured neutron yield, however, in fact
does not vary (within the limits of the measuring error of -10%) in the range 93-613?K
investigated.
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This fact can be explained as follows.
1. Even within the framework of a simple resonance dependence of the form (21) the value
of adtu, which varies with the temperature, can remain fairly large so that at all tempera-
tures in the range 93-613?K we have aatuTCD >> Ao. In this case (and with the condition that
AdtWCT a Ao) the yield of neutrons from reaction (27) is insensitive to Adt according to
u
Eq. (28).
2. Muonic to atoms, obtained in the stripping process du + t - to + d (the dominant
channel of their formation in a D2 + T2 mixture when CT 30 bar - this paper; p < 30 bar - [4].
have explained the effect as the action of collision's of the molecules on their translational
motion. At the same time it has been shown for a broad class of condensed and gaseous
materials that as the overall mobility of the particles scattering the neutrons decreases,
the degree of slowing down of their motion increases, and the effective mass increases the
total scattering cross section of cold neutrons decreases by virtue of the inelastic compo-
nent [4, 9]. And so as long as the gas pressure is so low that the molecules on the average
are.located at distances greater.than the effective action radius of intermolecular forces
and the neutron wavelength, and molecular collision events are rare relative to the inter-
action time of:cold neutrons with molecules, the scattering of neutrons occurs primarily the
same as in a gas of free-particles. As the gas pressure increases, the collision frequency
of molecules increases-and the average distance between them decreases. A neutron interacts
now with molecules. which are.to this or that extent under the influence of their nearest
neighbors, which increases the effective mass of the scattering particle. Upon a change in
pressure from 4 to 132 bar the distance between nitrogen molecules varies from 2.2 to 0.65
nm, and the time between their collisions varies from 30 to 2 psec. The time for a neutron
with wavelength 1 nm to traverse the region occupied by a nitrogen molecule is -1 psec. It
is natural to assume that it is precisely intermolecular collisions and interaction which
determine the observed dependence of the scattering cross section on gas pressure. Estimat-
ing according to the relationship mentioned [9] the variation of the effective mass of the gas
particles upon a change in pressure on the basis of the cross sections measured in this paper,
we obtain that the mass increases by 10 and 50% in the 30-130 bar range for neutrons with
wavelengths of 0.7 and 1.7, respectively.
Thus the data obtained in this paper on the scattering cross section of cold neutrons
in nitrogen have for the first time permitted making a reliable confirmation of the existing
computational results within the framework of ideal gas models in the energy range up to
approximately l0-4 eV. A dependence of the microscopic scattering cross section of neutrons
on gas pressure has been discovered which is qualitatively explained by the effect of inter-
molecular collisions. The possibility of experimental confirmation of dynamic scattering
models which would take account of intermolecular interaction in the gas is shown.
LITERATURE CITED
1.
J.
Ljvseth, Kjeller Report, No. 26 (1962).
2.
E.
Fermi and L. Marshall, Phys. Rev., 75, 578 (1949).
3.
V.
P. Vertebnyi et al. At. Energ., 18, No. 5, 452 (1965).
4.
V.
E. Zhitarev, Candidate's Dissertation, MIFI, Moscow (1980).
5.
A.
Fulinski, Acta Phys. Polon., 34, No. 1,. 119 (1968).
6.
S.
B. Stepanov et al., in: Neutron Physics [in Russian], Izd. FEI, Obninsk (1974),
p.
257.
7.
A.
A. Vasserman et al., Thermophysical Properties of Air and Its Constituents [in
Russian], Nauka, Moscow (1966).
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 Declassified and Approved For Release 2013/02/21: CIA-RDP10-02196R000300030004-5
8. Neutron Cross Sections, Vol. 1, Ses. ed. Supp. No. 2, BNL-325 (1964).
9. E. Yanik and A. Koval'skaya, in: The Scattering of Thermal Neutrons [in Russian],
Atomizdat, Moscow (1970), p. 377.
10. Y. Lefevre et al., in:. Neutron Inel. Scatt., Vienna (1972), p. 445.
COMPOSITION OF THE GASEOUS PHASE AND THE
BEHAVIOR OF XENON AND KRYPTON IN IRRADIATED
FUEL ELEMENTS OF A BOR-60 REACTOR
A. P. Kirillovich, Yu. I. Pimonov,
Yu. G. Lavrinovich, and O..S. Boiko UDC 621.039.546
The gaseous phase in irradiated fuel elements is a complex multicomponent system con-
sisting of a filler gas (He, Ar, and others), gaseous fission products (GFP), and nonfrag-
mentary (process) gases desorbed from the fuel [1]. The investigation of the amount and
composition of this phase has great meaning for raising the efficiency of the fuel elements
and ensuring the safety of the reactor as well as for accomplishing the subsequent regenera-
tion of the depleted fuel. Only a few publications [2-4] are devoted to the investigation of
the composition of the gaseous phase in irradiated fuel elements and to the behavior of GFP
at different stages of the fuel cycle, which is evidently associated with systematic and
technological difficulties.
The aim of this article is to develop a procedure for the sampling and mass spectro-
metric analysis of the chemical and isotopic composition of the gaseous phase in irradiated
fuel elements as well as to investigate the composition of the gases in regular and experi-
mental heat-generating assemblies (HGA) of a BOR-60 reactor and the behavior of Kr and Xe in
the stage of thermal opening (fusing) of the fuel element casing.
Procedure of the Investigation. In working out a procedure of quantitative mass
spectrometric analysis of complex gas mixtures the coefficients of relative,sensitivity of
the gas constituents being analyzed and the intake pressure. region in which these coefficients
remain constant were determined, and operating regimes of the mass spectrometer-ion.-source
which provide maximum sensitivity to the gases being determined were selected.
Fig. 1. Intake system: 1) sorption trap;
2) vacuum pump; 3) intake tank; 4) mixer;
5) valve; 6) sampler ampul; 7) manometer;
IS). to the ion source; HV) to the high
vacuum.
Translated from Atomnaya Energiya, Vol. 55, No. 6, pp. 405-407, December, 1983.
Original article submitted April 18, 1983.
0038-531X/83/5506- 0865$07.50 ? 1984 Plenum Publishing Corporation
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 Declassified and Approved For Release 2013/02/21: CIA-RDP10-02196R000300030004-5
Fig. 2. Layout of the sampler: 1) fuel
element;' 2)'mechanism for puncturing the
fuel elements; 3) needle; 4) apparatus
for thermal opening (fusing) of the casing;
5) filter; 6) absorbing column; 7) vacuum
pump; 8) valve;-9) sampling ampul;. 10)
manometer.
TABLE 1. Main Characteristics of the
Investigated Fuel Elements
Characteristics and
parameters of HGA
irradiation
Regular HGA
Exptl. HGA
Fuel comp.
U02 (90% enrich-
80% U02 (90% en-
ment in 235U)
richment in 235U) +
Av, depletion of
11
20% Pu02
heavy atom, %
Holding time after
4
7
irradiation, months
o. of
fuel e1e-
monts
HGA
-
Amt. of
cm3/
gas
1
- Content, vol. %
inveeti.
gated
,;
,
fuel ele-
ment
I
He He
I N2
I Oa
I Ar
I CO2
Kr
Xc
5
Regular
93?14
0,09?0,0104,30?0,26
0,16?0,02
0,040?0,003
0,040?0,003
0,009?0,002
14,0?0
3
81
4?0
8
4
Exptl.
257?8
0,07?0,05 2,6?1,3
1,15?0,69
0,29?,0,21
0,46?0,23
0,054?0,010
,
12,5?0,2
,
,
82,8?0,8
TABLE 2. Amount and Composition of the Gaseous Phase under the Casing of Some Fuel
Elements of a'BOR-60 Reactor.
TABLE 3. Release of Xe and Kr from the
Fuel upon Its Heating and the Fusing of
the Fuel Element Casing, cm3/fuel elements
Temperature in
fusing furnace, ?c
Regular HGA
25-900 0,1 0,325
900-1300 2,56 7,69
2300-1650 1,5 8,36
Experimental HGA
25-800 0,011 0,77
800-900 0,0094 0,048
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TABLE 4. Yield of Xe and Kr from the Irradiated Uranium and Uranium-Plutonium Fuel
of a.BOR-60 Reactor
Puncture of the fuel
element casing
Fusing of the fuel element
casing
Total measured amt. of gas
U02 I U62-Pu02
0,059?0,008
81,9
0,013?0,002
18,1
0,1)72?0,010
100
0,1141=E:0,0041
98,5
0,0117?0,0003
1,45
0,1158:f:0,0044
100
0,0064?0,0009
76,2
0,0020?0,0004
23,8
0,0084?0,0013
100
0,0110?0,0004
99.,1
0,00010?0,00(102
0,9
0,01110?0,00042
100
Remark: The yield of the elements, g/g of the depleted fuel, is indicated in the
numerator, and the percentage of the total measured amount of Xe and Kr is indicated
in the denominator.
The measurements were made with an MI-1201 mass spectrometer additionally equipped with
a gas intake system (Fig. 1) according to the procedure of [5]. We used the method of com-
paring the sample being analyzed with a control gas mixture prepared from certified gases,
i.e., gases which have been additionally checked for purity (the instrument was not calibrated
for tritium, since the sensitivity of the mass spectrometer to hydrogen - protium, deuterium,
and tritium - is identical [6]), in investigating the composition of the gaseous phase. In
order to raise the sensitivity when determining a small amount of tritium (T2) and hydrogen,
an ampul with the gas mixture to be analyzed was connected directly to the ion source through
a membrane with a 50-um hole and cooled to the boiling temperature of liquid nitrogen.. The
ion source operated in the maximum sensitivity regime of the instrument for hydrogen (10'lo g
(10' %)). No special measures were adopted to concentrate tritiated water (T20) and other..-
tritium compounds. The relative mean square deviation in the determination of Xe and Kr was
1 and 2%, respectively, and it did not exceed 15% for the remaining constituents of the gas
mixture (He, Ar, N2, 02, CO, and C02) when their volume content ranged from 0.01 to 100%.
Sampling of the gases from under the casings of irradiated fuel elements and from the
apparatus for thermal opening of the fuel element casing by fusing was done with the help of
a sampler system (Fig.. 2) located in a protective chamber. The ampuls and sampler system
were put under a vacuum in advance to a pressure of 6.6 and 13.3 Pa and checked for inleakage.
The residual background in nitrogen and oxygen in the ampuls did not exceed 1 x 10-13 and
5 x 10-14 A, respectively. Calculation of the amount of gas released from the fuel under the
fuel casing was done from the measured pressure (the accuracy class of the instrument was
1.5) and the known volume of the sampling system (see [5] for more detail).
Results of the Experiments and Discussion of Them. The characteristics and parameters
of the irradiation of the investigated fuel elements of regular and experimental HGA of a
BOR-60 reactor are given in Table 1, and the results of measurement of the amount and compo-
sition of the gaseous phase are given in Table 2. As follows from the latter, Kr and Xe,
whose average content varied from 8 to 14 and from 81.4 to 89.59% for various types of fuel
elements, are the main constituents of the gaseous phase of irradiated fuel elements with
uranium oxide and uranium-plutonium fuel. In addition to Kr and Xe, H2, He, N21 CO, 02, Ar,
and CO2 were detected under the casing of fuel elements. The presence of "nonfragmentary"
gases can be explained by the fabrication technology of the fuel and fuel elements and by
other factors (for example, by radiolysis of moisture, release or absorption of oxygen as a
result of the chemical interaction of the fuel with. constituents of the gaseous phase and the
casing, and so on).
Elemental tritium (T2) has not been detected in the gaseous phase of irradiated fuel
elements of regular and experimental HGA at the maximum sensitivity of the method. The ob-
served variations in the amount of gas (here and later the measurement data are given in cm3
at normal conditions) in individual fuel elements of regular HGA (from 77 to 112 cm3) exceed
the error in determining the volume of released gas and can be explained by small differences
in the fuel charging and in the conditions of fabrication of the individual fuel elements and
their irradiation in the reactor. I t turned out (see Table 2) that the amount of gas re-
leased from uranium-plutonium fuel under the fuel element casing exceeds by approximately a
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 Declassified and Approved For Release 2013/02/21: CIA-RDP10-02196ROO0300030004-5
TABLE 5. Isotopic Composition of Xe and Kr in the Fuel Elements of the BOR-60
Reactor
Regular
Exptl.
1.4,2?0,2
15,0?0,2
22,8?0,4
22,8+0,4
34,8?0,4
33,8?0,4
28,2:1-0 ,4
28,4?0,4
1.5,4?0,2
15,9=1:0,2
28,4?0,4
28, 3?0, 4
49,83=b0,3
40,17?0,30
6,57+0,13
6,63-1-.0,13
factor of two the volume of gases in irradiated fuel elements with uranium fuel. This may be
a consequence of a larger fuel charge in the fuel e]r'ements and (or) a higher depletion of
the uranium-plutonium fuel and specific linear loading, during irradiation of the HGA in the
reactor (see Table 1).
We investigated the behavior of Xe and Kr upon heating of the fuel elements to 700?C
in the puncturing mechanism (see Fig. 2) and upon thermal fusing of their casings. It was
established that when the fuel is heated part of the Xe and Kr is additionally released from
it. This release is described by the following empirical equations for uranium-plutonium
fuel in the 20-700?C temperature ranger
v$e = 212.8 + 0,0010t;
vKr=32.12+O.00009t,
where t is the temperature, ?C, and v is the amount of gas (Xe, Kr), cm9/fuel element.
Data on the release of Xe and Kr upon the heating of fuel elements in the thermal opening
apparatus (see Fig. 2) are presented in Table 3. The total amount of Xe and Kr measured in
the gaseous phase under the casing of fuel elements and when it is fused is given in Table 4.
The relationship of Xe and Kr is not identical for the different types of fuel elements, which
indicates a difference in the nature of the release of these GFP from the fuel both in the
course of its irradiation in the reactor (which is evidently associated with the different
irradiation regimes) and in the subsequent heating of the irradiated fuel elements and fusing
of the casing. Thus for uranium-plutonium fuel the Xe and Kr yield from under the casing were
98.5 and 99.1%, respectively, of the total measured amount, whereas for fuel elements with
uranium fuel it is equal to 81.9 and 76.2% (see Table 4). Part of the Xe and Kr remains in
the fuel after fusing of the fuel element casing (1650?C) and may be released from it during
chemical regeneration, for example.
A procedure is being developed at the present time for measurement of the amount of Xe
and Kr released during chemical regeneration of depleted fuel which will permit accomplishing
a balance in the GFP. The isotopic composition of xenon and krypton in the investigated fuel
elements with UO2 and U02-PU02 fuel irradiated in a BOR-60'reactor differs inappreciably
(Table 5).
The experimental data obtained on the composition of the gaseous phase in irradiated
fuel elements of a BOR-60 reactor with uranium and uranium-plutonium fuel and on the behavior
of Xe at the stage of preparation of fuel elements for regeneration can be used for physical
and engineering calculations in the design and operation of, nuclear and radiochemical fuel
regeneration facilities.
LITERATURE CITED
1. B. V. Samsonov and V. Sh. Sulaberidze, Gas Release from Oxidized Nuclear Fuel. Analytical
Review NIIAR V-16 [in Russian], Dimitrovgrad (1977).
2. A. T. Ageenkov et al., At. Energ., 40, No. 3, 203 (1976).
3. A. T. Ageenkov et al., At. Energ., Al, No. 1, 23 (1976).
4. A. T. Ageenkov and E. M. Valuev, At. Energ., Al, No. 2, 141 (1976).
5. A. P. Kirillovich et al., Preprint NIIAR-31 (439), Dimitrovgrad (1980), p. 27.
6. M. Goldblatt and W. Gones, Anal. Chem. 36, 431 (1964).
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 Declassified and Approved For Release 2013/02/21: CIA-RDP10-02196R000300030004-5
BASICS OF PULSED NEUTRON LOGGING WHEN STRONG
ABSORBERS ARE ESTABLISHED
D. K. Galimbekov, I. T. Ilamanova,
B. E. Lukhminskii, and A. I. Pshenichnyuk UDC 550.83
The space-time distribution of neutrons and y radiation from radiative capture has been
theoretically treated in [1, 2] in relation to problems of pulsed neutron logging. Some use-
ful qualitative conclusions in the theory of pulsed neutron logging have been drawn in [2] on
the basis of numerical results obtained for a homogeneous infinite medium with a uniformly
distributed absorber. Particularly, the authors of [3] have reported on experimental work in
which the possibilities of various techniques of pulsed neutron logging were evaluated for
establishing strong absorbers of neutrons. The investigations have confirmed that the pulse
techniques are more sensitive than the stationary techniques as far as the concentration of
neutron absorbers is concerned.
But a number of theoretical problems have not been solved: the possibilities of pulsed
neutron logging techniques with short delay times; precise quantitative estimates of the in-
fluence which imply changes in the borehole diameter and the position of the instrument in the
borehole; influence of the spectral composition of the y radiation from radiative capture
upon the readings of pulsed neutron y logging; and the possibilities of techniques based on
measurements of emission spectra.
These problems were solved with the Monte Carlo method. The features of the program are
as follows: the geometry of the problem assumes a water-filled borehole; the.pulsed neutron-
logging instrument comprises a neutron generator (energy 14.1 MeV); the ore layer is modeled
by sand (SiO2) containing cinnabar (HgS; isotropic elastic scattering, radiative capture, and
resonance capture of fast and epithermal neutrons and radiative capture in the thermal range
were taken into account; the program portion modeling the transfer and the recording of y
quanta was taken from [4].
The analysis of the results of the calculations made it possible to assess the time in
which the nonstationary distribution of the thermal neutrons and of the y radiation from
radiative capture becomes exponential, i.e., when the space-time dependencies are locally
multiplicative. In the case of a pulsed neutron logging instrument with a long probe (thick-
ness of the shield between the target of the generator and the detector) of 5-30 cm, the
distribution of thermal neutrons is locally multiplicative in logging work in boreholes with
a diameter of 59-76 mm, provided that the delay time after a pulse exceeds 300-400 usec for
the enclosing rock, and in practically the entire time interval in the case of rock with a
cinnabar volume concentration q = 1%.
Recommendations can be derived from the calculations for selecting the measurement con-
ditions in pulsed neutron-neutron logging work when neutrons are recorded in..bhree time
"windows" with center positions of 150, 350, and 800 usec. The neutron fluxes which can be
recorded in the windows then fall on the exponential portions which correspond to rock rich
in mercury, poor in mercury, or enclosing rock. The window width required for reaching the
necessary statistical accuracy is experimentally determined. The dependgnce of the attenua-
tion decrement An, measured on the exponential portion, upon the lifetime rn of the thermal
neutrons in the layer (or upon the cinnabar concentration) is shown in Fig. 1. It follows
from the figure, which may serve as a nomogram for determining the cinnabar concentration from
An, that in a borehole with a diameter of 59 mm, the differentiation by way of the decrement
for rocks with a cinnabar concentration between 0 and 1% amounts to 10 and does not depend
upon the position of the instrument in the borehole. In a borehole with a diameter of 76 mm,
the differentiation on the basis of A decreases to 5 and thereafter drops off sharply with
increasing borehole diameter. n
Translated from Atomnaya Energiya, Vol. 55, No. 6, pp. 407-409, December, 1983. Original
article submitted May 16, 1983.
0038-531X/83/5506-0869$07.50 ? 1984 Plenum Publishing Corporation
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 Declassified and Approved For Release 2013/02/21: CIA-RDP10-02196R000300030004-5
A , msoc'1
0 10
0 0,25 0,5 0,751
Fig. 1. Theoretical chart for interpreting
pulsed neutron-neutron logging in boreholes
(diameter of the instrument 42 mm); o refers
to a borehole diameter of 76 mm (instrument
in the center); 0 and A) 59 mm (instrument
in the center and at the wall, respectively);
- - - - -) dependence for an infinite homo-
geneous medium.
0 0,2- 0,' 0, 5 48 q,%
Fig. 2. Dependence of the flux 0(E, q)
( ) and of x(- - -) upon the volume con-
centration of cinnabar.
The field of the y radiation from radiative capture is basically given by the neutron
field. But the field is characterized by spectral features resulting from the material com-
position of the rock. An analysis of the dependency of the distribution has shown that the
recommendations made for the measurement conditions in pulsed neutron-neutron logging are
applicable to pulsed neutron y logging with integral recording of the y radiation from radia-
tive capture. The rock differentiation through the attenuation decrement ay is 12-13, i.e.,
it is higher than in pulsed neutron-neutron logging because the borehole has a lesser in-
fluence upon the readings of pulsed neutron y logging. The rock differentiation through the
flux 0(q) of the y radiation measured with a small delay time (e.g., 0-25 usec) is the same
as in the ay measurements. Though the amplitude characteristic of 0y is less affected by
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 Declassified and Approved For Release 2013/02/21: CIA-RDP10-02196R000300030004-5
disturbances in the case of short. times than the attenuation decrement (the influence of a
hydrogen-containing layer, changes in borehole diameter, instability of the neutron yield),
the spectral version of the technique has certain advantages over measurements of the decre-
ment: higher sensitivityto,?the absorber concentration and absence of the disturbing back-
ground of the natural radioactivity of:rocks.
Figure 2 depicts.the dependence of the flux @y (E, q) of y radiation from radiative
capture (flux normalized to; the flux 01(E, 0) in the enclosing rock) in the time interval
0-25 psec for various energy intervals (see numbers at the curves). The highest sensitivity
to changes in q is observed at 4-5 MeV. The low sensitivity at 5-10 MeV is explained by the
missing mercury emission in this interval. Figure 2 includes the quantity
U , (E, q) Te (0)
my (E, 0) T?e(q)
for 2-5 MeV. The quantities Te(0) and Te(q) are inverse quantities of the decrement A for
the enclosing. rock and'the ore, respectively. The quantity X varies by more than two orders
of magnitude in the energy interval under consideration.
Thus, based on.the results of the calculations, the laws governing the time distribu-
tion of the thermal neutron field and the Y radiation field from radiative capture were
determined for short delay times. The influence which changes in the borehole diameter and
the eccentricity of the instrument have upon the readings of pulsed neutron logging has been
determined with great precision. It has been shown that the pulsed neutron y logging tech-
nique with short delay times is promising. Furthermore, it has been shown that it is possible
to increase the efficiency of pulsed neutron y logging when the spectral version of the
technique is employed.
LITERATURE CITED
1.
A.
L.
Polyachenko and N. S. Erichman (Erikhman), Fiz. Zemli, No..8, 106 (1971).
2.
V.
F.
Zakharchenko, Izv. Akad. Nauk SSSR, Ser. Geofiz., No. 10, 1522 (1963).
3.
D.
K.
Galimbekov, in: Mathematical Modeling in Nuclear Geophysics [in Russian], Bashkir
Branch, Academy of Sciences of the USSR, Ufa (1979), p. 66.
4.
Nuclear Physics Techniques of Elementary Analysis and of Geophysical Sounding [in
Russian], Nedra, Moscow (1972).
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 Declassified and Approved For Release 2013/02/21: CIA-RDP10-02196R000300030004-5
NUCLEAR-PETROPHYSICAL BASICS OF NEUTRON
MEASUREMENTS IN ROCKS
A. I. Pshenichnyuk UDC 539.125.523:550.832.5
The modern orientation of nuclear geophysics toward indirect measurements of neutron
characteristics.of rocks [1] requires the compilation and analysis of a set of characteristics
for solving the corresponding direct and inverse problems. Since the number of parameters in
the set coincides with the number of independent characteristics of the radiation field, the
set determines the maximum information content of the technique of the measurements. Numeri-
cal calculations of neutron characteristics are important for the nuclear-petrophysical founda-
tions of research in geophysics..
The high requirements to the solution of the problem manifest themselves through the
measurement conditions: wide variability range of the hydrogen concentration; high neutron
energy (14 MeV) at which inelastic scattering, angular anisotropy, and neutron absorption in
reactions with emission of charged particles take place; and measurements at distances from
the source which are characterized by a flux attenuation of 3-4 orders of magnitude so that
the angular distribution of the radiation must be taken into account.
These requirements are met by the:method developed in [2,.3] for solving the transfer
equation. The method is based on a modified transport approximation and the technique of
spectral approximations.. The collision density of neutrons provided by a two-dimensional
isotropic monoenergetic source in a medium of some composition is of the form
70(s, a
h (o) arctg xX(0) (x, u) f g (x, u')
2s J s? (0) oxp r ixz -
h (u) M (x, u) J M du' dx
(x, u')
(the notation is the same as in [2, 3]). Let us approximate the transformation function in
Eq. (1) by the formula
t (x, u)=A (u) oxp [-B (u) 1+C (u) xa]
1~1 1-c (u) x2
which preserves the essential features of a precise transformation and which allows an analytic
treatment. The free parameters A, B, and C are obtained from the condition that the first
three spatial moments coincide. Then
'F (Z, u)= 1 'Y (u)oa(u) ( a(u)+1 11/2 K, ra (u) ~1+ a (u)+1 Zz I ( )
A 2t (u) / ( a2 (u) 2t (u) 2
a (u) 1=2 [1/ 3 Ex(u)-
J-t ;
;
Et (u)- Eta (u)/T2 (u). (4)
The notation is interpreted as follows: Ko(x), a Bessel function; `Y(u), neutron spec-
trum; r(u), age of the neutrons; and Ex(u), excess of the distribution of Eq. (1). The
Translated from Atomnaya Energiya, Vol. 55, No. 6, pp. 409-410, December, 1983.
Original article submitted May 30, 1983.
872 0038-531X/83/5506- 0872$07.50 ? 1984 Plenum Publishing Corporation
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 Declassified and Approved For Release 2013/02/21: CIA-RDP10-02196R000300030004-5
spectrum, the age, and T(u) are defined by the formulas
V(u)= h(0) expr- g(udu'1
M (u) l bl M (rc ) J
u
3-r (u)=,%'(0)+ 5 h(u')12 (u') [1-9(u')0(u')) M(uu") +Is(u)[1-h (u)0(u));
0
It
5; (u) = 8 1', (01+ 5 h (u') 14 (u') X
X { [1-g (u') 6 (u')) [! }- 9 -y h (u') g (u'1 M (u,)) M (u') --
+11(u){[1+h(u)0(u)) [18 I- 8 h(u)0Cu)]-9 h9 (u) M(u)
where we have introduced, for the sake of brevity, the notation
1' (u) (u) 0(u)=M-'(u) t 9i (u)+2g (u) 01(u))?
1-h(u) po(u)
(6)
The structure of Eq. (2) resembles that of the formula obtained'in the P2 approximation
of the theory of continuous moderation [4]. The parameter M(u) is given by the type of the
spectral approximation [3]. We observe a difference of at most 10% between the spatial dis-
tribution (calculated with Eq. (2)) of 1.46-eV neutrons in water (neutrons emitted from a
two-dimensional isotropic 235U fission source) and the experimental results of [5] in the
distance interval which is characterized by a flux-density attenuation of three orders of
magnitude. At a lower hydrogen concentration (water-saturated quartz sand) the comparison
with the results of the Monte Carlo method shows full agreement within the error limits. The
age figures for iron-water and aluminum-water mixtures and graphite differ from the experi-
mental results of [5-7] by at most 5%.
The space-time and the energy distributions of moderated neutrons were calculated on
the basis of the properties of the space-time multiplicativity of the distribution function
[1]. The time distribution is given by the total moderation time ts(u) and the dispersion
D[t(u)] of the pulse of moderated neutrons [8].
A program making use of the ENDL-2 library was compiled for routine calculations of the
neutron parameters (Eqs. (5)-(7)) and of t s (u) and D[t(u)]. Inelastic scattering, absorption,
and anisotropy of the scattering in the center-of-mass system were taken into account up to
the fourth order. Equations (5)-(7), supplemented by equations for the moderation time and.
the dispersion of the pulse, determine the unknown set of neutron characteristics of the
moderator and define the field of epithermal neutrons with a precision which is adequate for
applications in nuclear geophysics.
LITERATURE CITED
1. D. A. Kozhevnikov, Neutron Characteristics of Rock and Their Use in the Geology of the
Oil and Gas Industry [in Russian], Nedra, Moscow (1982).
2. D. A. Kozhevnikov and A. I. Pshenichnyuk, At. Energ., 51, No. 6, 366 (1981).
3. D. A. Kozhevnikov and A. I. Pshenichnyuk, At. Energ., 51, No. 6, 370 (1981).
4. I. A. Kozachok, V. V. Kulik, and V. I. Pirogov, At. Energ., 38, No. 3, 167 (1975).
5. R. Paschall, Nucl. Sci. Eng., 20, 451 (1964).
6. P. Palmedo, Nucl. Sci. Eng., 32, 302 (1968).
7. R. Paschall, J. Nucl. Energy, Part A, 20, 57 (1966).
8. D. A. Kozhevnikov, At. Energ., 40, No. 5, 338 (1976).
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 Declassified and Approved For Release 2013/02/21: CIA-RDP10-02196R000300030004-5
OPTIMIZATION OF RECTIFICATION AND EXCHANGE PILOT
PLANTS FOR ISOTOPE PRODUCTION
V. A. Kaminskii, G.. A. Tevzadze,
V. M. Vetsko, 0. A. Devdariani,
and G. A. Sulaberidze
The difficulty of separation and the high cost of isotope-enriched compounds make it
necessary to optimize the separating installations of all levels of output by using as many
process and apparatus parameters as possible; the most common characteristic reflecting the
efficiency and economy of a process is the cost of the enriched product.
It should be noted that the determination of an explicit cost function is impossible
unless we take account of the design solution of the installation. In the general approach
to optimization described in our previous article [1], two main design variants are possible.
The first, described in [1], assumed that in approximating an ideal cascade by a cascade of
rectangularly sectioned form we use a parallel and series set of columns of the same diameter
and height with individual phase-conversion devices. This variant is intended for industrial
large-scale production of isotopes.
For installations of pilot-plant scale, it is advisable to use one column of different
diameter in each section. These columns are connected according to a scheme in which the
flow of liquid in the following section is part of the flow in the preceding one, and the
gas flow in the following section is fed directly to the preceding one. At the joints between
the sections there is only partial phase conversion. Unlike the first design variant, in the
present case each section is characterized not only by different values of the diameter of
the columns with stage height Hn but also by the relative density of the flow, y , and in
n
the general case it contains packing with elements of different dimensions. The sections of
such a cascade are usually placed vertically one under the other, and the stripping section
constitutes a single whole with the column of the first section. Therefore, it is desirable
to introduce into the optimization conditions either a limitation on the height of the
installation, if the latter is situated in a definite room,
SnHn Dmfn;
In > imin-
Moreover, these parameters are related by the formula
In (N) < Da,
which reflects the condition that the packing elements must be close against the column walls.
In the second design variant the pressure drops arising in each column of the cascade
are added together. In this case, when rectification takes place along the cascade, there
is an increase in the temperature of the process, and, as arule, the separation coefficient
decreases. The variation of H and c along the length of the installation makes it impossible
to use the concept of ideal cascade, so that the only possible method consists in.the direct
determination of the optimum volume of a real cascade within the framework of the general
problem of optimization. Then the cost function can be written as follows:
=I
(VP CO Pta Or - tp) Y'capvi x-11 aaipLi a wpmi+F ~j a Ft '} P (t 1 _ t l X
i P_ td
x (L1 LJ ao,Li+F yao,F,+aWg ase ) ,
i i
F=P cp-cw
op-CW
The cost function directly involves such optimization parameters as the flow L1 in the
first section and the tailings concentration cW. The other parameters (the number N of
packing elements per unit volume, the relative load y, the flows Ln in the subsequent sections,
the number n of such sections, and also the temperature and pressure of the process) are used
in calculating the values of Vplant and tp, which occur in the cost formula.
The values of Q*, Reg, rot gcap' qts' r, m, h, and H are determined from the formula
given in [1], Sn and SW from the formulas of [2], and A and tp by the methods described in
[3] and [4], respectively. According to [5], the pressure drop per unit column length is
equal to
_ V'o.g,acip.PpXc
z 8 (F fr -cst -'Pd-qp)8 (4Vo:gpg/?g?cap1n e
where Vo.g for the case of rectification is given by the formula
VQI P I RT (P).
Vo g= 6OMP
Pg= MPIRT (p).
In laminar flow (when Red op