Multi-group modeling of protection against neutron and gamma radiation by materials based on titanium hydride
The paper provides a comparative calculation of the radiation protective efficiency of various composite materials based on titanium hydride using multi-group modeling methods using the ANISN program. The calculations showed the high efficiency of titanium hydride composites with respect to neutron...
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Національний науковий центр «Харківський фізико-технічний інститут» НАН України
2020
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| Цитувати: | Multi-group modeling of protection against neutron and gamma radiation by materials based on titanium hydride / R.N. Yastrebinsky, A.A. Karnaukhov // Problems of atomic science and tecnology. — 2020. — № 2. — С. 119-124. — Бібліогр.: 13 назв. — англ. |
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irk-123456789-1943732023-11-23T16:21:26Z Multi-group modeling of protection against neutron and gamma radiation by materials based on titanium hydride Yastrebinsky, R.N. Karnaukhov, A.A. Physics and the technology of construction materials The paper provides a comparative calculation of the radiation protective efficiency of various composite materials based on titanium hydride using multi-group modeling methods using the ANISN program. The calculations showed the high efficiency of titanium hydride composites with respect to neutron and gamma radiation. The relaxation length of the fast neutron flux density in titanium hydride materials is 5.1…7.0 cm. The spatial-energy distribution of neutron radiation in materials is formed by fast neutrons. The dose rate of gamma rays behind the material is determined mainly by capturing gamma rays arising in the initial layer of protection. Introduction to the composition of the protection of boron atoms reduces the level of capture gamma radiation, but does not affect the attenuation of fast neutrons. Дано порівняльна розрахункова оцінка радіаційно-захисної ефективності різних композиційних матеріалів на основі гідриду титану з використанням методів багатогрупового моделювання за програмою ANISN. Розрахунки показали високу ефективність композитів на основі гідриду титану по відношенню до нейтронного і гамма-випромінювання. Довжина релаксації щільності потоку швидких нейтронів у матеріалах на основі гідриду титану становить 5,1…7,0 см. Просторово-енергетичний розподіл нейтронного випромінювання в матеріалах формується швидкими нейтронами. Потужність дози гамма-квантів за матеріалом визначається в основному захватними гамма-квантами, що виникають у початковому шарі захисту. Введення до складу захисту атомів бору знижує рівень захватного гамма-випромінювання, але не впливає на ослаблення швидких нейтронів. Дана сравнительная расчетная оценка радиационно-защитной эффективности различных композиционных материалов на основе гидрида титана с использованием методов многогруппового моделирования по программе ANISN. Расчеты показали высокую эффективность композитов на основе гидрида титана по отношению к нейтронному и гамма-излучению. Длина релаксации плотности потока быстрых нейтронов в материалах на основе гидрида титана составляет 5,1…7,0 см. Пространственноэнергетическое распределение нейтронного излучения в материалах формируется быстрыми нейтронами. Мощность дозы гамма-квантов за материалом определяется в основном захватными гамма-квантами, возникающими в начальном слое защиты. Введение в состав защиты атомов бора снижает уровень захватного гамма-излучения, но не влияет на ослабление быстрых нейтронов. 2020 Article Multi-group modeling of protection against neutron and gamma radiation by materials based on titanium hydride / R.N. Yastrebinsky, A.A. Karnaukhov // Problems of atomic science and tecnology. — 2020. — № 2. — С. 119-124. — Бібліогр.: 13 назв. — англ. 1562-6016 http://dspace.nbuv.gov.ua/handle/123456789/194373 621.039.539.7 en Вопросы атомной науки и техники Національний науковий центр «Харківський фізико-технічний інститут» НАН України |
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Digital Library of Periodicals of National Academy of Sciences of Ukraine |
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Physics and the technology of construction materials Physics and the technology of construction materials |
| spellingShingle |
Physics and the technology of construction materials Physics and the technology of construction materials Yastrebinsky, R.N. Karnaukhov, A.A. Multi-group modeling of protection against neutron and gamma radiation by materials based on titanium hydride Вопросы атомной науки и техники |
| description |
The paper provides a comparative calculation of the radiation protective efficiency of various composite materials based on titanium hydride using multi-group modeling methods using the ANISN program. The calculations showed the high efficiency of titanium hydride composites with respect to neutron and gamma radiation. The relaxation length of the fast neutron flux density in titanium hydride materials is 5.1…7.0 cm. The spatial-energy distribution of neutron radiation in materials is formed by fast neutrons. The dose rate of gamma rays behind the material is determined mainly by capturing gamma rays arising in the initial layer of protection. Introduction to the composition of the protection of boron atoms reduces the level of capture gamma radiation, but does not affect the attenuation of fast neutrons. |
| format |
Article |
| author |
Yastrebinsky, R.N. Karnaukhov, A.A. |
| author_facet |
Yastrebinsky, R.N. Karnaukhov, A.A. |
| author_sort |
Yastrebinsky, R.N. |
| title |
Multi-group modeling of protection against neutron and gamma radiation by materials based on titanium hydride |
| title_short |
Multi-group modeling of protection against neutron and gamma radiation by materials based on titanium hydride |
| title_full |
Multi-group modeling of protection against neutron and gamma radiation by materials based on titanium hydride |
| title_fullStr |
Multi-group modeling of protection against neutron and gamma radiation by materials based on titanium hydride |
| title_full_unstemmed |
Multi-group modeling of protection against neutron and gamma radiation by materials based on titanium hydride |
| title_sort |
multi-group modeling of protection against neutron and gamma radiation by materials based on titanium hydride |
| publisher |
Національний науковий центр «Харківський фізико-технічний інститут» НАН України |
| publishDate |
2020 |
| topic_facet |
Physics and the technology of construction materials |
| url |
http://dspace.nbuv.gov.ua/handle/123456789/194373 |
| citation_txt |
Multi-group modeling of protection against neutron and gamma radiation by materials based on titanium hydride / R.N. Yastrebinsky, A.A. Karnaukhov // Problems of atomic science and tecnology. — 2020. — № 2. — С. 119-124. — Бібліогр.: 13 назв. — англ. |
| series |
Вопросы атомной науки и техники |
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AT yastrebinskyrn multigroupmodelingofprotectionagainstneutronandgammaradiationbymaterialsbasedontitaniumhydride AT karnaukhovaa multigroupmodelingofprotectionagainstneutronandgammaradiationbymaterialsbasedontitaniumhydride |
| first_indexed |
2025-07-16T21:38:09Z |
| last_indexed |
2025-07-16T21:38:09Z |
| _version_ |
1837841140872642560 |
| fulltext |
ISSN 1562-6016. PASТ. 2020. №2(126), p. 119-124.
UDC 621.039.539.7
MULTI-GROUP MODELING OF PROTECTION AGAINST NEUTRON
AND GAMMA RADIATION BY MATERIALS BASED
ON TITANIUM HYDRIDE
Roman N. Yastrebinsky, Alexander A. Karnaukhov
Belgorod State Technological University Named After V.G. Shoukhov,
Belgorod, Russian Federation
E-mail: yrndo@mail.ru
The paper provides a comparative calculation of the radiation protective efficiency of various composite
materials based on titanium hydride using multi-group modeling methods using the ANISN program. The
calculations showed the high efficiency of titanium hydride composites with respect to neutron and gamma
radiation. The relaxation length of the fast neutron flux density in titanium hydride materials is 5.1…7.0 cm. The
spatial-energy distribution of neutron radiation in materials is formed by fast neutrons. The dose rate of gamma rays
behind the material is determined mainly by capturing gamma rays arising in the initial layer of protection.
Introduction to the composition of the protection of boron atoms reduces the level of capture gamma radiation, but
does not affect the attenuation of fast neutrons.
INTRODUCTION
The design of radiation protection of nuclear power
plants (NPP) is a multi-stage process based on the
search for optimal solutions to ensure the radiation
safety of maintenance personnel and to protect the
reactor plant equipment from unacceptable radiation
damage, i.e. weakening of the basic values of the
functionals of ionizing radiation to the maximum
permissible values [1].
In the process of designing radiation protection, the
necessary modeling of protective properties is carried
out using multi-group and statistical calculation
methods, the results of which determine the
composition, geometry and thickness of the protection.
For a calculated or known radiation intensity of a
nuclear reactor, it is necessary to select the composition
of protection in such a way that it provides thermal and
biological protection, as well as equipment protection to
the established maximum permissible standards. At the
same time, biological protection is of primary
importance, which ensures the protection of staff and
the public from radiation [2].
The design of the radiation protection of NPP is a
multi-stage process based on the search for optimal
solutions to ensure the radiation safety of maintenance
personnel and to protect the reactor installation
equipment from unacceptable radiation damage, i.e.
weakening of the basic values of the functionals of
ionizing radiation to the maximum permissible values
[1].
In the process of designing radiation protection, the
necessary modeling of protective properties is carried
out using multi-group and statistical calculation
methods, the results of which determine the
composition, geometry and thickness of the protection.
For a calculated or known radiation intensity of a
nuclear reactor, it is necessary to select the composition
of protection in such a way that it provides thermal and
biological protection, as well as equipment protection to
the established maximum permissible standards. At the
same time, biological protection is of primary
importance, which ensures the protection of staff and
the public from radiation [2].
The task of designing protection can be seen as the
opposite in relation to the calculation of the attenuation
of radiation from the given sources. In this case, the
initial data used by the maximum permissible radiation
levels for protection, and the desired quantities are the
parameters of protection. This task is performed by the
method of successive approximations based on the
approximations of the security settings using the
following program code calculates the basic functionals
of the radiation. The results of the comparison of the
values obtained functionals with the maximum
permissible values of the adjustments accepted part of
the protection with the subsequent performance of the
calculation. The final composition of protection,
thickness and geometry of the layers is produced when
the coincidence of the calculated values of the
functionals with the specified maximum permissible
values [1, 2].
Thus, the main challenge in designing is finding the
necessary optimality criterion, and choosing the best
variant of development of the set of the admissible in
accordance with the quality indicators. The optimality
of the solution selected depends on the number of
options considered and the variation area of the main
technical parameters affecting performance. The most
important parameters when designing radiation
shielding reactor materials are nuclide composition of
the fillers, the binder matrix used for the calculation of
the basic functionals of attenuation of neutron and
gamma radiation as well as thermal, structural,
mechanical, chemical properties and radiation resistance
of the final composite. Used fillers, as a rule, should
provide multi-level performance of radiation protection.
In this direction promising the use of functional fillers
having high radiation shielding properties, but also a
high compatibility with the binder matrix [3, 4].
When choosing a filler, special attention is paid to
materials with a high specific density of hydrogen
atoms, which effectively slow down fast neutrons. The
choice of binders is limited by the requirements for the
mobility of mixtures due to the need to fill radiation
protection structures of complex geometry and
inaccessible places [5–8].
To a large extent, the requirements for compact
protection of transport NPP are met by designers based
on titanium hydride materials and, above all, products
based on compact titanium hydride (CTH) with
hydrogen atoms up to 3.5 wt.% аnd titanium hydride
fractions (THF) with a hydrogen atom content of up to
3.6 mass%, which have found application in a number
of projects [9–11].
To assess the protective effectiveness of various
composite materials based on titanium hydride, this
paper presents comparative calculated data using multi-
group modeling methods. The performed calculations
are estimates and made for hypothetical compositions.
MATERIALS AND METHODS
To conduct computational studies used a
composition based on a CTH, THF, the fraction of
titanium hydride and portland cement (PC) with
different water-cement ratio (CTHF). It also describes
boron-containing compositions of the composites as
well as comparative analysis with serpentinite concrete.
Variation calculations of spatial-energy distribution
of neutrons and γ-rays in materials protection were
carried out according to the program ANISN [12],
designed for solving one-dimensional transport
equations of the theory of radiation transport by discrete
ordinates, taking into account the scattering anisotropy
in the multi-group approximation. ANISN solves any
order of approximation of the scattering phase function
with Legendre polynomials (PN) and using the boundary
conditions in the form of albedo, that specify for each
energy group of neutrons returning to the system at the
boundary [13].
Spectrum of neutrons in the program ANISN
calculated for the 22-group splitting of the energy
interval. A gamma spectrum was 18-group splitting. For
comparison of the protective properties of materials are
all design options normalized to one and the same
reactor power. The distribution of fluxes of neutrons
and γ-rays, as well as the distribution of dose from
neutrons and γ-rays in materials protection were
considered in plane geometry.
Concentrations of elements used in the calculations
were determined on the basis of elemental and chemical
composition of materials (Tables 1–3). The
compositions and volumetric mass of the investigated
materials are presented in Table 4.
Table 1
The elemental composition of the material CTH
Element, wt.%
Ti H Fe Si Al C O N Cl F
95.0 3.44 0.2 0.1 1.0 0.07 0.15 0.04 810
-4
310
-4
Table 2
Elemental composition of THF material
Element, wt.%
Ti H Al C Fe Si O N
95.63 3.50 0.40 0.05 0.20 0.08 0.10 0.04
Table 3
Chemical composition of PC, wt.%
СаО SiО2 Аl2О3 Fе2О3 MgO SO3
65.8 20.8 4.9 3.5 2.9 2.1
Table 4
Compositions of protection materials
No.
Material,
Composition*
Bulk
weight,
kg/m
3
Nuclear
concentration
of hydrogen,
×10
24
1/cm
3
1 CTH 3800 0.0747
2 CTH in the masonry,
taking into account
block gaps
3400 0.0646
3 THF 2526 0.0542
4 CTHF (80.6 wt.%
DHT, PC,
W/C = 0.8)
2790 0.0482
5 CTHF (75.8 wt.%
DHT, PC,
W/C = 0.3)
2970 0.0476
6 CTHF (90.4 wt.%
DHT, PC,
W/C = 0.9)
3127 0.0568
7 CTHF (85.5 wt.%
DHT, PC,
W/C = 0.27)
3262 0.0558
8 CTHF with boron
(84.5 wt.% DHT,
PC, BA,
W/C = 0.27) **
3262 0.0552
9 Serpentinite concrete
(serpentinite –
80.4%, PC – 17.2%,
W/C = 0.75)
2100 0.0163
*Water-cement ratio (W/C) of the prepared mixture;
**boron-containing additive (BA) 5 wt.%.
In variant calculations, compositions consisting of a
reactor core, reactor structural elements, a reflector, and
a layer of the protective material under study 1 m thick
(which corresponds to the thickness of the hydrogen-
containing layer in real protection) were considered.
Moreover, in the calculations it was assumed that the
protective material is located immediately after the steel
reactor vessel with a thickness of 125 mm, which allows
gamma rays to leak from the core, the reactor structures,
and the reactor vessel.
RESULTS AND ITS DISCUSSION
The results of calculating the functionals are
presented in Figs. 1–4, which shows the nature of the
distribution of the flux density of fast and thermal
neutrons, as well as the dose rate of neutron and gamma
radiation over the thickness of the protection materials.
As follows from the figures, in a first approximation,
neutron attenuation can be described by a simple
expression (1):
( )
, (1)
where Ф0 – neutron flux density without protection;
Ф(d) – neutron flux density behind a layer of protection
thick d, cm; λ – neutron relaxation length, cm.
Fig. 1. Distribution of fast neutron flux density (Фf )
in materials No. 1–9
Fig. 2. Distribution of the flux density of thermal
neutrons (Фt ) in materials No. 1–9
Fig. 3. Distribution of neutron dose rate (Pn)
in materials No. 1–9
The calculations showed that the spatial-energy
distribution of neutron radiation in the material is
formed by fast neutrons. The nature of the distribution
of the dose rate of gamma radiation (Pr) over the
thickness of the protection and its value behind the
protection determines the radiation flowing onto the
front wall and the capture gamma radiation generated in
the initial layer of the material, a few centimeters thick
(and for the materials under consideration, in our
calculation composition reactor protection, the first
component is less than the second). In this case, with an
increase in the layer thickness, an equilibrium spectrum
is established for fast and thermal neutrons. Therefore,
when comparing the protective properties of materials
in this case, the value of the fast neutron flux density
can be considered as the main criterion. The same
pattern is followed by a weakening of the dose rate of a
neutron study.
Fig. 4. The distribution of the equivalent dose rate of
gamma-rays (Pγ) in materials No. 1–9
The introduction of boron atoms into the
composition (composite No. 8), which has a high
absorption cross section of slow neutrons (thermal and
epithermal), significantly reduces the thermal neutron
flux density and capture gamma radiation. Moreover, it
has almost no effect on the fast neutron flux density (Фf)
and neutron dose rate (Pn). A similar material without
boron is material No. 7. Compared to material No. 7,
behind material No. 8, the value Фt at a thickness of 1 m
decreases 95 times, and the value of Pγ decreases 3.7
times.
Table 5
The relaxation length of fast neutrons (f.n, cm) on the
thickness of the protective layer (h, cm) of materials
No.
material
f.n. (h), cm
0…30 30…60 60…100
1 4.3 5.4 6.3
2 4.8 6.0 6.9
3 6.8 7.9 9.3
4 5.9 6.9 7.9
5 5.8 6.7 7.6
6 5.2 6.3 7.3
7 5.1 6.1 7.0
8 5.2 6.2 7.1
9 9.3 9.7 10.1
The relaxation lengths of the fast neutron flux
density and the dose rate of gamma radiation in the
materials under study, obtained on the basis of the
calculation data for the attenuation of the functionals of
neutron and gamma radiation, are presented in Tables 5
and 6.
The relaxation length λ at the layer thickness d = d1–
d2 was determined from the relationship (2) and 3):
( ) ( ( ) ( )), (2)
( ) ( ( ) ( )), (3)
1,00E-09
1,00E-08
1,00E-07
1,00E-06
1,00E-05
1,00E-04
1,00E-03
1,00E-02
1,00E-01
1,00E+00
1,00E+01
0 20 40 60 80 100
F
as
t
n
eu
tr
o
n
f
lu
x
d
en
si
ty
,
Ф
f
1
/(
cm
2
·s
)
Material thickness, cm
9
4
5
6
8
3
7
1
2
1,00E-10
1,00E-09
1,00E-08
1,00E-07
1,00E-06
1,00E-05
1,00E-04
1,00E-03
1,00E-02
1,00E-01
1,00E+00
1,00E+01
1,00E+02
0 20 40 60 80 100
T
h
er
m
al
n
eu
tr
o
n
f
lu
x
d
en
si
ty
,
F
T
1
/(
cm
2
·s
)
Material thickness, cm
9
4
5
6
1
3
7
8
2
1,00E-08
1,00E-07
1,00E-06
1,00E-05
1,00E-04
1,00E-03
1,00E-02
1,00E-01
1,00E+00
1,00E+01
1,00E+02
0 20 40 60 80 100
N
eu
tr
o
n
d
o
se
r
at
e,
P
n
,
μ
S
v
/h
Material thickness, cm
9
4
5
6
8
3
7
1
2
1,00E-06
1,00E-05
1,00E-04
1,00E-03
1,00E-02
1,00E-01
1,00E+00
1,00E+01
0 20 40 60 80 100
P
o
w
er
o
f
th
e
eq
u
iv
al
en
t
d
o
se
o
f
γ
-r
ay
s,
R
γ,
µ
S
v
/h
Material thickness, cm
9
4
5
6
1
3
7
8
2
where Ф(d1), Ф(d2) and Р(d1), Р(d2) – accordingly, the
neutron flux density and the dose rate of gamma rays on
the thickness of the protection layer d1 and d2.
Table 6
The relaxation length of the dose rate of gamma rays
(t, cm) on the thickness of the protection layer (h, cm)
of materials
No.
material
г (h), cm
0…30 30…60 60…100
1 8.3 8.8 9.3
2 9.1 9.7 10.3
3 12.3 13.6 14.9
4 10.4 11.4 12.4
5 9.9 10.8 11.7
6 9.6 10.4 11.1
7 9.2 9.9 10.5
8 10.1 10.1 10.5
9 13.6 14.7 15.7
In the physical sense, the relaxation length λ
indicates the thickness of the layer of material, where
the density of neutron flux is attenuated in e times
(e = 2.718).
Analysis of the obtained data (see Tables 5 and 6)
shows that with the increase of the thickness calculated
layer of the investigated material, the value of the
relaxation length of rapid neutrons λf.n. gradually
increases, tending to a constant value.
This is due to tightening of neutron spectrum on the
thickness, which in turn is connected with the energy
dependence of the scattering of fast neutrons on
hydrogen atoms of materials. The attenuation of the
neutron dose rate by protection from materials based on
THF actually corresponds to the attenuation of the fast
neutron flux density.
The increase of hydrogen content in the material
leads to a decrease in the value λf.n. (see Table 5), and
this dependence decreases with increase in the thickness
of the material. This may be due to the energy loss of
neutrons to the thickness of the protection and reduction
of the processes of elastic scattering on hydrogen. This
trend is most pronounced in the composite of No. 7,
where the values of the relaxation length of rapid
neutrons with a layer thickness of 60…100 cm is
comparable to λf.n. for material No. 2 on the basis of the
CTH in the masonry.
In the composites the presence of hydrogen due to
the content of the hydride phase. The presence of
hydrogen due to its content in the associated water of
the cement stone has a much smaller impact on the
protective properties of materials. In this regard,
compositions with a reduced content of bound water,
which is removed after heat treatment, its protective
properties are not inferior to the materials of the same
density with PC in which part of the bound water
remains.
Since the protection materials under consideration
have weaker protective properties with respect to
gamma radiation than with thermal neutrons (γ > t),
the primary gamma radiation that arises before the
protection and capture gamma radiation generated in the
initial layers of protection will be weaken less than the
profit of new gamma rays generated by the capture of
thermal neutrons. In this regard, with increasing
thickness of the protection γ increases (see Table 6).
However, due to the considerable attenuation of thermal
neutrons arising secondary gamma radiation is not a
significant contributor to the total dose rate of gamma
rays for protection. This explanation is particularly well
illustrated by the example material No. 8 with boron
with the highest value of attenuation of thermal
neutrons, where the value γ throughout the thickness of
the protection is practically the same. Thus, the dose of
gamma rays in a material is mainly determined by the
capture gamma quanta arising in the initial layer of
protection and the formation capture gamma rays across
the thickness of protection are irrelevant. As a result, the
relaxation length of the dose rate of gamma radiation in
these materials is almost independent of the content of
hydrogen.
Calculations show high efficiency of the described
materials. The length of the relaxation of Фf materials
based on THF is an average of 5.1; 6.1, and 7.0
(material No. 7) for thicknesses of 0 to 30; 30 to 60 and
60…100 cm, respectively. The formula of attenuation
material No. 7 on the layer thickness 1 m to more than
~ 3 orders of magnitude greater than is widely used in
protecting nuclear facilities serpentinite concrete. For
products of the SCC in the masonry with the actual
density 3.4 g/cm
3
, the corresponding values of the
lengths of the relaxation of Фf is 4.8; 6.0, and 6.9 cm.
The introduction of boron atoms into the composite
slightly increases the value of γ (materials No. 8 and 7)
at a thickness of 30…60 cm due to the tightening of the
neutron spectra in the thermal and previous thermal
regions. At the same time, the absolute value of the flux
density of thermal neutrons is reduced. The resulting
total effect is a substantial reduction of dose rate of
gamma radiation.
Data analysis of Tables 46 shows that increasing
the density of the material decreases the length of the
relaxation, as for neutron and gamma radiation. For
gamma radiation this process is more pronounced and
amounts to 26% in the thickness of the layer 30…60 cm
and 28% and a thickness of 60…100 cm, i.e. almost
independent of the layer thickness.
Calculations show that increasing the density of
protection to 3.2…3.3 g/cm
3
allows reducing the length
of neutron relaxation to 6.0…7.0 cm, which is 2.2…2.4
times less than in serpentinite concrete. This means that
with the same neutron attenuation factor, the thickness
of protective composites based on titanium hydride and
PC will be ~ 2 times less than serpentinite concrete.
Thus, on the basis of the multi-group modeling, the
high efficiency of materials based on titanium hydride is
shown when used in the protection of transport NPP
instead of serpentinite concrete. Composite materials
based on fractions of titanium hydride and a PC binder,
which have a monolithic protective structure in
comparison with materials based on CTH, are most
effective. In this case, the introduction of boron atoms
into the material of protection leads to a decrease in the
flux density of thermal neutrons and the level of dose
rate of capture gamma radiation
CONCLUSION
Multi-group modeling of the processes of interaction
between the radiation of the active zone of a nuclear
power reactor and materials based on titanium hydride
has shown that gamma radiation behind the protection is
formed by capture radiation flowing from the steel
reactor shell. Secondary gamma radiation generated by
the passage of a neutron flux through the thickness of
the composite does not significantly affect the value of
the gamma ray functional for protection.
The calculations showed the high efficiency of
materials based on THF. According to the efficiency of
attenuation of neutron and gamma radiation, materials
based on fractions of titanium hydride and cement
binder (CTHF) are not much inferior to the CTH
material in the masonry, and they are an order of
magnitude higher than the protective properties of the
fraction without a binder (THF) due to the absence of
neutron slip and denser particle packing. In comparison
with serpentinite concrete, the dose rate of neutrons and
gamma rays in CTHF is much less. The introduction of
boron atoms into the protective composition reduces the
level of capture gamma radiation, but does not affect the
attenuation of fast neutrons. The use of fractions of
titanium hydride without a binder is unreasonable
because of the possible formation of voids in the
protection and slip of neutrons.
ACKNOWLEDGEMENTS
The reported study was funded by RFBR, project
number 19-38-90024.
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Article received 18.02.2020
МНОГОГРУППОВОЕ МОДЕЛИРОВАНИЕ ЗАЩИТЫ ОТ НЕЙТРОННОГО
И ГАММА-ИЗЛУЧЕНИЯ МАТЕРИАЛАМИ НА ОСНОВЕ ГИДРИДА ТИТАНА
Роман Н. Ястребинский, Александр А. Карнаухов
Дана сравнительная расчетная оценка радиационно-защитной эффективности различных
композиционных материалов на основе гидрида титана с использованием методов многогруппового
моделирования по программе ANISN. Расчеты показали высокую эффективность композитов на основе
гидрида титана по отношению к нейтронному и гамма-излучению. Длина релаксации плотности потока
быстрых нейтронов в материалах на основе гидрида титана составляет 5,1…7,0 см. Пространственно-
энергетическое распределение нейтронного излучения в материалах формируется быстрыми нейтронами.
Мощность дозы гамма-квантов за материалом определяется в основном захватными гамма-квантами,
возникающими в начальном слое защиты. Введение в состав защиты атомов бора снижает уровень
захватного гамма-излучения, но не влияет на ослабление быстрых нейтронов.
БАГАТОГРУПОВЕ МОДЕЛЮВАННЯ ЗАХИСТУ ВІД НЕЙТРОННОГО І ГАММА-
ВИПРОМІНЮВАННЯ МАТЕРІАЛАМИ НА ОСНОВІ ГІДРИДУ ТИТАНУ
Роман Н. Ястребинський, Олександр А. Карнаухов
Дано порівняльна розрахункова оцінка радіаційно-захисної ефективності різних композиційних
матеріалів на основі гідриду титану з використанням методів багатогрупового моделювання за програмою
ANISN. Розрахунки показали високу ефективність композитів на основі гідриду титану по відношенню до
нейтронного і гамма-випромінювання. Довжина релаксації щільності потоку швидких нейтронів у
матеріалах на основі гідриду титану становить 5,1…7,0 см. Просторово-енергетичний розподіл нейтронного
випромінювання в матеріалах формується швидкими нейтронами. Потужність дози гамма-квантів за
матеріалом визначається в основному захватними гамма-квантами, що виникають у початковому шарі
захисту. Введення до складу захисту атомів бору знижує рівень захватного гамма-випромінювання, але не
впливає на ослаблення швидких нейтронів.
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