Forecasting changes in the neutron-physical characteristics of fuel-containing materials
The paper presents the results of calculations for determining neutron-physical characteristics (NFC) of fuelcontaining materials (FCM), including the possibility of formation of critical masses.
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irk-123456789-1591122019-09-25T01:25:32Z Forecasting changes in the neutron-physical characteristics of fuel-containing materials Borysenko, V.I. Goranchuk, V.V. Sidoruk, N.M. Проблеми Чорнобиля The paper presents the results of calculations for determining neutron-physical characteristics (NFC) of fuelcontaining materials (FCM), including the possibility of formation of critical masses. Представлено результати розрахунків по визначенню важливих параметрів паливовмісних матеріалів (ПВМ), що визначають їхні нейтронно-фізичні характеристики (НФХ), у тому числі й можливість утворення критичних мас. Представлены результаты расчетов по определению важных параметров топливосодержащих материалов (ТСМ), определяющих их нейтронно-физические характеристики (НФХ), в том числе и возможность образования критических масс. 2017 Article Forecasting changes in the neutron-physical characteristics of fuel-containing materials / V.I. Borysenko, V.V. Goranchuk, N.M. Sidoruk // Проблеми безпеки атомних електростанцій і Чорнобиля: наук.-техн. зб. — 2017. — Вип. 29. — С. 56-61. — Бібліогр.: 9 назв. — англ. 1813-3584 http://dspace.nbuv.gov.ua/handle/123456789/159112 621.039.5 en Проблеми безпеки атомних електростанцій і Чорнобиля Інститут проблем безпеки атомних електростанцій НАН України |
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Проблеми Чорнобиля Проблеми Чорнобиля |
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Проблеми Чорнобиля Проблеми Чорнобиля Borysenko, V.I. Goranchuk, V.V. Sidoruk, N.M. Forecasting changes in the neutron-physical characteristics of fuel-containing materials Проблеми безпеки атомних електростанцій і Чорнобиля |
| description |
The paper presents the results of calculations for determining neutron-physical characteristics (NFC) of fuelcontaining materials (FCM), including the possibility of formation of critical masses. |
| format |
Article |
| author |
Borysenko, V.I. Goranchuk, V.V. Sidoruk, N.M. |
| author_facet |
Borysenko, V.I. Goranchuk, V.V. Sidoruk, N.M. |
| author_sort |
Borysenko, V.I. |
| title |
Forecasting changes in the neutron-physical characteristics of fuel-containing materials |
| title_short |
Forecasting changes in the neutron-physical characteristics of fuel-containing materials |
| title_full |
Forecasting changes in the neutron-physical characteristics of fuel-containing materials |
| title_fullStr |
Forecasting changes in the neutron-physical characteristics of fuel-containing materials |
| title_full_unstemmed |
Forecasting changes in the neutron-physical characteristics of fuel-containing materials |
| title_sort |
forecasting changes in the neutron-physical characteristics of fuel-containing materials |
| publisher |
Інститут проблем безпеки атомних електростанцій НАН України |
| publishDate |
2017 |
| topic_facet |
Проблеми Чорнобиля |
| url |
http://dspace.nbuv.gov.ua/handle/123456789/159112 |
| citation_txt |
Forecasting changes in the neutron-physical characteristics of fuel-containing materials / V.I. Borysenko, V.V. Goranchuk, N.M. Sidoruk // Проблеми безпеки атомних електростанцій і Чорнобиля: наук.-техн. зб. — 2017. — Вип. 29. — С. 56-61. — Бібліогр.: 9 назв. — англ. |
| series |
Проблеми безпеки атомних електростанцій і Чорнобиля |
| work_keys_str_mv |
AT borysenkovi forecastingchangesintheneutronphysicalcharacteristicsoffuelcontainingmaterials AT goranchukvv forecastingchangesintheneutronphysicalcharacteristicsoffuelcontainingmaterials AT sidoruknm forecastingchangesintheneutronphysicalcharacteristicsoffuelcontainingmaterials |
| first_indexed |
2025-07-14T11:41:30Z |
| last_indexed |
2025-07-14T11:41:30Z |
| _version_ |
1837622405403508736 |
| fulltext |
ISSN 1813-3584 ПРОБЛЕМИ БЕЗПЕКИ АТОМНИХ ЕЛЕКТРОСТАНЦІЙ І ЧОРНОБИЛЯ 2017 ВИП. 29 56
УДК 621.039.58
V. I. Borysenko, V. V. Goranchuk, N. M. Sidoruk
Institute for Safety Problems of Nuclear Power Plants NAS of Ukraine, Lysogirska str., 12, Kyiv, 03028, Ukraine
vborysenko@ispnpp.kiev.ua
FORECASTING CHANGES IN THE NEUTRON-PHYSICAL CHARACTERISTICS
OF FUEL - CONTAINING MATERIALS
The paper presents the results of calculations for determining neutron-physical characteristics (NFC) of fuel-
containing materials (FCM), including the possibility of formation of critical masses. The lack of reliable information at
present about the material composition of some places of FCM requires additional studies to substantiate the nuclear
safety of such clusters of FCM. An important and, in some cases, the only, method of investigating nuclear safety of
places of FCM accumulation is the calculation method for determining the criticality parameters of FCM and predicting
the change of neutron-physical characteristics important for nuclear safety during the planned time of FCM location in
the «Ukryttya» object. Ensuring a reliable control of the intensity of fissions in FCM, as well as clarifying the
supplementary cementitious material composition, requires the introduction of additional methods and channels for
monitoring FCM, for example, on the activity of air and water environments determined by short-lived products of
fission of nuclear fuel in FCM. The paper presents information on the predicted change in the neutron-physical
characteristics of FCM for 120 years from the time of formation of FCM. In particular, the following characteristics are
considered: neutron activity; alpha-activity; power of residual energy releases; change in the ratio of main alpha and
neutron emitters; the effect of changes in other characteristics of FCM on subcriticality, as well as the conditions and
configuration of possible criticality in FCM. For a long period of "planned" storage of FCM in the «Ukryttya» object
facility, some NFC FCM will undergo significant changes. The main decline in residual energy releases, and,
accordingly, FCM activity, including neutron activity, occurred in the first years after the accident. After more than 30
years, the rate of change of these characteristics of FCM significantly decreased.
Keywords: neutron-physical characteristics, fuel-containing materials, criticality parameters, nuclear safety,
«Ukryttya» object.
The code system
To calculate the NFC FCM, the SCALE code [1] was applied, as well as the analytical dependencies
between the amount of nuclide and its activity.
The SCALE software system was developed at the Oak Ridge National Laboratory of the United
States, commissioned by the US Nuclear Regulatory Commission. SCALE is a software tool for analyzing
nuclear safety and designing fuel-based systems. The first version of the SCALE was released in 1980, and
since then it has been widely used both in the US and abroad to perform analyzes of criticality, radiation
safety, heat transfer, burnup [2].
The SCALE software complex was used to justify the nuclear safety of the Spent Fuel Storage
Facility (SFSF) of the ZNPP, SFSF-2 of the ChNPP and the Central Spent Fuel Storage Facility (CSFSF) [3].
In accordance with [2] SCALE is used to substantiate the nuclear safety of spent nuclear fuel (SNF) storage
systems in many countries, including: Bulgaria, Germany, Hungary, Slovakia, Sweden, USA, Japan.
Validation of the radionuclide concentration determination model for fuel burnup.
The estimation of the number of radionuclides in RBMK nuclear fuel at the time of the accident was
carried out by simulating the operation of the RBMK fuel channel for 700 effective days at a capacity of
2 MW, which corresponds to fuel burn-out of ~ 11 MW·day/kg U, defined as the average fuel burnup at the
time Accident [4, 5]. Validation of the calculation model was performed by comparing the calculated and
experimental values [6] of the concentration of the determining radionuclides for the fuel burn-out depth
RBMK-1000 ~ 11 MW·day/kg U.
Residual energy releases of FCM
In Fig. 1 shows the calculated values of the change in the contribution of β-, γ- and α-emitters to the
power of residual energy releases. The contribution to the power of residual energy releases of FCM from α-
emitters is now less than 10 %, and the contribution from stimulated fission caused by spontaneous neutrons
and neutrons (α, n) reactions is insignificant and can be ignored. Analytical assessment of the contribution to
the residual energy release of FCM from the forced fuel division into FCM is presented below.
© V. I. Borysenko, V. V. Goranchuk, N. M. Sidoruk, 2017
mailto:vborysenko@ispnpp.kiev.ua
FORECASTING CHANGES IN THE NEUTRON-PHYSICAL
________________________________________________________________________________________________________________________
ISSN 1813-3584 ПРОБЛЕМИ БЕЗПЕКИ АТОМНИХ ЕЛЕКТРОСТАНЦІЙ І ЧОРНОБИЛЯ 2017 ВИП. 29
57
1.E-02
1.E-01
1.E+00
1.E+01
1.E+02
0 20 40 60 80 100 120
Time, Years
P
o
w
e
r
o
f
re
s
id
u
a
l
e
n
e
rg
y
r
e
le
a
s
e
s
,
W
/k
g
betta
gamma
alfa
Fig. 1. Changes in the contribution of β-, γ- and α-emitters to the power of residual energy releases.
Change in the activity of α-emitters
In contrast to the intensity of β- and γ-emitters, which are continuously decreasing after their
accumulation in the spent fuel, the intensity of the α-emitters after the decay, in the first years after
"unloading" the fuel from the reactor, grows up to a 60-year storage life, and then begins continuous decline.
In Fig. 2 graphs of the activity of the main α-emitters are presented.
Despite the high activity of the α-emitters, their contribution to the power of the residual energy
releases is currently less than 0.04 W/kg, which is equivalent to ~ 10 % of the total residual energy release,
but over time, the contribution of α-emitters to the total power of the residual energy releases increases, and
110 years later will exceed the relative contribution from the other radiators. At the same time, the total
power of residual energy releases throughout the entire period is constantly decreasing (see Fig. 1).
1.0E+09
2.1E+10
4.1E+10
6.1E+10
0 20 40 60 80 100 120
Time, Years
A
lf
a
-a
c
ti
v
it
y
,
a
c
t/
(s
k
g
)
Am-241
Sum
Pu-240
10
20
30
40
0 20 40 60 80 100 120
Time, Years
A
lf
a
-a
c
ti
v
it
y
,
%
.
Fig. 2. Change in the α-activity of FCM Fig. 3. The change in the total α-activity of FCM.
The change in the total α-activity is shown in Fig. 3. After reaching an intermediate minimum of
~ 20 % of the initial α-activity of FCM, the relative increase in α-activity after 60 years from accident will
reach its maximum ~ 32 % of the initial α-activity of FCM, after which there will be a continuous decrease.
The contribution of spontaneous fission neutrons to the residual energy release of FCM
Consider the algorithm for determining the upper estimate of energy release from fission of fuel
nuclei in FCM, based on the following provisions.
For today (~ 30 years after the accident), the estimated neutron activity of fuel due to spontaneous
fission neutrons and neutrons from the (α-n) reaction on light nuclei is ~ 2.5·10
6
n/t (see Fig. 4).
At Kef = 0.999, ignoring the neutron loss, we get that the maximum number of neutrons produced in the
subcritical state is 2.5·10
9
n/t. If even ALL neutrons cause the fission of nuclear fuel - this will correspond to
– 2.5·10
9
acts of fissions per ton of fuel. Considering that 200 MeV of energy is released in the fission event,
3.1·10
10
fissions/s are needed for 1 W power. Thus, the contribution from fission to energy release of FCM
is: 0.082 W/t.
V. I. BORYSENKO, V. V. GORANCHUK, N. M. SIDORUK
________________________________________________________________________________________________________________________
ISSN 1813-3584 ПРОБЛЕМИ БЕЗПЕКИ АТОМНИХ ЕЛЕКТРОСТАНЦІЙ І ЧОРНОБИЛЯ 2017 ВИП. 29
58
Variations in the power of the neutron source to 1.0·10
7
n/t, as well as other "assumptions", can give
an upper estimate of the power output from fission to 1 W/t, which is less than 0.5 % of the residual energy
release on "today".
0.0E+00
5.0E+02
1.0E+03
1.5E+03
2.0E+03
2.5E+03
3.0E+03
20 40 60 80 100 120
Time, Years
N
u
m
b
e
r
o
f
n
e
u
tr
o
n
s
,
n
/k
g
alfa-n neutrons
spontaneous neutrons
sum
Fig. 4. Contribution to neutron activity of FCM from spontaneous fission neutrons
and neutrons (α, n) reactions.
Some authors [7] use another algorithm for determining the forced fission power in a subcritical
reactor, which requires additional "justifications". For example, it is necessary to point out the incorrect
application in formula (2) [8] (analogue of formula (1) of the present paper) instead of the lifetime of the
instantaneous neutrons linst, the effective lifetime of the neutrons lef,.
1
inst
ef
Q
n v v
K
(1)
Let us consider why such a replacement is erroneous. From the definition:
FLUX OF NEUTRONS - φ = n·v, is the product of the neutron density (n/cm
3
) of a given energy and the
neutron velocity (cm/s) of a given energy.
Despite the deceptive dimension of n/(cm
2
·s), these are not neutrons in cm
2
per 1 s, but, in
accordance with the definition - the neutron flux - the total path traveled by neutrons of a given energy, in
1 cm
3
per 1 s.
Thus, a neutron can contribute to the neutron flux only when it "lives" and has the corresponding
energy. To determine the thermal neutron flux at an energy of 0.0253 eV, and, correspondingly, the velocity
v = 2200 m/s, it is necessary to know how long this thermal neutron lives at an "average" velocity
v = 200 m/s. This time corresponds to the neutron diffusion time. Therefore, in formula (1) there should not
be linst = lmod,+ ldif, where ldif is the diffusion time of a thermal neutron, and lmod, is the moderation time of a
neutron, which can be neglected in this analysis. Usually, the condition that lmod < ldif is fulfilled, therefore the
formula (1) is acceptable.
If we consider the effect of delayed neutrons on the formation of the thermal neutron flux, then they
practically do not differ in any way from the instantaneous neutrons. Namely, they have the same diffusion
time as the instantaneous neutrons - therefore, there is no difference in the contribution to the neutron flux
between instant neutrons and delayed neutrons.
The delayed neutron contributes to the neutron flux only when it "lives", i.e. already emitted from
the precursor nucleus and slowed down to thermal energies, and by formula (2) [8] it turns out that the
delayed neutron already "forms" the neutron flux, while still in the precursor nucleus - which does not
correspond to the neutron flux definition.
It should also be noted that as the degree of subcriticality increases, the neutron lifetime will
decrease due to an increase in neutron absorption macro cross-section.
A correct application of formula (1) gives:
4 5
2
1
10 2.2 10 22000
1 1 0.999 cm s
inst
eff
Q n
n v v
K
.
The macro-section of the
235
U fission can be obtained based on the conditions that the density of the FCM is
3 g/cm
3
, the uranium content is 10 %, the enrichment is 1.15 %, respectively: Σf = 0.00509 cm
-1
.
FORECASTING CHANGES IN THE NEUTRON-PHYSICAL
________________________________________________________________________________________________________________________
ISSN 1813-3584 ПРОБЛЕМИ БЕЗПЕКИ АТОМНИХ ЕЛЕКТРОСТАНЦІЙ І ЧОРНОБИЛЯ 2017 ВИП. 29
59
Accordingly, the number of fission reactions (the upper bound)
3
112
cm
f
fis
R .
And the number of fission reactions in 1 m
3
= 1.12·10
8
, or the number of fission reactions per 1 ton
of fuel = 3.7·10
8
. Thus, the contribution to the energy release from stimulated fission in FCM caused by the
spontaneous fission neutron flux does not exceed 0.1 W/t of fuel, which correlates well with the maximum
estimation of the forced fission power given above.
Nuclear Safety of FCM
A large number of publications [4, 5] is devoted to the nuclear safety of places of FCM clusters.
Let us consider the conservative conditions under which criticality can be attained in the various
locations of FCM clusters in the chemical composition.
In accordance with the accepted classification, the main FCM can be divided by chemical
composition into brown and black ceramics [4, 5, 9].
In Fig. 5 shows the dependence of the neutron multiplication factor Kef on the water volume content
for homogeneous FCM with the following parameters: density 3 g/cm
3
, mass content UO2 – 60 %,
enrichment according to
235
U – 1.15 %.
Figures 6 and 7 show the dependence of the neutron multiplication factor on the water volume
content for heterogeneous FCM with the following parameters: the volume filled with cylindrical fuel rods
located in 2.0 cm steps is a cylinder of radius 2 m and height 1 m surrounded by a concrete reflector
thickness 0.5 m, enrichment at
235
U – 1.15 %. The density and size of the fuel pellet are chosen on the basis
that the mass fraction of fuel in the simulated volume remains constant and is 60 %.
In Fig. 6 simulation data are presented for the case when only fuel and water are present in FCM.
Fig. 7 shows the modeling data for the case when there are some other materials in FCM, in addition to fuel
and water, according to [9].
0.92
0.96
1.00
1.04
1.08
0 0.2 0.4 0.6 0.8
Volume fraction of water
c
o
e
ff
ic
ie
n
t
o
f
n
e
u
tr
o
n
m
u
lt
ip
lic
a
ti
o
n
.
1
2
0.5
0.6
0.7
0.8
0.9
1.0
1.1
0.0 0.2 0.4 0.6 0.8 1.0
Volume fraction of water
c
o
e
ff
ic
ie
n
t
o
f
n
e
u
tr
o
n
m
u
lt
ip
lic
a
ti
o
n
.
1
2
3
Fig. 5. Change in the coefficient of neutron multi-
plication in homogeneous FCM: 1 - for an infinite
FCM medium; 2 - for a cylinder with a radius of 2 m,
and a height of 1 m, surrounded by a concrete reflector
0.5 m thick.
Fig. 6. The change in the multiplication factor of
neutrons in heterogeneous FCM, consisting of fuel (UO2)
and water. (Tablet radius and density UO2, respectively:
1 - 5.75 mm, 10.0 g/cm
3
, 2 – 7.5 mm, 5.88 g/cm
3
,
3 – 9.0 mm, 4.084 g/cm
3
.)
Thus, it can be concluded that for the material composition of black and brown ceramics, under the
assumption of a homogeneous model of the multiplying medium, it is impossible to obtain conditions for
achieving criticality. Criticality in a homogeneous model of the multiplying medium at a ceramic density of
3 g/cm
3
can be achieved starting with a mass content of 60 % in fuel, which is not yet found in any of the
fuel core samples.
For the case of a heterogeneous model of the multiplying medium, the criticality estimates are
sufficiently identical.
In Table the simulation results for determining the minimum cylinder size consisting of fuel rods
with a diameter of 11.5 mm and surrounding water are presented. Fuel rods are presented in the form of
tablets UO2 enrichment of 1.15 %, density of 10 g/cm
3
.
V. I. BORYSENKO, V. V. GORANCHUK, N. M. SIDORUK
________________________________________________________________________________________________________________________
ISSN 1813-3584 ПРОБЛЕМИ БЕЗПЕКИ АТОМНИХ ЕЛЕКТРОСТАНЦІЙ І ЧОРНОБИЛЯ 2017 ВИП. 29
60
The dependence of Kef on the radius of the multiplying medium 1.0 m high, consisting
of fuel cells arranged along a triangular lattice
Cylinder radius, m 0.5 0.55 0.60 0/65 0.75 1.00 1.50
Kef 0.9937 1.0015 1.0089 1.0149 1.0227 1.0356 1.0437
Fig. 8 shows the results of modeling to determine the most optimal step in the location of RBMK
fuel to achieve the greatest Kef.
0.5
0.6
0.7
0.8
0.9
1.0
1.1
0.0 0.2 0.4 0.6 0.8 1.0
Volume fraction of water
c
o
e
ff
ic
ie
n
t
o
f
n
e
u
tr
o
n
m
u
lt
ip
lic
a
ti
o
n
. 4
5
0.7
0.8
0.9
1.0
1.1
12 14 16 18 20 22
Step of a triangular fuel grid, mm
co
ef
fi
ci
en
t
o
f
n
eu
tr
o
n
m
u
lt
ip
li
ca
ti
o
n
.
Fig.7. The change in the multiplication factor of
neutrons in heterogeneous FCM, depending on the
density of water in the interstitial space. The radius of
the fuel pellet is 5.75 mm. In the figure, numbers
indicate the simulation results for conditions with
different fuel tablet compositions: 4 - 60 % UO2, 31 %
SiO2, 5 - 45.7 % UO2, 31 % SiO2, 4 % Ca, 4 % Zr,
4 % Na, 3 % Mn, 2.3 % Mg.
The density of the tablet material is 10 g/cm
3
.
Fig. 8. The dependence of Kef on the pitch of the fuel grid.
Conclusions
The results of the predictive modeling of the change in important NFC FCM show that during the
period of forced storage of FCM in the «Ukryttya» object, the continuous decline is observed in terms of
residual energy release, β-, γ-activity, and neutron activity. As for α - activity, after a significant decline in
the first 4 years after the accident α - activity will increase up to 60 years, after which its continuous decline
will begin. The growth of α-activity is determined by the accumulation of
241
Am, due to β-decay of
241
Pu
with a short half-life of ~ 14 years.
The modeling results show that for a multiplying medium with the most conservative parameters,
chosen from the experimental determination of the material composition of brown and black ceramics,
subcritical conditions are ensured, including for possible flooding with water in an amount optimal for
obtaining the maximum multiplying properties of FCM.
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FORECASTING CHANGES IN THE NEUTRON-PHYSICAL
________________________________________________________________________________________________________________________
ISSN 1813-3584 ПРОБЛЕМИ БЕЗПЕКИ АТОМНИХ ЕЛЕКТРОСТАНЦІЙ І ЧОРНОБИЛЯ 2017 ВИП. 29
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В. І. Борисенко, В. В. Горанчук, М. М. Сидорук
Інститут проблем безпеки АЕС НАН України, вул. Лисогірська, 12, Київ, 03028, Україна
ПРОГНОЗУВАННЯ ЗМІНИ НЕЙТРОННО-ФІЗИЧНИХ ХАРАКТЕРИСТИК
ПАЛИВОВМІСНИХ МАТЕРІАЛІВ
Представлено результати розрахунків по визначенню важливих параметрів паливовмісних матеріалів
(ПВМ), що визначають їхні нейтронно-фізичні характеристики (НФХ), у тому числі й можливість утворення
критичних мас. Відсутність у даний час достовірної інформації про матеріальний склад деяких місць скупчень
ПВМ вимагає проведення додаткових досліджень для обґрунтування ядерної безпеки таких скупчень ПВМ.
Важливим, у деяких випадках і єдиним, методом дослідження ядерної безпеки місць скупчення ПВМ є розра-
хунковий метод визначення параметрів критичності ПВМ і прогнозування зміни важливих для ядерної безпеки
нейтронно-фізичних характеристик протягом планованого часу знаходження ПВМ в об'єкті «Укриття».
Забезпечення надійного контролю за інтенсивністю поділу у ПВМ, а також уточнення матеріального складу
ПВМ вимагає впровадження додаткових методів контролю ПВМ, наприклад по активності повітряного й
водного середовищ, визначених за короткоживучими продуктами поділу ядерного палива в ПВМ. У роботі
представлено інформацію про прогнозовану зміну нейтронно-фізичних характеристик ПВМ протягом 120 років
із часу створення ПВМ, зокрема таких як: нейтронна активність, альфа-активність, зміна співвідношення
основних нейтронних і альфа-випромінювачів, вплив зміни інших характеристик ПВМ на підкритичність, а
також умови й конфігурація можливої критичності ПВМ. За тривалий період «планованого» зберігання ПВМ в
об'єкті «Укриття» деякі НФХ ПВМ зазнають істотних змін. Основний спад залишкових енерговиділень, а
відповідно й активності ПВМ, включаючи й нейтронну активність, припав на перші роки після аварії. Після
більше 30 років швидкість зміни зазначених характеристик ПВМ істотно зменшилася.
Ключові слова: нейтронно-фізичні характеристики, паливовмісні матеріали, параметри критичності,
ядерна безпека, об'єкт «Укриття».
В. И. Борисенко, В. В. Горанчук, Н. М. Сидорук
Институт проблем безопасности АЭС НАН Украины, ул. Лысогорская, 12, Киев, 03028, Украина
ПРОГНОЗИРОВАНИЕ ИЗМЕНЕНИЯ НЕЙТРОННО-ФИЗИЧЕСКИХ ХАРАКТЕРИСТИК
ТОПЛИВОСОДЕРЖАЩИХ МАТЕРИАЛОВ
Представлены результаты расчетов по определению важных параметров топливосодержащих матери-
алов (ТСМ), определяющих их нейтронно-физические характеристики (НФХ), в том числе и возможность
образования критических масс. Отсутствие в настоящее время достоверной информации о материальном
составе некоторых мест скоплений ТСМ требует проведения дополнительных исследований для обоснования
ядерной безопасности таких скоплений ТСМ. Важным, в некоторых случаях и единственным, методом
исследования ядерной безопасности мест скопления ТСМ является расчетный метод определения параметров
критичности ТСМ и прогнозирования изменения важных для ядерной безопасности нейтронно-физических
характеристик в течение планируемого времени нахождения ТСМ в объекте «Укрытие». Обеспечение надеж-
ного контроля интенсивности делений в ТСМ, а также уточнение материального состава ТСМ, требует внедре-
ния дополнительных методов контроля ТСМ, например по активности воздушной и водной сред, определяемых
короткоживущими продуктами деления ядерного топлива в ТСМ. В работе представлена информация о
прогнозном изменении нейтронно-физических характеристик ТСМ в течение 120 лет с момента образования
ТСМ, в частности таких как: нейтронная активность, альфа-активность, изменение соотношения основных
нейтронных и альфа-излучателей, влияние изменения других характеристик ТСМ на подкритичность, а также
условия и конфигурация возможной критичности в ТСМ. За продолжительный период «планированного»
хранения ТСМ в объекте «Укрытие» некоторые НФХ ТСМ претерпят существенные изменения. Основной спад
остаточных энерговыделений, а соответственно и активности ТСМ, включая и нейтронную активность, при-
шелся на первые годы после аварии. По прошествии более 30 лет скорость изменения указанных характеристик
ТСМ существенно уменьшилась.
Ключевые слова: нейтронно-физические характеристики, топливосодержащие материалы, параметры
критичности, ядерная безопасность, объект «Укрытие».
Надійшла 29.09.2017
Received 29.09.2017
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