Study of materials for reactors employing molten fluoride salts or PB-BI coolant using an electron irradiation test facility
Overview of results obtained within the framework of the STCU Project #294 is given Наведено огляд результатів, отриманих в рамках проекту УНТЦ № 294. Приведен обзор результатов, полученных в рамках проекта УНТЦ № 294...
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| Опубліковано в: : | Вопросы атомной науки и техники |
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| Дата: | 2005 |
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| Мова: | Англійська |
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Національний науковий центр «Харківський фізико-технічний інститут» НАН України
2005
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| Назва журналу: | Digital Library of Periodicals of National Academy of Sciences of Ukraine |
| Цитувати: | Study of materials for reactors employing molten fluoride salts or PB-BI coolant using an electron irradiation test facility / V.M. Azhazha, A.S. Bakai, I.V. Gurin, I.M. Neklyudov, A.A. Omelchuk, V.F. Zelenskiy, F.A. Garner // Вопросы атомной науки и техники. — 2005. — № 4. — С. 3-19. — Бібліогр.: 40 назв. — англ. |
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Digital Library of Periodicals of National Academy of Sciences of Ukraine| _version_ | 1859861272696193024 |
|---|---|
| author | Azhazha, V.M. Bakai, A.S. Gurin, I.V. Neklyudov, I.M. Omelchuk, A.A. Zelenskiy, V.F. Garner, F.A. |
| author_facet | Azhazha, V.M. Bakai, A.S. Gurin, I.V. Neklyudov, I.M. Omelchuk, A.A. Zelenskiy, V.F. Garner, F.A. |
| citation_txt | Study of materials for reactors employing molten fluoride salts or PB-BI coolant using an electron irradiation test facility / V.M. Azhazha, A.S. Bakai, I.V. Gurin, I.M. Neklyudov, A.A. Omelchuk, V.F. Zelenskiy, F.A. Garner // Вопросы атомной науки и техники. — 2005. — № 4. — С. 3-19. — Бібліогр.: 40 назв. — англ. |
| collection | DSpace DC |
| container_title | Вопросы атомной науки и техники |
| description | Overview of results obtained within the framework of the STCU Project #294 is given
Наведено огляд результатів, отриманих в рамках проекту УНТЦ № 294.
Приведен обзор результатов, полученных в рамках проекта УНТЦ № 294
|
| first_indexed | 2025-12-07T15:45:41Z |
| format | Article |
| fulltext |
PACS: 28.41.Qb
STUDY OF MATERIALS FOR REACTORS EMPLOYING MOLTEN
FLUORIDE SALTS OR PB-BI COOLANT USING AN ELECTRON
IRRADIATION TEST FACILITY
V.M. Azhazha, O.S. Bakai, I.V. Gurin, I.M. Neklyudov, A.A. Omelchuk, V.F. Zelenskiy
National Science Center Kharkiv Institute of Physics&Technology, 61108 Kharkiv, Ukraine;
Frank Garner
Pacific Northwest National Laboratory, Box 999, Battelle Boulevard, Richland WA,
99352, USA
Overview of results obtained within the framework of the STCU Project #294 is given.
1. INTRODUCTION
The experience in development and use of the nucle
ar energy for more than a half of century brought us, as
it had to be, good and bad news. Advantages of the
atomic energy industry are:
− low, as compare to fossil fuel sources, price of ener
gy;
− low level of pollutions and environmental contami
nations;
− large amount of available nuclear fuel;
− developed, investigated and applied in industry ad
vanced fuel elements, structural materials, technolo
gies and safety securement for atomic power engi
neering is a good base for the further advance of the
nuclear power engineering.
The most significant disadvantages of the nuclear
power are:
− increasing amount of the stored nuclear waste, high
ly toxic radionuclide, including military Pu and oth
er minor actinides (MA);
− radioactive contaminations of the environment due
to the nuclear industry and accidents and disasters
on the nuclear stations;
− lack of strict control proliferation of the nuclear ma
terials around the world inspite of serious efforts by
international institutions to restrict this process. The
main difficulty of the non-proliferation activity is
connected with the fact that the declared peaceful
use of the nuclear energy allows to store military Pu
and other fissile materials which can be used in
“pure” and “dirty” bombs.
One can see that advantages and disadvantages are
of a global scale. They cannot be ignored or postponed
and transferred to the next people generations without
proper solutions. Fast exhausting of gas and oil makes
the problems extremely sharp. Evidently that a proper
solution has to be found in historically short time. It
worth to note that the problems of nuclear power indus
try in Ukraine have additional specification. Many nu
clear stations were built and are in use in Ukraine pro
viding nearly half of the electric energy generation. Pro
duction of main nuclear fuel constituents, powerful met
allurgical industry, advanced scientific and technologi
cal base of Ukraine were tightly incorporated in the nu
clear industry of USSR. Nowadays the processing of the
nuclear fuel cycle as well as industry of used nuclear
fuel storage and reprocessing is on a very beginning de
velopment stage in our state. For this reason an optimal
roadmap for nuclear energetic development in Ukraine
for short-range and long-range time has to be developed
taking into account results of the global evolution in this
field and using existing industrial and scientific links
and collaboration.
Recently an international community elaborated a
Technological Roadmap for Generation IV Nuclear Sys
tems (G-IV) [1]. Concepts of the next generation nucle
ar systems are proposed also in other projects [2,3]. It
includes description of an optimal nuclear strategy and
considerations and estimations of several projects of the
nuclear reactors of the next generation. New ideas on
fission fuel cycle, nuclear transmutation, use of the re
leased energy and ecological safety are formulated. The
most general ideas which one can meet in the project in
different combinations are following:
1) Considerably higher, than in the reactors of
current generation, outlet temperature, 600…
1200 °C . Due to this feature generation of hy
drogen, as an ecologically pure fuel, in thermo
chemical and electrochemical cycles becomes
possible.
2) Accelerator driven systems (ADS) are most
probable candidates for G-IV. Subcritical, with
neutron reproducing coefficient κ≈98 , reac
tors can successfully be used with external neu
tron sources. Necessitated 2% of the neutrons
will be generated by proton (or electron) beams
projected on a liquid metallic target. Molten
Pb, Pb-Bi eutectic (PBE) or Hg are considered
as possible candidates for the target.
3) Metallic melts (Pb,PBE, Na) look like attrac
tive coolants. In opposite to gaseous coolants
(e.g. He) the metallic melts are efficiently oper
ating at low pressures. Especially interesting
are Pb and PBE because they are not as chemi
cally aggressive as Na in a case of leakage.
4) Fluid fuel (in form of molten fluoride salts) is
considered as a promising non-traditional fuel
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3
in several projects. Use of the fluid fuel simpli
fies the fuel processing and reprocessing cycle.
A worth noting is that a minor experience in long-
term use of components and prototypes of the G-IV sys
tems is stored. Some data are obtained and used systems
are molten salts reactors constructed and tested in Oak-
Ridge (see [4-6] and refs. quoted) fast reactors with Pb
and PBE coolants designed and used in submarines and
satellites in USSR and Russia [7,8]. But this experience
is not enough for designing of constructions and use of
the proposed G-IV reactors for a long time.
A key problem of the G-IV program is R&D of con
struction materials compatible with molten salts and Pb
and PBE melts at high temperatures. A brief summary
of the state of the art materials for the nuclear systems
of the next generation is given in ref. [9]. Ni-Mo alloys
(Hastelloys) were used in MSR [4-6] and ferritic-
martensitic chromium steels were applied in reacrors
with Pb and PBE [7,8]. The experimental data obtained
show which way of the advanced materials designing is
promising but we are staying at the very beginning of
this way.
The STCU Project 294 (approved for financing on
March 1st 2003) is devoted to designing and test of
materials for MSR and ADS. Promising construction
materials have to be tested in reactors using special
loops reproducing all conditions of real MRS and ADS.
Experiments of this type take a lot of time and need spe
cial technique and testing procedures An idea to imitate
reactor conditions using e.g. electron or proton and irra
diation for material tests is not a new one. Many impor
tant results of the nuclear science where got by this way.
But it is well known that there are no proper scaling
laws allowing to predict material behavior in reactor
condition on the base of data obtained at different tem
peratures, irradiation fields, doses and dose rates. It is
because of strong connection of different cooperative
processes in solids under irradiation due to large local
non-equilibrium generated by radiation damages. Nev
ertheless the nature of kinetic processes induced, modi
fied and activated by irradiation on microscopic level
are very similar independently on the type of irradiation.
For this reason investigations of the kinetic processes
under electron irradiation gives a valuable information
which can be used for prediction of the material behav
ior in reactors using a proper rescaling of the kinetic co
efficients.
A special remark concerning the corrosion kinetics
in surface layers has to be made. The corrosion is con
trolled by chemical reactions. The reaction rate is rather
sensitive to the local energy deposition due to irradia
tion. The irradiation excites electron states and can pro
duce atomic replacements and displacements. These ra
diation induced elementary processes need excitation
energy from one to some tens of eV per atom. Acceler
ated electrons and γ -quanta are ideal agents for activa
tion of this processes. Therefore an electron accelerator
of a proper energy (10 MeV per electron) was chosen to
imitate impact of the reactor irradiation on the corrosion
kinetics. In parallel, the electron irradiation impacts dif
fusion and phase transformations in bulk and, especial
ly, within the grain boundaries. This general idea was
used as basic one in planned within the framework of
the Project 294 investigations.
It is worth to note that the performed corrosion tests
in metallic melts and salts without irradiation not only
provide a reference data base but allow estimating ther
modynamic driving forces controlling the corrosion and
aging kinetics.
To realize the project a methodology of tests was de
veloped in detail and tested materials were chosen (see
below). Really we have performed pilot experiments
which have to be continued in larger scale but they al
lowed to get an important information on impact of
electron and γ -irradiation on corrosion, compositional
and mechanical properties of the tested materials.
In this overview paper we summarize and discuss
the main results obtained within the framework of the
Project. In detail the results obtained are presented in
separated papers of this issue. Purposive goal of the
overview is to describe concepts, methodology and to
consider the main results obtained in their
interconnection.
2. METHODOLOGY
2.1. ELECTRON IRRADIATION TEST FACILITY
(EITF)
Main reasons for use electron irradiation for investi
gation of the radiation effect on materials for MRS and
ADS are formulated in Introduction. Some features of
the electron irradiation which are helpful in this kind of
investigations are following:
1. Because energetic thresholds of nuclear reactions in
duced by electrons and γ - rays are nearly 10MeV
or something less, use of the electron beams of 10M
eV energy excludes considerable radioactivity of the
irradiated materials. For this reason hot cells are not
needed in postirradiation investigations of the mate
rials.
2. The penetration length of electrons with 10 MeV en
ergy is of some centimeters. On this scale of length
the electron energy decreases from 10 to 1 MeV or
even less due to bremsstrahlung and another inelas
tic processes. It means that, from one side, the test
ampoules can have reasonable size to mimic large
scale experiments and, from other side, the irradia
tion conditions are dramatically changing on com
paratively small length. Therefore we can use many
specimens of sub millimeter thickness within one
ampoule to investigate dependence of corrosion and
radiation damage processes versus energy deposited
by e- and γ - fields.
3. The electron beam can be used as an efficient heater
of the irradiated target, to keep a needed temperature
if the electron beam power, target geometry and con
struction are chosen by a proper way.
The EITF construction [10] is shown in Fig.1. It was
designed taking into account calculations of the e- and
γ - fields as well as calculations of the thermal balance
and temperature field [11].
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4
2.2. SIMULATIONS OF THE e- AND γ - FIELDS
AND DEPOSITED ENERGY DISTRIBUTION
The electron and γ -radiation fields, the deposited
energy profiles in salt and samples as well as the point
defects generation in Haselloy have been calculated by
means simulation of the radiation transport in the test
bench geometry [11]. The computer code based on the
CERN GEANT4 Monte Carlo toolkit [13] was used for
this simulation. The results of simulations are needed
for optimization of the ampoule construction as well as
for obtaining quantitative dependencies of the specimen
properties on the dose.
2.3. TEMPERATURE DISTRIBUTION
Equation of the heat balance taking into account the
integral deposited energy and the heat irradiation from
the surface is a base for estimation of the temperature
balance of irradiated target in total. To calculate temper
ature field within an ampoule the deposited energy dis
tribution and heat transport equations have to be de
duced and solved [12]. The temperature distribution is
an important characteristic of the irradiated target. In
our experiments the goal was to make this distribution
as homogeneous as possible. The calculations per
formed show that the temperature variance is really
small [12].
2.4. MECHANICAL TESTS
Changes of macroscopic mechanical properties due
to impact of the molten fluoride salts or metallic
coolants can be revealed by standard mechanical tests
[14]. Because these changes are caused by structural
alternations and, first of all, by alternations of the grain
boundaries and dislocational system, as well as by
compositional heterogeneities. Therefore systematic
microindentations and nanoindentations of the materials
in the vicinity of surface and in bulk were performed.
Nanoindentations allow to measure local value of
Young modulus which is sensitive to local
compositional and structural changes [15].
Mechanical tests with impact of ultrasonic vibrations
on the been tested material are of special interests be
cause from one side they give information on the acti
vated slipping of dislocations and, from another side,
plastic deformation of vibrating or shocked elements of
the reactor construction can be very different as com
pare to that without vibrations. Within the project
framework a home-made device for mechanical tests
under ultrasonic impact was designed and constructed
[16].
In combination with metallography and
microanalysis these methods are rather informative.
2.5. MICROANALYSIS
Compositional and phase changes can be revealed
by mean of microanalysis. It is important to to reveal
these changes on all scales in surface layers and in bulk
of the tested materials. The X-ray phase analysis was
performed on a DRON-UM diffractometer (CuKα-radi
ation), the atomic absorption analysis on a Pionicum
SP-9 spectrophotometer, the IR spectroscopic investiga
tions on a Specord 80M instrument, the X-ray mi
croanalysis and visual control of the state of the sample
surface on a REM-101M scanning electron microscope-
microanalyzer. Secondary ion mass spectrometry
(SIMS) also was used to detect compositional changes
in surface layers of tested samples [17, 18].
Microanalysis by mean of high resolution transmission
electron microscopy is in progress.
2.6. CORROSION TESTS
The corrosion of alloys and their constituents in a
molten eutectic sodium fluoride-zirconium fluoride
mixture has been studied by cyclic voltammetry, X-ray
analysis, scanning electron microscopy and
metallography. The corrosion rates of the samples ex
posed in salt at 650 °C for different times were meas
ured afterward in a fresh molten salt by the voltamperic
method. Impact of electron irradiation on the corrosion
rate of the alloys in molten fluorides has been
investigated [15, 19].
2.7. ROENTGEN SPECTROSCOPY
After irradiation by 10MeV electrons a small
account of nuclear reactions leading to formation of γ -
active nuclei happens. Therefore a possibility to reveal
products of the alloys corrosion in molten salt under
irradiation exists. Inspite this method does not produce a
proper quantitative compositional analysis it gives a
semiquantitative and qualitative information [20].
2.8. THEORETICAL MODELS
For quantitative analysis of the experimental data
appropriate theoretical models have to be developed.
Main goal of these model is to describe corrosion
kinetics and phase transformation kinetics taking into
account impact of electron irradiation. Refs. [21, 22] are
devoted to this problem solution.
2.9. THERMODYNAMICS OF ZR IN ALKALI
HALIDE MELTS
Equilibrium constants determine the thermodynamic
driven forces responsible for chemical reactions and
dissolution of alloys. Because Zr and alkali metals are
basic elements of all used fluoride salts of MSR, it is
worth to know their thermodynamic characteristics.
Investigations of the thermodynamics of Zr in alkali
melts are performed within the framework of this
project [23].
3. MATERIALS
3.1. HASTELLOYS
Ni-Mo alloys (Hastelloys) showed a satisfactory cor
rosion resistance in contact with molten fluoride salts
ZrF 4 -NaF and ZrF 4 -LiF-BeF at temperatures 600…
800 °C [4-6]. It was revealed that comparatively small
compositional changes of alloys can lead to consider
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5
able decrease or increase of the corrosion resistance. For
example alloying with a few percents of Nb improves
the corrosion resistance of the alloy. In [5] is proposed
an “optimized” composition of Ni-Mo alloy that showed
acceptable radiation resistance under neutron irradiation
up to dose ~1020 n/cm 2 . The data obtained [4-6] can
serve very much as a starting point and background in
further R&D materials compatible with fluoride molten
salts at temperature above 600 °C . Therefore for the
corrosional tests of Ni-Mo alloys we have chosen the al
loys of two compositions (alloy A and alloy B, see Ta
ble 1) that are very similar to those reported in Refs. [5].
In the Table 1 also the composition of alloy U which
was tested under neutron irradiation is presented. The
compounds of high purity were used to exclude impact
of uncontrolled impurities. It has to be noted that micro
scopic mechanisms of the structural and compositional
changes of Hastelloys in the molten salts under and
withoutirradiation were never studied in detail. Very
scant information about precipitations and hetero
geneities of these alloys is available.
Table 1
Compositions of alloys A, B, and U
Element Composition (wt %),
Alloy A Alloy B Alloy U
Nickel 78.2 78.2 78.2
Molybdenum 11.7 11.7 11…12
Chromium 6.7 6.2 5…7
Titanium 0.5 0.5 0.5
Aluminum 0.8 0.8 0.8…1.2
Iron 1.5 1.5 1.5
Manganese 0.5 0.5 0.5
Silicon 0.15 0.15 0.15
Niobium - 0.5 0.5
Yttrium - 0.05 -
Carbon - - 0.038
Sulfur - - 0.012
3.2. W AND Mo COVERS
Hastelloys are corroding very fast in molten Pb, PbE
and Bi because these melts are good dissolvent for Ni.
Ferritic/martensitic (F/M) chromium steels with oxided
surface and steels covered by W or Mo are considered
as materials with much better corrosion resistance in the
mentioned melts. Beside W and Mo have small solubili
ty in Pb and one can expect that these metals can be
used as acceptable constitutive materials compatible
with Pb abd PbE.
In NSC KIPT are developed different efficient tech
nologies for covers formation. One of these technolo
gies is suitable for W and Mo covers using thermolise of
carbonyls of these metals. To make clear whether this
technology provides W and Mo covers of a proper qual
ity we have used the Hastelloy A as substrate. If these
covers demonstrate good protective properties on this
wittingly “bad” substrate, one can expect that they will
be more efficient on (F/M) chromium steels which have
an appropriate corrosion resistance by themselves. Up to
now only corrosion resistance of alloy A covered by W
and Mo in molten ZrF 4 -NiF salt was studied and fur
ther experiments are planned.
3.3. CARBON-CARBON COMPOSITE
Different carbon materials are very useful in nuclear
industry from very beginning. These materials can be
used at high temperatures. Graphites are radiation
resistant and have comparatively low chemical activity
in contacts with fluoride salts [5]. Graphite has low
mechanical properties what is a serious limitation for
many applications. But C-C composites have much
better mechanical properties. An original technology of
C-C composite preparation is developed and efficiently
used for different applications in NSC KIPT for many
years. Tests of the corrosion resistance of these
materials in molten salts under irradiation can give
quantitative results on their corrosion properties.
4. RESULTS
4.1. HOME MADE DEVICES
4.1.1. ELECTRON IRRADIATION TEST
FACILITY
EITF has been constructed and built at Linac-10
electron linear accelerator in NSC-KIPT (Fig. 1 and
Fig. 2) [24].
Schematically construction of the ampoule is shown
in Fig. 3 and Fig. 4.
A container assembly that consists of ampoules
holding samples imbedded in fluoride salt is disposed in
a chamber in Ar atmosphere. The container assembly
includes 16 ampoules made of the C-C composite.
Electron beam with the 10 MeV energy and up to 1mA
of average current (up to 1 kW of power) is in use. The
electron beam is scanning over the chamber inlet
window. The container assembly temperature is
monitored with three thermocouples. The temperature is
kept to be (650оС ± 15)оС and controlled by the beam
current.
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6
Fig. 1. Construction of the EITF-KIPT
Fig. 2. General view of EITF
Fig. 3. Scheme of the ampoule construction
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ВОПРОСЫ АТОМНОЙ НАУКИ И ТЕХНИКИ. 2005. №.4.
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38
12
25
4
0
3
10
7
Fig. 4. Frontal section of the loaded ampoule
4.1.2. ULTRASONIC MECHANICAL TEST
FACILITY (UMTF)
The UMTF was designed and constructed for
mechanical tests with impact of ultrasonic vibrations on
the material [16]. A scheme of the UMTF is shown in
Fig. 5. Ultrasonic vibrations are being generated by
magnetostrictive converter and through concentrator
focused on the tested specimen. The vacuum furnace
allows to keep specimen temperature up to 1500 °C .
The ultrasound frequency range is 18…22 Khz. The
ultrasonic generator has controlled power 2…4 kw but
the transferred to specimen fraction of this power is yet
not known. Storage of the measurement data will be
made in digital form. Impact of the ultrasonic vibrations
on the plastic deformation of irradiated and unirradiated
specimens is planned.
Fig. 5. Scheme of the USTF for mechanical tests under
ultrasonic impact:
1. Water-jacket; 2. Magnetostrictive converter; 3. Ul
trasound concentrator; 4. Specimen; 5. Vacuum fur
nace; 6. Flange; 7. Supportive blocks; 8. Dynamometer
with piezometers of load; 9. Sylphon;
10. Worm gears; 11. Configuration of reduction system;
12. Feed spindle; 13. Dynamometer with piezometers of
strain; 14. Diffusion vacuum pump
4.2. RESULTS OF SIMULATIONS OF THE
e- AND γ - FIELDS AND DEPOSITED ENERGY
DISTRIBUTION
To calculate the electron and γ -radiation fields, the
deposited energy profiles in salt and specimens of
Haselloy, the detailed Monte Carlo modeling was per
formed. The computer code based on the CERN
GEANT4 Monte Carlo toolkit [13] was used for the
simulation [11].
The results of the simulations were used for optimiz
ation of the ampoule construction. Namely, the number
of specimens in the salt and their disposition within the
ampoule were chosen to use the irradiating electron
beam as efficiently as possible. Within the scope of the
experimental geometry proposed the maximal variabili
ty of energy deposition on all interfaces of Hastelloy
and molten fluoride salt is achieved. The modeling of
complex multicomponent heterogeneous system of the
assembled ampoule taking into account all valuable
physical processes that determine the relativistic elec
trons energy deposition in substance provides informa
tion on the kinetics of development of electron-photon
processes in this system. Results of the simulations al
low obtaining quantitative dependencies of the speci
men properties changes on the dose of deposited energy.
The simulations data on data on depth profiles of the
fluence and deposited energy are shown in Fig. 6. The
digital data of these parameters and doses of the point
defects (vacancies and interstitials) generations are pre
sented in Table 2.
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123
4
5
Electron beam
1
40
2
0.6
30
1 – body of the ampoule (C-C
composite)
2 – double plates (Hastelloy)
3 – fluoride melt
4 – auxiliary plates (C-C
composite)
5 – inserts (graphite)
8
1.0 1.2 1.4 1.6 1.8 2.0 2.2 2.4 2.6 2.8 3.0
10-3
10-2
10-1
100
101
e-, total
e-, primary
gamma
El
ec
tro
n
en
er
gy
fl
ue
nc
e,
M
eV
/c
m
2 p
er
e
- /c
m
2
Depth, cm
carbon, 15.1 mm
fluoride, 2.0 mm
hastalloy, 0.6 mm
10 MeV
15 16 17 18 19 20 21 22 23 24 25
10 -2
10 -1
100
101
D
ep
os
ite
d
en
er
gy
, M
eV
/c
m
3 p
er
e
- /c
m
2
Depth, mm
total
primary
hastalloy, 0.6 mm
fluoride, 2 mm
а b
Fig. 6. Depth dependencies (a) of the particles normalized energy fluencies, incl. primary electrons and produced
gammas, and (b) the deposited energy profile
The results of simulations were used for analysis of
the dose dependencies of the corrosion stability and the
mechanical properties of irradiated specimens. Besides
the calculations of the temperature distribution within
the ampoule are also based on these data.
It is seen that the energy deposited on the first sur
face (5066 eV/at) is two orders in magnitude larger than
that on the last surface (63 eV/at) and that the deposited
energy gradually decreases from the first to last surface.
The point defects generation dose is decreasing from
2.12×10–3 dpa within the first specimen to 4.42×10–6 dpa
within the last one. Usually the ratio of the unstable
Frenkel pairs to the stable ones is ~102. Therefore the ir
radiation induced replacements can modify the transport
and segregation phenomena, especially in first and sec
ond specimens.
The deposited in the fluoride salt energy is several
times less than that in the Hastelloy specimens. Never
theless it is large enough to impact the chemical reac
tions within the salt.
Table 2
Deposited energies in the interface (near-surface) regions of the Hastelloy plates and the melt
and atomic concentrations of Frenkel pairs in Hastelloy
Sample Sample
Surface Material
Depth Deposited Energy Point Defects
cm eV/atom percentage dpa percentage
fluoride 1.7075 2221.52
1
1…1
1…2
2
2…1
2…2
Hastelloy
1.7125 5066.72 100.0 100.0 2.12×10–3 100.0 100.0
1.7375 4906.46 96.84 96.84 2.01×10–3 94.44 94.44
1.7425 4815.29 95.04 95.04 1.95×10–3 91.94 91.94
1.7675 4208.23 83.06 83.06 1.62×10–3 76.20 76.20
fluoride
1.7725 1794.76
1.9675 1010.82
3
3…1
3…2
4
4…1
4…2
Hastelloy
1.9725 2347.23 46.33 100.0 7.00×10–4 32.96 100.0
1.9975 1698.40 33.52 72.36 4.80×10–4 22.59 68.54
2.0025 1563.33 30.85 66.60 4.33×10–4 20.37 61.80
2.0275 969.88 19.14 41.32 2.44×10–4 11.48 34.83
fluoride
2.0325 375.17
2.2275 90.93
5
5…1
5-2
6
6-1
6-2
Hastelloy
2.2325 214.55 4.23 100.0 2.75×10–5 1.30 100.0
2.2575 107.01 2.11 49.88 1.06×10–5 0.50 38.64
2.2625 95.04 1.88 44.30 8.83×10–6 0.42 32.07
2.2875 63.82 1.26 29.74 4.42×10–6 0.21 16.07
fluoride 2.2925 24.39
The parameters of the irradiation effect on the salt
and specimens for 700 hours exposure were estimated to
be appropriate for investigations of the irradiation im
pact on corrosion and mechanical properties and com
positional heterogeneities of hastelloy contacting with
molten fluorides.
4.3. TEMPERATURE DISTRIBUTIONS
The calculations of the temperature distribution
within the ampoule are based on the simulated distribu
tions of the deposited energy distribution within am
poule reported in the previous section [12]. It is found
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9
that the temperature variance within an ampoule is not
more than 15˚. This variance can be neglected when ex
perimental data are analyzed.
4.4. RESULTS OF TESTS OF C-C COMPOSITE
There exists a big family of C-C composites differ
ing in the fabrication technology and properties. In the
tests performed we have used two types of the C-C
composite differing in density of the surface layers. To
make more dense composite an additional densification
with pyrocarbon was used [25]. The tests performed
show that exposure in the molten salt, without and under
e-irradiation, and lead does not change properties and
composition of the tested specimens and ampoules. It
looks like the test conditions were not “hard” enough
for this material.
In result of temperature cycling from room temperat
ure to 600…700°C and isothermal soaking at high tem
peratures (for up to 700 h) it was shown [25] that C-C
composite is impermeable to molten fluoride mixture
and lead. This material is indestructible in an inert atmo
sphere and no considerable structure changes are re
vealed after 700 hours e-irradiation in inert atmosphere.
It can be recommended for use in transmutation techno
logies.
Very eloquent are results of nanoindentations of the
tested material. The results of the nanoindentations are
presented in Table 3 and Fig. 7.
Samples of C-C composite were investigated with
out (sample 1) and with densification with pyrocarbon
(samples 2). In samples 1, two constituents have been
distinguished: fibers and binding material. The hardness
and elastic modulus were determined from the depth of
indentation according to the Oliver and Farr technique
[26, 27]. A microscopic examination did not reveal in
dentations on the samples as this material has a high ca
pacity for the elastic recovery of the original shape. This
is evidenced by loading and unloading curves (Fig. 7).
Table 3
Mechanical properties of carbon-carbon composites
Sample Elastic modules (GPa) Hardness (Gpa)
Sample 1 (binding material) 17±2 2.83±0,14
Sample 1 (fiber) 19±1 2.11±0,23
Sample 1 (fiber) 17±0 1.72±0,38
Sample 2 (fiber) 20±2 2.72±0,35
0 100 200 300 400 500 600 700
0
2
4
6
8
10
Lo
ad
(
m
N
)
Displacement (nm)
a b c
Fig. 7. Loading and unloading curves for carbon-carbon composites (a) without contact with the fluoride melt
and without irradiation; (b,c) after electron beam irradiation in a sodium fluoride – zirconium fluoride melt for 700
h at 650 °C: with deposited energy (b) (5066 MeV/atom) and (c) (64 MeV/atom)
The loading-unloading curves generally diverge on
unloading. For the samples under investigation, these
curves first diverge and then converge, and almost com
plete recovery of the original shape is observed on com
plete unloading. This character of dependencies and
elastic modules are also the same as those of glassy car
bon [27]. It looks like this material is a variety of glassy
carbon (amorphous carbon in which sp2 bonds predom
inated).
The hardness and elastic modules of the fibers and
binding materials of the composites in samples 1 differ
not so much what means that the mechanical properties
are isotropic.
4.5. IMPACT OF ZRF4 – NAF MOLTEN SALT
AND E-IRRADIATION ON PROPERTIES
OF ALLOYS A AND B
4.5.1. METALLOGRAPHY
Metallography was used to reveal macroscopic
structural changes of the tested specimens. In the case of
a considerable intercrystalline corrossion ( observed e.g.
in refs [5,6]) it is clearly seen in form cracks and etched
grain boundaries. The depth and density of the
intercrystalline cracks can be used as a corrosion
parameter [5]. Metallography of irradiated and unirradi
ated specimens in salt are shown in figs 12-15 [28].
In Fig. 8 a specimen of the alloy A exposured in salt
at T=650°C for 700 hours is shown. The size of the
grains is 60 μm . Precipitates in grain bulk and bound
aries of some μm in size are seen. The specimen was
not polished before the test and its roughness is not a re
sult of corrosion. Attentive inspection of the grain
boundaries shows that the depth of the “cracking” is not
more than 1…3 μm .
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10
Fig. 8. Structure of the Hastelloy A after 700 hr expo
sure in the molten salt at T=650°C
Considerably different is morphology of the alloy A
specimens irradiated in the molten salt for 700hr at
T=650°C (Fig. 9). The depth of intercrystalline cor
rosion essentially depends on the deposited energy. The
depth is 25…30 μm at Ed=5066 eV /at and 5…
10 μm at Ed=64 ev /at .
The statistics of precipitates of secondary phases is
yet not considered quantitatively but the size of precipi
tates within grain boundaries are larger in irradiated ma
terial (Fig. 10).
a b
Fig. 9. Structure of the alloy A after 700 hr exposure in the molten salt under irradiation at T=650°C with
the deposited energy (a) Ed=5066 eV /at and (b) Ed=64 ev /at . Corrosion and precipitates within bound
aries are larger at Ed=5066 eV /at
The metallography data show considerable depen
dence of the intercrystalline corrosion on the alloy com
position (Fig. 11). The structure specimens of alloy B is
changed not so much after irradiation to large (fig. 11,a)
and smaller (fig. 11,b) doses. Twines and many precipi
tates in grain bulk are seen. The grain size occurs to be
not depending on Ed. It is seen that the alloy B possess
ing an “optimized composition” really has much higher
corrosion resistance than the alloy A. But the mechani
cal tests show tha the irradiation changes microstructure
of the alloy B too.
Fig. 10. Precipitates within the grain boundaries of irradiated ( Ed=5066 eV /at ) alloy A
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11
a b
Fig. 11. Structure of the alloy B after 700 hr exposure in the molten salt under irradiation at T=650°C with
the deposited energy (a) Ed=5066 eV /at and (b) Ed=64 ev /at . No considerable corrosion is seen
4.5.2. CORROSION KINETICS AND COMPOSI
TIONAL CHANGES OF THE ALLOYS A AND B
IN MOLTEN SALTS
The metallography data gives a macroscopic picture
of the structural changes due to the corrosion of the
alloys in molten salt. Investigations of compositional
changes reveal peculiarities of the irradiation impact on
the alloy recomposition under irradiation in molten salt.
Dissolution of the components in salt and penetration of
the salt constituents in specimen determine the
corrosion kinetics. Voltammograms and corrosion
current measurements are important characteristics of
the corrosion determining the corrosion rate of material
[15, 17-19].
An important characteristic of the corrosion is con
tent of the dissolved elements in the salt. An informative
parameter is the ratio of concentration of an element in
salt to its concentration in alloy. When thermodynamic
equilibrium is archived this ratio determines the equilib
rium constant, K i :
K i=
c i , l
salt
ci , l
alloy =exp− μi
salt−μ i
alloy
T . (1)
Here c i , l
salt , c i , l
alloy are the equilibrium concentra
tions, K i is the equilibrium constant of ith element,
and μi
salt , μ i
alloy are chemical potential of this ele
ment in salt and alloy respectively.
In the case that the system is not in the equilibrium
the dissolution rate, dc i
salt /dt , depends on the trans
port rate of ith element to the surface and the dissolution
rate on the surface. The last one is proportional to
Sinh[ μi
alloy− μi
salt /T ] while the transport rate
strongly depends on phase structure and compositional
heterogeneities of alloy. Considered in ref. [29] corro
sion kinetics gives a constructive idea on the nature of
these processes.
The compositional changes of the alloy A exposured
in molten salt for 100, 200, 500 and 700 hr without ir
radiation revealed being in surface layers of 1…3 μm
in the thick [15, 18].
Within the salt all main components of the alloy are
revealed. The data on ratios c i
salt /c i
alloy are presented
in Table 5. It has to be noted that we have at the mo
ment a poor statistics and that the variance of the data
obtained is unknown. Nevertheless one can conclude
that without irradiation the basic elements Ni and Mo
have rather low solubility in salt, K i10−2 . For Ti
the equilibrium constant is also comparatively small, 2
¿10−2 , while for Al it is 10 times larger. Evidently
concentrations of Cr and Fe in the salt after 700hr are
not saturated. Of course larger statistics is needed to get
more accurate data.
Quantitative analysis of the salt composition after
the test shows that the content of dissolved Ni, Mo, Al,
Cr, Ti is small. An usefull parameter of corrosion kinet
ics is the ratio of an dissolved element content in salt
and in alloy. These parameters are presented in Table 4.
Table 4
Element Composition of
alloy А, at. % Ratio
ci , l
salt at
c i , l
alloy at
after exposure in salt
100 hr 200 hr 500 hr 700 hr
K i
Ni 81.041 0.00151 0.00022 0.00086 0.00108 1.23 10-5
Mo 7.578 0.00141 0.00141 0.00141 0.00141 1.84 10-4
Cr 7.808 0.00506 0.01772 0.00760 0.04295
Fe 1.627 0.11303 0.18088 0.11313 0.24827
Ti 0.148 0.03391 0.03392 0.03394 0.03385 0.23
Al 1.797 0.21194 0.21197 0.21211 0.21159 0.118
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12
-100 0 100 200 300 400 500 600 700 800
0.00
0.05
0.10
0.15
0.20
0.25
Alloy B, Ed=64 eV/at
Alloy B, E
d
=5066 eV/at
C
or
ro
si
on
ra
te
,
m
m
/y
ea
r
Exposure in salt, hours
Alloy A, E
d
=5066 eV/at
Alloy A, Ed=64 eV/at
Alloy A, E
d
=0
Fig. 12. Corrosion rates of the alloy A and B depending
on the exposure time in salt at 650 °C without and un
der e-irradiation at Ed= 5066 eV/at and Ed= 64 eV/at.
The points depicted the corrosion rates of the unirradi
ated alloy A and irradiated alloy B at Ed= 64 eV/at with
700hr are overlapping
The corrosion rates of the alloy A in fresh molten
salt after exposures 100, 200, 500 and 700 hr at
T=650°C without irradiation and for alloys A and
B after 700 hr under irradiation are presented in Fig. 12.
It is seen that without irradiation the corrosion rate satu
rates on a rather low level ~2 ¿ 10 −4 mm / yr . The
presented in table 5 data are consisting with this conclu
sion but Cr and Fe dissolution are not saturated. For this
reason the equilibrium constant for these elements is not
determined. It is worth noting that applied electronic
field in voltammetric measurements violets the equilib
rium and that measured corrosion rate using this method
can differ from that in absence of the electric field. Nev
ertheless the measured quantities can be accepted as re
liable ones.
0 2 0 4 0 6 0 8 0 1 0 0
0 .0
0 .2
0 .4
0 .6
0 .8
1 .0
1 .2
1 .4
1 .6
1 .8
2 .0
2 .2
2 .4
Ti
ta
ni
um
c
on
te
nt
, a
t %
D e p th , µm
a
0 2 0 4 0 6 0 8 0 1 0 0
0
1
2
3
4
5
6
7
8
C
hr
om
iu
m
c
on
te
nt
, a
t %
D e p th , µ m
b
Fig. 13. Redistribution of a) Ti content and b) Cr content in surface layers of irradiated and non-irradiated
samples after 700 hours exposure in salt () –no irradiation; ( ) – Edep =64 eV/atom; () - Edep =5066 eV/atom
The corrosion rate of irradiated in molten salt of al
loy A is 0.1 −4 mm / yr and is something higher for
specimens with larger deposited energy Ed . The cor
rosion rate of alloy B to be essentially depending on
Ed . It is ~10−1 mm/ yr at Ed=5066 eV /at
while for Ed=64 eV /at it is ~10−3mm/ yr .
This phenomenon have to be investigated in more detail
but now one can conclude that the corrosion rate of al
loy B in molten salt is sensitive to the deposited energy.
REMMA microanalysis and SIMS were used to in
vestigate compositional changes of the surface layers of
specimens. Results show considerable compositional
heterogeneities in the surface layers. Ti heterogeneities
and Cr, Al and Mo depletion zones were revealed in the
surface layer. As example data on redistribution of Ti
and Cr without and under e-irradiatrion in alloy A is
shown in Fig. 13. Minor compositional changes are seen
in samples after exposure in salt without irradiation. The
scale and amplitude of compositional heterogeneities
are depending on the deposited energy dose, Ed . Data
of SIMS measurements are rather scattered but they
clearly show that penetration depth of Zr and Na is de
pending on Ed [18]. Both of these elements are dif
fusing mainly in boundaries. A REMMA measurements
data (Fig. 14) show that in a cavern within the grain
boundary closely to the surface layer the salt is located
and that Zr and Na penetrate in the grain of alloy A on
1.25 μm while the depth of these elements penetration
is twice shorter in the case of alloy B at the same dose
Ed = 5066 eV/at.
0
10
20
30
40
50
60
70
80
90
-10 -7.5 -5 -2.5 -1.25 0 0.5 1.25 2.5 5
AlK:
SiK:
TiK:
CrK:
FeK:
MoL:
NiK:
ZrL:
NaK:
0.0
10.0
20.0
30.0
40.0
50.0
60.0
70.0
80.0
90.0
-3 -2 -1 -0.5 0 0.5 1 3
AlK:
SiK:
TiK:
CrK:
FeK:
MoL:
NiK:
ZrL:
NaK:
a b
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13
Fig. 14. Distribution of elements in the vicinity of a cavity within a boundary of a) alloy A and b) alloy B after expo
sure in molten salt under irradiation, Ed = 5066 eV/at . Distance from edge of cavity is given in μm , the refer
ence point coincides with the cavity edge. Negative values of the distance belong to inner part of the cavity
In bulk considerable compositional heterogeneities
are seen in the vicinity of the grain boundaries (Fig. 15).
Ni- Ti-precipitates enriched by and Si are forming with
in the boundaries and in grains. Average precipitate size
depends on Ed
Ni, aloy A
Ti, alloy A Overaly, alloy A
Fig. 15. Ni-Ti precipitates within grain boundaries of alloy A after 700 hr exposure in the molten salt at
T=650°C for 700 hr without irradiation [28]
It is worth noting that the variance of the composi
tional heterogeneities is rather large and that more care
ful investigation of them is needed to make a conclusion
concerning impact of the irradiation on the alloy decom
position and corrosion kinetics. The precipitation kinet
ics also has to be investigated in more detail.
4.5.3. MECHANICAL PROPERTIES
OF THE ALLOYS A AND B
The mechanical properties of reactor construction
materials determine workability of the construction ele
ments. The alloys A and B aged for 50hr at
T=675°C possess rather high (more than 50%) plas
ticity and yield stress σ B , up to 1000 MPa at room
temperature. At T=600°C the yield stress of both
alloys decreases on 50% while the plasticity threshold
decreases much less, approximately on 25%. As result
the plasticity is equal to 14% for alloy A and 19% for
alloy B at T=600°C [14]
Exposure of the alloy A in molten salt for 700hrs is
resulting in increase of σ B up to 20%. It is a result of
the alloy aging and partial decomposition. The precipi
tates seen on the metallographic images, within grains
and grain boundaries, increase in size.
The electron irradiation reduces both considerably
σ B and σ O2 of the alloy A and alloy B and the re
duction is larger at Ed=5066 eV /at than at
Ed=64 eV /at . At the same time, the plasticity does
not depend on composition and deposited energy, at
T=650°C , δ=15 .
In ref.[5] data of mechanical tests of alloy U, which
in composition differs not so much from alloy B, are re
ported. This alloy was exposured to neutron irradiation
in nitrogen atmosphere with neutron energy more than
0.5 Mev and flux 4 1020 neutron/cm2. As result the alloy
was strengthened as compare to unirradiated material
but its plasticity decreases from 59.2 to 33.3 at
T=650°C . At the moment we have not enough ex
perimental data on impact of neutron irradiation on Ni-
Mo alloys to compare with our data.
Nanoidentation gives an additional information on
impact of aging, molten salt and irradiation on the al
loys. A typical size of nanoindentor imprint is 50 nm.
Therefore the nanoindentation allows to measure hard
ness and Young modulus of rather small regions. If
compositional and structural heterogeneities have scale
larger than the indenter imprint, changes of the proper
ties due to these heterogeneities can be revealed.
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14
An example of irradiated with Ed=64 eV /at
specimen of the alloy A after nanoindentation is shown
in Fig. 16 [30].
Revealed impact of e-irradiation on the surface and
intercrystalline corrosion tells not so much on changes
within grains and specimen bulk. Measurements per
formed show that Young modulus of both alloys A and
B irradiated and unirradiated in molten salt are very the
same, in bulk E ≈ 240 GPa and that these quantities are
not sensitive to the dose deposited energy. Both Hμ and
E considerably decrease in the surface layer. The width
of this layer is 3…5 μm without irradiation and 7…
15μm for irradiated specimens. At the same time, the
hardness the unirradiated specimens exposured in salt
for 700hr increase, in bulk Hμ ≈ 6 GPa. For irradiated
specimens of the alloy A this quantity is Hμ ≈ 4.3 GPa
and for the alloy B Hμ= 5 GPa. It occurs that the
nanohardness is sensitive to the alloy composition but it
is not sensitive to the dose of the deposited energy. Pos
sible the dose effect saturates at Ed > 60 ev/at.
Fig. 16. Surface layer of the alloy A irradiated in salt
with Ed=64 eV /at . Intercrystalline cracks and in
dentor imprints are seen
Worth noting is that nanohardness correlates with
the yield stress and σ O2 of specimens.
It occurs that small addition of La (wt%) in salt con
siderably changes the corrosion rate and nanohardness
of the alloy A [19]. In this case the nanoindentations al
lows determine the width of corroded layer very well. In
Fig. 17 the structure of surface layer of the alloy A after
exposure in molten salt with La impurities for 500hr at
T=600°C . It is seen that the corrosion stroked the
sample on depth 20…25 μm.
Fig. 17. SEM of corroded surface layer of the alloy A
after exposure for 500hr in molten salt with La impuri
ties at T=600°C . Three large imprints are marks of
the investigated area. A row of small imprints of the
nanoindentor are seen
The nanohardness measurement data are shown in
Fig. 18.
0 5 10 15 20 25 30 35 40
1
2
3
4
5
6
N
an
oh
ar
dn
es
s,
M
P
a
Distance from edge, µm
Fig. 18. Depth dependence of the nanohardness
of the specimen shown in Fig. 17 [30]
4.8. PROTECTIVE W- AND MO- COVERS
It was found that the Hastelloys are not corrosion
resistant in molten metals: lead, lead-bismuth eutectics
and bismuth. For example, in the molten Bi the standard
specimen of alloy A was dissolved for 300 hr at T = 550
°C. Tungsten and molybdenum are known as corrosion
resistant materials in the mentioned metallic coolants.
These metals were used as protective covers of the alloy
A [31]. We aware that these covers would be more effi
cient on alloys possessing higher corrosion resistance,
e.g. ferritic-martensitic steels. Nevertheless the alloy
with poor corrosion resistance was used to check quality
of the covers. Indeed, if a cover is penetrable for Pb or
Pb-Bi than the substrate will be dissolved after expos
ure in the melt. It is possible that the alloy components
(Ni, Cr, etc.) penetrate trough the cover but this process
is much slower than dissolution of the alloy contacting
with melt.
Samples of the alloys with W and Mo coatings
(coating thickness 60 and 90 µm) were soaked at 550 °C
for over 300 hours in molten metals in absence of oxy
gen. The state of the samples Pb and Pb-Bi melts was
satisfactory. No traces of corrosion were observed visu
ally. No mass loss was noted either. A microsection of
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15
tungsten-coated sample of composition A after contact
with molten lead is shown in Fig. 19.
Fig. 19. Lateral microsection o tungsten-coated
sample of composition A after exposure in molten Pb at
550°C for over 300 hr
In the Bi melt bath, as it was noted, the uncoated
specimens of alloys A and B dissolved completely for
300 hours while the coated samples were not corroded
considerably. The percentage of the alloy components in
melts after exposure of the alloy A specimen covered by
W is given in Table 4.
Table 4
Percentage of the alloy constituents in molten metals-coolants
Alloy constitu
ent
Percentage in the original
alloy
Percentage in the metal melt bath
Pb Pb–Bi Bi
Ni 78.15 0.016 0.0241 0.225
Fe 1.5 0.019 0.0185 0.013
Cr 6.7 0.004 0.005 < 0.003
After the isothermal soaking of alloy samples in
molten metals, microsections were made, and nanohard
ness was measured. The results are following. No
change in hardness has been observed in coating or al
loy of the samples soaked in a molten lead-bismuth eu
tectic (Fig. 20). Very the same results were observed for
the covered alloys specimens after exposure in a molten
lead-bismuth eutectic.
0 20 40 60 80 100 120
0
2
4
6
8
Mo Hastelloy
H
ar
dn
es
s,
G
P
a
Distance from the edge, µm
Exposure in Pb
0 20 40 60 80 100 120
0
2
4
6
8
Alloy BW
H
ar
d
ne
ss
, G
P
a
Distance from the edge, µm
Exposure in Pb
Fig. 20. Nanohardness of molybdenum-coated samples of alloy A (above) and tungsten-coated samples
(below) after contact with molten lead for 300 hours
X-ray microanalysis shows that lead does not penet
rate either into the coating or into the alloy. Traces of
lead get onto the surface of sample during its polishing.
In the coated samples, a redistribution of both the main
alloy constituent (nickel and molybdenum) and the al
loying constituents is observed at the coating-alloy in
terface. In Fig. 21 and Fig. 22 distributions of Ni and Fe
in covers are shown. It is seen that these components
considerably diffuse in covers. The diffusivity of Ni in
Mo and W is very the same while Fe in W diffuses
faster than in Mo. For this reason a depleted Fe zone is
formed in the alloy A in the vicinity of interface. The
data of chemical analysis of the melts after exposure of
covered specimens (Table 4) confirm this conclusion.
Thus these tests show that the W and Mo covers were of
good quality and that these materials can be used for
protection of materials from corrosion in molten fluor
ide salts.
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16
-80 -60 -40 -20 0 20 40 60 80 100
0
20
40
60
80
w
t.
%
Distance from alloy-coating boundary, µm Ni in Mo cover,
Ni in W cover
-80 -60 -40 -20 0 20 40 60 80 100
0,0
0,5
1,0
1,5
2,0
Distance from alloy-coating boundary, µm Fe in Mo cover,
Fe in W cover
C
w
t.
%
Fig. 21. Distributions of Ni in Mo and W covers
and in alloy A
Fig. 22. Distribution of Fe in Mo and W covers and
in alloy A
4.9. PILOT TESTS OF A STEEL USING USTF
Full-scale tests of irradiated and unirradiated speci
mens of alloys A and B will be done in the near future.
Currently we have performed some pilot tests of differ
ent materials to check efficiency of the constructed
USTF. The tests show that the used ultrasound impacts
considerably the mechanical properties of alloys even at
room temperature [16]. The ultrasound generator output
power was 3,5±0,5 kw but the amplitude of the speci
men ultrasonic vibrations was not measured due to lack
ing of a proper tools. An example of the pilot measure
ments is presented in Fig. 23.
0 5 10 15 20 25 30 35 40 45 50 55 60 65
0
5
10
15
20
25
30
35
40
45
50
55
60
65
70
2.
1.
σ,
k
g/
m
m
2
ε, %
Fig. 23. Strain-stress curves of the stainless steel
Х18Н10Т (Cr – 18%, Ni – 10%, Ti – 0.45%) (1) with
out and (2) with impact of ultrasound at room tempera
ture. The strain rate was nearly 0.15 min-1
It is seen that a strengthening and a decrease of plas
ticity (from 55 to 50%) takes place in result of the ultra
sound action. Usually a strengthening of material ap
pears due to multiplication and pinning of dislocations.
It is likely that an additional pinning of dislocation is
caused mainly by vacancy complexes generated by the
ultrasound [32-34].
0 10 20 30 40 50 60
-10
0
10
20
30
40
50
60
70
σ,
k
g/
m
m
2
ε, %
1.
2.
3.
Fig. 24. Strain-stress curves (1) without and (2,3) with
impact of ultrasound at room temperature. The strain
rate was nearly 0.15 min-1 in absence at the ultrasound,
(curve 1). It was equal to 0.15 min-1 (curve 2) and 0.009
min-1 (curve 3) at ultrasound vibrations
In Fig. 24 are shown the stress-strain curves without
sound impact (curve 1) at a large strain rate (dε/dt =
0.15 min-1) and under ultrasound action at smaller stain
rate, dε/dt = 0.03 min-1 (curve 2) and dε/dt = 0.009 min-1
(curve 3). The ultrasound generator output power was
nearly twice less then in previous measurements shown
in Fig. 23. One can see that the specimen hardening and
the plasticity loss due to the ultrasound action increases
when the strain rate decreases. This result is expected
because generation of vacancies and multiplication of
dislocations (if the last one happens) takes place for a
longer time at slower deformation. For this reason the
density of the generated by ultrasound defects is larger
for smaller strain rates.
Performance of full-scale experiments and investiga
tions in a wide temperature range have to be made to
study the microscopic nature of the observed phenome
na. Especially interesting is issue on difference of me
chanical properties of irradiated and unirradiated speci
mens under ultrasonic vibrations.
4.10. THEORETICAL MODELS OF KINETIC
PROCESSES UNDER IRRADIATION
Without irradiation a nonequilibrium conservative
system relaxes approaching the stable equilibrium state.
Thermodynamic driven forces and thermal activations
determine the kinetics of relaxation. A system under ir
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ВОПРОСЫ АТОМНОЙ НАУКИ И ТЕХНИКИ. 2005. №.4.
Серия: Физика радиационных повреждений и радиационное материаловедение (87), с. 3-19.
17
radiation is non-conservative. For it the equilibrium
states and relaxation kinetics have to be described with
in frameworks of properly developed models. Kinetic
coefficients of these models depend on both thermody
namic and irradiation parameters.
Corrosion of multicomponent alloys in molten salts
or metallic coolants is a complicate process including
transport of components, dissolution and a wide spec
trum of chemical reactions. An irradiation is resulting in
local strong damages (generation of the Frenkel pairs,
formation of the displacement cascades, compositional
replacements of atoms) and soft damages (generation of
unstable Frenkel pairs, excitations of atoms). In our ex
periments the electron and x-ray irradiation produces
mainly soft damages which impact considerably the cor
rosional kinetics on the interface alloy-melt.
Features of the kinetic processes on interfaces under
electron irradiation are investigated, taking as example
amorphization of Zr Fe precipitates under electron irra
diation, in ref [36]. This example was chosen because it
was investigated experimentally [37]. A theoretical
model proposed in [38] is based on assumption that just
the strong damages, Frenkel pairs and compositional
disordering, are responsible for this phase transforma
tion. It is shown in [36] that the soft damages can en
force the amorphization inspite in this case the thermo
dynamically stable crystalline phase transforms into
metastable amorphous state. This example shows that
the soft radiation damages can even change the direction
of the phase transformation and therefore they can con
siderably impact the corrosion of alloys in metals. The
stored experimental data on corrosion under irradiation
are yet not enough to develop a proper quantitative de
scription of this processes but a progress in its under
standing is made.
Corrosion rate of a metal in molten metallic coolant
depends on its solubility, coolant velocity, surface struc
ture etc. To decrease the corrosion rate, protective cov
ers are useful. As it is known, in Pb and PBE coolants
protective oxide layers covering the metal surface, are
efficient (see e.g. [39] and references quoted). If the
coolant is saturated by oxygen then the oxide layer can
grow permanently on the surface of metal contacting
with the coolant. Therefore description of the kinetics of
a metal with covered by an oxide surface layer and ki
netics of the oxide growth is a problem of great interest.
In ref. [29] this problem is considered within the frame
work of proposed there kinetic equations.
The metal atoms diffuse through the oxide layer and
dissolve in molten coolant due to a difference of the
chemical potential of the atoms in the metal and solu
tion. Due to the vacancy mechanism of diffusion, the
currents of metal atoms and vacancies are equal in abso
lute value and excess vacancies are pouring into the
metal. Therefore the kinetics of both metal atoms and
vacancies have to be considered to determine the corro
sion rate.
If occurs that the corrosion current of metal atoms is
equal to
∣ jm∣= D vm
ox cvm
ox
Lox
Dm
Pb cm
Pb
Lm
(2)
The width of the oxide layer is Lox and Lm is
the width of diffusional region of the metal atoms with
in the coolant; D vm
ox is the diffusion coefficient and
cvm
ox is the equilibrium concentration of the metallic
vacancies within the oxide; Dm
Pb and cm
Pb is the dif
fusion coefficient and equilibrium concentration of the
metal atoms in the coolant.
To describe the growth kinetics of oxide the diffu
sional transport of both metal and oxygen atoms are tak
en into account. The growth rate is proportional to t
in all cases but the proportionality coefficient depends
on the oxide layer width, diffusion coefficients, chemi
cal potentials, etc.
It is shown [29] that at a high concentration of oxy
gen in the coolant the oxide layer growth can be blocked
due to formation of nuclei of a complex oxide including
atoms of both metal and coolant. It seems that this phe
nomenon was observed in some experiments [40 ].
In the theoretical models developed in [29] the irra
diation role on the kinetics is not yet taken into account
but the results obtained can be helpful for analysis of the
big amount of data of experiments performed in loops
without of an irradiation.
5. DISCUSSION
Inspite the Ni-Mo alloys possess good corrosion
resistance in molten salts the stored up to now data [4-6]
do not allow to select a best alloy for long-term use in
MSR. Our data show that
− irradiation considerably impacts the corrosion;
− corrosion resistivity is rather sensitive to compara
tively small compositional changes. The alloys A
and B have minor composition difference but their
corrosion resistivity without and under irradiation
are considerably different. Beside, the corrosion rate
of alloy B is more sensitive to the deposited energy
Ed;
− impurities of La in the ZrF4-NaF molten salt enhance
the corrosion of the alloy A even without irradiation.
The results obtained show that systematic investiga
tions of corrosion of Ni-Mo alloys of different composi
tions in molten salts containing fluorides of actinides,
lanthanides and other impurities expected to appear in
blanket of MSR is severely needed . An interesting issue
is dependence of Ni-Mo alloys corrosion dependence on
Nb and other impurities content. In ref [6] a lot of data
on dependence of Ni-Mo alloys corrosion on different
alloying elements are published. These data form a good
base for tests of corrosion similar alloys under irradia
tion.
The fact that C-C composite occurred to be rather re
sistant to corrosion and to impurity penetration in
molten salt as well as in Pb and PBE melts at long-term
exposure shows that this material is a good candidate
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ВОПРОСЫ АТОМНОЙ НАУКИ И ТЕХНИКИ. 2005. №.4.
Серия: Физика радиационных повреждений и радиационное материаловедение (87), с. 3-19.
18
for MSR and ADTS. Workability of the composite is
rather good but its functional properties in specified re
actor conditions, especially its mechanical properties,
have to be investigated in more detail. It is desirable to
test the C-C composites at more hard condition includ
ing higher temperature (above 1000 °C ), larger de
posited energy and carbide forming impurities to make
clear under which conditions this material can be used.
A correlation of the nanohardness and yield stress of
the irradiated specimens is revealed in the performed
measurements. Similar correlation of microhardness and
yield stress of irradiated alloys was demonstrated in ref.
[35]. Inspite that a poor data base on this correlation is
stored, it is believed that the nanoindentations can be
used for semi quantitative estimations of the
macroscopic mechanical properties of irradiated
materials.
The specimens of alloy A and B covered by W and
Mo films have considerably better corrosion resistance
in Pb and PBE melts than the uncovered specimens do.
Besides the W and Mo covers are impenetrable for Pb
and Bi. From other side atoms of Ni and Fe, being
components of the alloys A and B, penetrated in the
covers on large depth for 300 hr. It means that
composition of steel under W or Mo cover has to be
thoroughly chosen to diminish solubility of Ni and Fe in
covers and transportation of these elements into metallic
coolant through covers. First of all the ferritic-
martensitic steels, showing acceptable corrosion
resistance in molten Pb and PBE, and modified by
impurities steels of this family have to be used as basic
alloys for covers. It is interesting to compare protective
properties of oxide films and W,Mo covers of steels in
molten Pb and PBE.
An important issue is to make clear to which
measure the corrosion processes of the materials tested
under electron irradiation are simulating impact of
neutron irradiation on the corrosion in molten salts and
metallic coolants. For that additional tests are needed.
Anyway one can expect that the results of tests under
electron irradiation correlate with those under neutron
irradiation and that comparatively fast tests using EITF
should be very useful for design and selection of
advanced materials for MSR and ADTS.
6. CONCLUSIONS
− EITF is an efficient irradiation tool for corrosion
tests of ADTT candidate construction materials.
− No visible changes of macroscopic mechanical,
nano-plastic, compositional properties or corrosion
resistance of C-C composite material fabricated in
NSC KIPT were found after long contact with mol
ten fluoride or metallic coolants.
− Both irradiation ampoules and test samples from C-
C composite show no changes of properties.
− C-C material is therefore proposed for use in con
struction of ADTT.
− Hastelloy alloys A and B have good corrosion resist
ance in fluoride salt at 650°C. Voltamperic measure
ments show that the corrosion rate of both alloys is
negligible after 700 hour exposure without irradi
ation.
− After electron irradiation in EITF for 700 hours the
voltamperic data provide a preliminary estimation of
the corrosion rate to be ~ 0.1 mm/year.
− Changes of the strength and yield stress of the
Hastelloys after e-irradiation in salt in EITF are sim
ilar to those observed in samples after fast neutron
irradiation to dose 3·1020 cm-2 at 650°C.
− These results show that the tools and methods de
veloped in this program can be used for tests of
metallic alloys of different compositions, castings
and thermal treatments.
− The deposition methods developed for W and Mo
coatings ensure corrosion protection coatings which
are resistant to attack in metallic coolants.
− To make clear to which measure the corrosion
processes of the materials tested under electron
irradiation are simulating impact of neutron
irradiation on the corrosion in molten salts and
metallic coolants additional reactor tests are needed.
ACKNOWLEDGEMENT
This research was partially supported by Science &
Technology Center in Ukraine (STCU) within the
framework of Project # 294.
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ИССЛЕДОВАНИЕ МАТЕРИАЛОВ ДЛЯ ЖИДКОСОЛЕВЫХ РЕАКТОРОВ И РЕАКТОРОВ С Pb-Bi ОХЛА
ДИТЕЛЕМ С ПОМОЩЬЮ ЭЛЕКТРОННОГО ОБЛУЧАТЕЛЬНОГО ИСПЫТАТЕЛЬНОГО СТЕНДА
В.M. Ажажа, А.С. Бакай, И.В. Гурин, И.M. Неклюдов, A.A. Омельчук, В.Ф. Зеленский, Ф. Гарнер
Приведен обзор результатов, полученных в рамках проекта УНТЦ № 294.
ДОСЛІДЖЕННЯ МАТЕРІАЛІВ ДЛЯ РІДКОСОЛЬОВИХ РЕАКТОРІВ ТА РЕАКТОРІВ З Pb-Bi ОХОЛОДЖУВА
ЧЕМ ЗА ДОПОМОГОЮ ЕЛЕКТРОННОГО ОПРОМІНЮВАЛЬНОГО ВИПРОБУВАЛЬНОГО СТЕНДУ
В.M. Ажажа, О.С. Бакай, І.В. Гурін, І.M. Неклюдов, A.О. Омельчук, В.Ф. Зеленський, Ф. Гарнер
Наведено огляд результатів, отриманих в рамках проекту УНТЦ № 294.
_______________________________________________________________________________
ВОПРОСЫ АТОМНОЙ НАУКИ И ТЕХНИКИ. 2005. №.4.
Серия: Физика радиационных повреждений и радиационное материаловедение (87), с. 3-19.
20
|
| id | nasplib_isofts_kiev_ua-123456789-80541 |
| institution | Digital Library of Periodicals of National Academy of Sciences of Ukraine |
| issn | 1562-6016 |
| language | English |
| last_indexed | 2025-12-07T15:45:41Z |
| publishDate | 2005 |
| publisher | Національний науковий центр «Харківський фізико-технічний інститут» НАН України |
| record_format | dspace |
| spelling | Azhazha, V.M. Bakai, A.S. Gurin, I.V. Neklyudov, I.M. Omelchuk, A.A. Zelenskiy, V.F. Garner, F.A. 2015-04-18T19:06:53Z 2015-04-18T19:06:53Z 2005 Study of materials for reactors employing molten fluoride salts or PB-BI coolant using an electron irradiation test facility / V.M. Azhazha, A.S. Bakai, I.V. Gurin, I.M. Neklyudov, A.A. Omelchuk, V.F. Zelenskiy, F.A. Garner // Вопросы атомной науки и техники. — 2005. — № 4. — С. 3-19. — Бібліогр.: 40 назв. — англ. 1562-6016 PACS: 28.41.Qb https://nasplib.isofts.kiev.ua/handle/123456789/80541 Overview of results obtained within the framework of the STCU Project #294 is given Наведено огляд результатів, отриманих в рамках проекту УНТЦ № 294. Приведен обзор результатов, полученных в рамках проекта УНТЦ № 294 This research was partially supported by Science & Technology Center in Ukraine (STCU) within the framework of Project # 294 en Національний науковий центр «Харківський фізико-технічний інститут» НАН України Вопросы атомной науки и техники Study of materials for reactors employing molten fluoride salts or PB-BI coolant using an electron irradiation test facility Дослідження матеріалів для рідкосольових реакторів та реакторів з Pb-Bi охолоджувачем за допомогою електронного опромінювального випробувального стенду Исследование материалов для жидкосолевых реакторов и реакторов с Pb-Bi охладителем с помощью электронного облучательного испытательного стенда Article published earlier |
| spellingShingle | Study of materials for reactors employing molten fluoride salts or PB-BI coolant using an electron irradiation test facility Azhazha, V.M. Bakai, A.S. Gurin, I.V. Neklyudov, I.M. Omelchuk, A.A. Zelenskiy, V.F. Garner, F.A. |
| title | Study of materials for reactors employing molten fluoride salts or PB-BI coolant using an electron irradiation test facility |
| title_alt | Дослідження матеріалів для рідкосольових реакторів та реакторів з Pb-Bi охолоджувачем за допомогою електронного опромінювального випробувального стенду Исследование материалов для жидкосолевых реакторов и реакторов с Pb-Bi охладителем с помощью электронного облучательного испытательного стенда |
| title_full | Study of materials for reactors employing molten fluoride salts or PB-BI coolant using an electron irradiation test facility |
| title_fullStr | Study of materials for reactors employing molten fluoride salts or PB-BI coolant using an electron irradiation test facility |
| title_full_unstemmed | Study of materials for reactors employing molten fluoride salts or PB-BI coolant using an electron irradiation test facility |
| title_short | Study of materials for reactors employing molten fluoride salts or PB-BI coolant using an electron irradiation test facility |
| title_sort | study of materials for reactors employing molten fluoride salts or pb-bi coolant using an electron irradiation test facility |
| url | https://nasplib.isofts.kiev.ua/handle/123456789/80541 |
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