Experimental possibilities of research fast
reactor BOR-60
Efimov V.N., Zhemkov I.Yu., Korolkov A.S.
FEDERAL STATE UNITARY ENTERPRISE STATE SCIENTIFIC CENTER
OF RUSSIAN FEDERATION RESEARCH INSTITUTE OF ATOMIC REACTORS
Research fast reactor BOR-60 is one of the leading
experimental facilities of the country and of the world intended
for testing of a variety of fuel, absorbing and structural materials
that are offered for creation of advanced fast, pressurized water,
gas-cooled and fusion reactors and serving for substantiation of
the VVER and BN-type reactor service life extension. The
reactor has been in effective and reliable operation for more
than 35 years already and at present it is practically the only
research fast reactor that, apart from well equipped material
science laboratories and pilot-scale production engaged in fuel
fabrication and reprocessing, has unique experimental
possibilities for complex investigation activities in different
research lines.
Table 1
Some physical characteristics of the reactor
Characteristic
Reactor heat power, MW
Inlet temperature of coolant, С
Outlet temperature of coolant, С
Fuel
235U enrichment, %
Maximum Pu concentration, %
Maximum volumetric power in the core, kW/l
Maximum neutron flux density, cm-2·s-1
Average neutron energy, MeV
Neutron fluence per 1 year, cm-2
Value
60
310-330
530
UO2 or UO2-PuO2
45-90
40
1100
3.7·1015
0.4
3·1022
Damage dose accumulation rate, dpa/y
Up to 25
Fuel burnup rate, %/y
Up to 6
Power non-uniformity factors:
Axial
Radial
1.14
1.15
1 - reactor;
2 - intermediate heat
exchanger;
3 - circulating pump of the
first circuit;
4 - steam generator;
5 - sodium-air heat
exchanger;
6 - circulating pump of the
second circuit;
7 - blow fan;
8 - turbine;
9 - turbine condenser;
10 - deaerator;
11 - condensate pumps;
12 - feed pumps;
13 - low pressure heater;
14-high pressure heater
Fig. 1. Simplified schematic diagram of the BOR-60 reactor facility
1 – inlet branch pipe,
2 – high pressure chamber,
3 – basket,
4 – thermal and neutron reactor
vessel shielding,
5 – protective casing,
6 – support flange,
7 – refueling channel,
8 – driving mechanism of the
control and safety rods,
9 – support flange,
10 – large rotating plug,
11 – small rotating plug,
12 – core and reflector
assemblies
Fig. 2. The BOR-60 reactor section
1 – pressure plenum chamber;
2 – throttle plug;
3 – adjustable plug;
4 – inlet chamber;
5 – inlet chamber bottom;
6 - throttle;
7 - throttle;
8 - throttle;
9 - gasket;
10 – shell with displacers;
11 - displacer
Fig. 3. Pressure plenum
Reactor loading possibility
Cells quantity
for S/A
for absorbing rods
instrumented cells
265
156
7
3
State S/A quantity
85124
Maximum quantity of the
experimental non-fuel S/A in
the core
instrumented
cell
control rod
cora S/A
screen
S/A
ZrH
ronfuel S/A
Maximum quantity of the
experimental fuel S/A in the
core
Fig. 4. Cartogram of the BOR-60 reactor
12
156
3.5 E+ 15
350
F n, M C U
300
2.5 E+ 15
250
E n , M CU
2.0 E+ 15
200
1.5 E+ 15
150
1.0 E+ 15
100
5.0 E+ 14
50
0.0 E+ 00
0
см -2с -1.
-2 -1
Neutron
flux density,
s
Плотность
потокаsm
нейтронов,
F n(0.1 ), M CU
0
1
2
3
4
5
6
7
8
9
Layer
Ряд
Fig. 5. Radial distribution of average neutron energy (En), integral
energy (Fn) and neutron flux density with Е>0.1 Mev (Fn(0.1))
Средняя
энергия
нейтронов,
кэВ
Average
neutron
energy,
keV
3.0 E+ 15
1.E+ 00
Value, relative units
Значение , отн.ед .
1.E-01
1.E-02
1 (Б 31)
1.E-03
2 (Б 39)
3 (Б 43)
1.E-04
4 (В 05)
1.E-05
5 (В 11)
1.E-06
1.E-07
1.E -01
1.E +0 0
1.E+ 01
1 .E+ 02
1 .E+ 03
1.E +04
1.E+ 05
1.E +06
1 .E+ 07
эв
E,Е,eV
Fig. 6. Neutron spectrum of the BOR-60 reactor core - layer
(cell number)
Value,З relative
начениеunits
, отн.ед .
1.E + 00
1.E-01
1.E-02
6 (В 19)
7 (В 24)
8 (В 25)
1.E-03
9 (Г 01+ ZrH )
1.E-04
1.E -01
1.E+0 0
1.E+ 01
1 .E + 02
1.E +03
1.E+ 04
1.E + 05
1.E + 06
1.E+07
, эВ
E,EeV
Fig. 7. Neutron spectrum of the BOR-60 reactor reflector – layer
(cell number)
- For instrumented irradiation a special thermometric channel is used
allowing allocating experimental devices directly in the core (D23). The lower
part of the experimental device looks like a standard S/A (a fixture and a
hexagonal tube of 44 mm of “across flats dimension”).
- In two cells (А43 and D35) it is possible to display limited data
(thermocouples, neutron sensors, etc.).
- Peripheral cell G01 of the reflector is shielded by three assemblies with
zirconium hydride that allowed mitigating the cell neutron spectrum and
using it for radioisotope production and other purposes.
- The reactor is equipped with a horizontal (HEC) and 9 vertical (VEC)
channels outside of the reactor vessel. The channels are used mainly for
irradiation of electro technical materials and silicon radiation doping. By the
results of the HEC neutron physical characteristics study it was concluded
that the channel can be used for medical investigations.
Table 2
Testing conditions of materials and products in cell D-23
Parameter
Neutron flux density, sm-2·s-1
Specific radiation energy release in structural
materials (with atomic number Z = 2630), W/g
Absorbed gamma-radiation dose rate, Gy/s
Coefficient of non-uniform radiation density
distribution along the core height (450 mm):
for neutrons
for gamma-radiation
Sodium flow rate, m3/h:
when fed from high pressure chamber
when fed from low pressure chamber
Value
2·1015
4
4.5·103
1.13
1.25
up to 8
up to 2
Table 3
Neutron-physical characteristics of the BOR-60 instrumented
cells (Wreactor=55 MW)
Cell, row
Е31, 1
А43, 3
D23, 5
D35, 8
Radius of the cell center location against the core
center, mm
45
135
196
360
Neutron flux density, 1015 sm-2s-1:
-E>0.0 MeV (F0)
-E>0.1 MeV (F0.1)
3.4
2.8
3.1
2.5
2.5
2.0
1.2
0.6
Damage accumulation rate in steel (DPA), 10-6 d.p.a./s
1.4
1.3
1.0
0.2
F0
1.15
1.16
1.15
1.12
F0.1
1.17
1.17
1.17
1.15
DPA
1.18
1.18
1.18
1.16
F0
1.00
1.05
1.09
1.13
DPA
1.01
1.06
1.11
1.31
Neutron flux density fraction with energy exceeding 0.1
MeV, relative unit
0.83
0.82
0.80
0.50
Average neutron energy, keV
350
320
250
40
E>0.0 MeV
5.5
5.0
4.1
1.9
E>0.1 MeV
4.6
4.1
3.3
1.0
24
21
17
4
Kz(AP), relative unit
Kr(CCP), relative unit
Neutron fluence, 1022 sm-2
Steel damage dose, d.p.a.
1 year of irradiation - WT≈ 250 000 MW×h, Kz and Kr – axial and radial non-uniformity coefficient.
0.20
Value,
relativeотн.ед.
units
Значение,
(Г01+B4C)
(Г01)
0.15
(Г01+ZrH)
0.10
0.05
0.00
1.E-01
1.E+00
1.E+01
1.E+02
1.E+03
1.E+04
1.E+05
1.E+06
1.E+07
E,E,eV
эВ
Fig. 8. Neutron spectrum of cell G01 of the BOR-60 reflector
1 - HEC,
2 - sand,
3 - oxide,
4 – disperser drive,
5 – cast iron,
6 - graphite,
7 - concrete,
8 - VEC
Fig. 9. BOR-60 HEC and VEC location scheme
1.E+00
Value,
units
relative
Значение,
отн.ед.
1.E-01
1.E-02
ММК
1.E-03
ОКС-РОЗ-6
ОКС-РОЗ-6 (3см-ZrH2)
1.E-04
1.E-05
1.E-01
1.E+00
1.E+01
1.E+02
1.E+03
1.E+04
1.E+05
1.E+06
1.E+07
эВ
E,E, eV
Fig. 10. Neutron spectrum of the BOR-60 VEC (calculations were made
on the basis of MMK and OKS-ROZ-6 programs)
Value,
relative units
О тносительное
значение
0.50
В ходinlet
ГЭ К
HEC
0.40
В ых од
Г ЭК
HEC
outlet
В ых
од Pb
Pb
yield
0.30
0.20
0.10
0.00
1.0E-01
1.0E+00 1.0E+ 01 1 .0E+02 1.0E+0 3 1.0E+ 04 1.0 E+05 1.0E+06 1.0E +07
E,
En,eV
эВ
Fig.11. Neutron spectrum at the HEC inlet and outlet
Table 4
Neutron flux and gamma-quantum density at the BOR-60 HEC outlet (sm-2s-1)
HEC
Experiment
value
Calculated value
Fn
Fg
Fn
Without Pb-screen, Еn>0 MeV
(0.841.2)1010
9.6108
(2.93.4)108
Without Pb-screen, Еn>1.2 MeV
(6.28.6)107
-
(5.76.5)107
With Pb-screen
3.6109
2.9106
-
60
N,
МВт
50
40
30
20
10
00
11
22
33
44
55
66
77
88
12
9 9 10 10 11 1112
Month
Month
Fig. 12. Typical diagram of the BOR-60 reactor operation
Long term investigation of neutron physical,
heat-hydraulic and dynamic reactor characteristics
allowed detailed study of reactor behavior in
different operation modes, creating a complex of
computation programs for reactor operation and
experiments performance. As a result, calculations
authenticity increased to support experimental
programs, reactor operation and its safety
substantiation. On the basis of great experience of
reactor characteristics investigations and a verified
complex of computation programs different
methods were developed that enable high
accuracy control of operation modes and
parameters of materials irradiation in the noninstrumented reactor cells.
Table 5
Irradiation parameters errors, %
Parameter
Measurement Calculation
Intel (outlet) reactor
temperature
1,2
-
Reactor power
-
2,5
Reactor flow rate
3
-
Neutron flux (fluence)
7
10
Experimental devices (ED)
flow rate
2
-
ED power
7
10
Intel ED temperature
1
1,5
Outlet ED temperature
1
1-3
•
For irradiation of a variety of materials and products at different operation
modes and parameters a complex of specialized test devices is used. The test
devices consist of capsule devices, dismountable material science assemblies,
autonomous instrumented channels, special instrumented S/As etc.
•
Simple design of the devices and possibility to install them practically into
any core or reflector cell can be considered an undoubted advantage of the
devices.
The main task the developer of the test devices faces is creation of the
required temperature modes at the specimens. For this purpose thermal
insulating clearances, intensive cooling or additional heating due to radiation
energy release or fuel fission are used. Temperature stabilization is achieved
as a result of thermistor change in the scheme of heat transfer due to the
coolant temperature change or as a result of the heat removal intensification by
using liquid metal under boiling condition. These devices help to provide the
specified axial and azimuthal temperature non-uniformity.
1 – specimens;
2 – shell;
3 – heater;
4 – body
Fig. 13. Flow experimental assemblies
with gas heat insulation:
1,2 – outer and inner
bodies;
3 – clearance;
4 – shells;
5 - specimens
Fig. 14. Device with evaporative
thermosiphon:
1
2
3
4
5
6
7
8
1575
9
300
761
3
10
310
The lower boundary of the irradiation temperature range
that is ensured in the BOR-60 reactor makes up 300-310оС. It
significantly expands the scope of reactor work, including
experiments on investigation of physical and mechanical
properties of zirconium alloys and materials of the VVER-type
reactor internals. At relatively high coolant flow rate the
dismountable assembly allows irradiating structural materials
specimens at the temperature close to the reactor inlet
temperature. This assembly is one of the simplest and widely
used experimental devices helping to perform intermediate
reloading procedures and investigation of specimens with their
subsequent irradiation. The dismountable assembly is also
used for irradiation of fuel elements.
1 - thermometric probe;
2 - detachable head;
3 - spacer tubes;
4 - probe thermocouples;
5 - wrapper;
6 - gas clearance;
7 - inner pipe;
8 - capsule assembly;
9 - core center;
10 - fixture
Fig. 15. Dismountable assembly with a hot probe for irradiation of
structural materials
1, 2 – leak-tight capsules with different
type specimens in lithium-4 medium;
3 – inner capsule cladding from Inconeltype heat-resistant steel;
4 – outer capsule cladding from stainless
steel;
5 – ampoule sodium;
6 – ampoule clearance;
7 – leak-tight wrappers from Inconel-type
heat-resistant steel with
thermocouples
Fig.16. Cross-section of the experimental device with capsules for
irradiation of vanadium in lithium medium
1 - fixture
2 - throttling orifice
3 - filter
4 - block of tungsten rods
5 - gas clearance
6 - block of steel rods
7 - head
Fig. 17. Scheme of sodium boiling generator
1 – fuel assembly
2 - nozzle body
3 – tube with sensors
4 – flow regulator
5 – sodium vapor filter
6 – electric engine
Fig. 18. Scheme of the instrumented nozzle
1 – sodium vapor catcher
2- level gauges inside of the channel
3 – maximum sodium level in the channel
4- KGO pipe
Cross-section of the loop channel core center
5 – sodium flow regulator
6 – sodium yield from electromagnetic pump
7 – MGD pump
8 – fuel assembly body
9 – sodium upflow in the channel
10 – sodium down flow in the channel
11- upflow of reactor sodium
12 – heat insulating gas clearance of FA
in the channel
13 – channel body
14 - neutron sensors
15 – inner wrapper of the channel
16 – fuel elements
17 – membrane
18 – tube for sodium channel filling
19 – throttling orifice
20 – channel tail
21 – inlet of reactor sodium into the channel
from the BOR-60 high pressure chamber
22 – protective membrane
1-8 – thermocouples
Fig. 19. Scheme of the capsule loop with the MGD-pump
9
14
13
12
11
a)
15
8 b)
7 b)
6 10
9
8
16
c)
5
7
6
4
3
2
5
4
1- channel tail
2- flow meter (for sodium)
3- channel vessel
4- heat-insulating gas gap
5- FA
6- Electric heater
7- “lead-sodium” heat-exchanger
8- pump wheel
9- pump shaft
10- oxygen sensor (2 pc.)
11- hydrogen supply tube
12- magnetic clutch
13- pump electric drive
14- electric cables output
15- gas output tube
16- gas input tube
Thermocouples:
1- lead-fuel pins input
2- lead-downcomer section input
3- lead-fuel pins output 1
4- lead-fuel pins output 2
5- sodium output
6- lead in upper part of loop
7- temperature of oxygen sensor
8- gas cavity
9- electric motor surface
 9,4*0,5 (4 шт.)
 35*0,6
 37,5*0,6
 41*0,5
ILCC cross-section in the core central plane
1
3
a)
b)
c)
d)
to gas rig
lead level
Na output
Na input
2
1
d)
Fig. 20. Scheme of the lead loop
Main directions of investigation
- Study of safety issues. A series of experiments on
substantiation of fast sodium reactor safety was performed. Among
them are: feeding of gas into the core, sodium boiling, blocking of
coolant flow in the experimental FA resulting in fuel elements damage,
intercircuit leaks in steam generators etc. Detailed study of different
normal and off-normal processes at the BOR-60 reactor allowed testing
and adjusting of methods and means of abnormities diagnostics.
- Testing of fuel, absorbing and structural materials. Irradiation
programs are paid special attention to, among them:
•
Mass testing of fuel elements and fuel assemblies up to the burn
up of 30% h.a. under steady-state and transition conditions;
•
Testing of different neutron absorbing materials;
•
Radiation testing of structural reactor materials;
•
Testing of electric insulating, magnetic and refractory materials for
fussion reactors;
• Investigations in radiation material science:
Determination of deformation, long-term strength and fracture
toughness dependence at temperature of 320-1000оС up to the
dose of 200 dpa;
• Study of the technology of long-lived radionuclides transmutation
and burning out from spent fuel of different reactors;
• Radiation silicon alloying for radio electronics.
In 1981 fuel elements with vibropacked fuel columns on the basis of
power-generated plutonium were applied for the reactor core for the
first time. Positive results of mass testing of fuel elements with
vibropacked uranium-plutonium oxide fuel in the BOR-60 reactor up
to the burn up of more than 30%, as well as of 6 experimental fuel
assemblies up to the burn up of 9,6% in the BN-600 reactor can
serve a real basis for large-scale experiments in fast power reactors
to increase their efficiency and to enhance their safety.
Testing of fuel elements containing weapon grade
plutonium-based fuel was started in 1998.
In the frame of the program on development of closed fuel
cycle elements much is being done on burning out and
transmutation of plutonium and minor actinides (MA).
Design-experiment investigations and analysis of the
isotope content of microcapsules (40 pieces) with
different MA sets irradiated in the BOR-60 reactor were
performed. The obtained design-experiment results can
be used for adjustment of physical constants.
Results on investigation of different fuel compositions serve the basis
for development of a fuel cycle of advanced fast reactors with
enhanced safety. Among these is the BREST-OD-300 reactor with
lead coolant and nitride fuel.
The first stage of testing of BREST-OD-300 pilot fuel elements took
place at the BOR-60 reactor.
• Short-cut testing of different structural materials is carried out:
• Steels used for fabrication of vessel internals (VI) for VVER reactors;
• Zirconium alloys for VVER cores;
• Vanadium-based alloys in lithium medium for fusion reactors;
• Graphite for RBMK reactors.
Table 6
Reactor materials tested in the BOR-60 reactor
Material
Fuel
Absorbing
Structural
Type
Ceramics
UO2, UO2-PuO2, UC, UN, UPuN,
UPuCN
Metal
U, UPu, UpuZrNb
Ceramal
U-PuO2, UO2-U, UN-U
Samples
Ta, Hf, Dy, Sm, Gd, AlB6, AlB12,
EuO3
CPS rods
CrB2, B4C, Eu2O3, Eu2O3+H2Zr
Stainless steels
OX18H9, X18H10T, ЭП-450, ЭП823 03Х16Н9М2, ЭП-912, ЭИ847, ЭП-172, ЧС-68, ВХ-24
High-nickel alloys
РЕ-16, Х20Н45М4Б, ВЦУ
Refractory
materials
V, W, Mo, Nb
Zirconium alloys
Э-110, Э-635, Э-125
Graphites
ГРП-2-125, МП6-6, ГР-280, АРВ,
IG-11, ПГИ
Material
Electrotechnical
Others
Type
Insulation
Al2О3, SiO2, Si, mica
Cables
КТМС, КНМС(Н)
Magnets
ЮНДК
Special ceramics
ГБ-7, ИФ-46, ЦТС, LiNbO3
Biological
Concretes
shielding materials
Isotope accumulation for medical purposes
Taking into account physical peculiarities of a fast reactor,
commercial radionuclide accumulation parameters were investigated.
The radionuclides were produced by the threshold neutron reactions:
32P, 33P, 35S, 89Sr (reaction (n, р)) and 117mSn (reaction (n,n')). Besides,
indices of the 153Gd radionuclide accumulation process were also
determined. The radionuclide was produced by reaction of radiation
neutron capture (n,g) in the BOR-60 irradiation cells with specially
heated neutron spectrum. At present serial production of strontium-89
from yttrium targets (for production of “strontium-89 without carrier”
preparation) and gadolinium-153 from europium targets is realized for
production of sources and preparations.
Plans for future reactor facility operation
The BOR-60 reactor has been in operation for 35 years already, the
design service life makes up 20 years and calculated life is equal to 40 years.
Decision on possible reactor service life extension was made up taking into
account the equipment and materials state, strength of the equipment and
sodium circuit pipelines – these are the components that contribute much to
the reactor safety and that were fabricated in accordance with the current
calculation norms. Long-term plans concerning the above mentioned problems
are made for several decades.
There are plans on reactor reconstruction aiming at the reactor service life
extension for not less than 30 years in comparison with the calculated
resource. During reconstruction it is important to expand the reactor
experimental possibilities and to enhance its safety. A draft design of a new
reactor has been prepared already and at present design work on installation
of a new reactor within the operating reactor facility is being in process.
Thank you for attention!