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 Research fast reactor BOR-60 is one of the leading experimental facilities of the country and of the world intended experimental facilities of the country and of the world intended for testing of a variety of fuel, absorbing and structural materials for testing of a variety of fuel, absorbing and structural materials that are offered for creation of advanced fast, pressurized that are offered for creation of advanced fast, pressurized water, gas-cooled and fusion reactors and serving for water, gas-cooled and fusion reactors and serving for substantiation of the VVER and BN-type reactor service life substantiation of the VVER and BN-type reactor service life extension. The reactor has been in effective and reliable extension. The reactor has been in effective and reliable operation for more than 35 years already and at present it is operation for more than 35 years already and at present it is practically the only research fast reactor that, apart from well practically the only research fast reactor that, apart from well equipped material science laboratories and pilot-scale equipped material science laboratories and pilot-scale production engaged in fuel fabrication and reprocessing, has production engaged in fuel fabrication and reprocessing, has unique experimental possibilities for complex investigation unique experimental possibilities for complex investigation activities in different research lines.activities in different research lines.

Table 1Some physical characteristics of the reactor

Table 1Some physical characteristics of the reactor

Characteristic Value

Reactor heat power, MW 60

Inlet temperature of coolant, С 310-330

Outlet temperature of coolant, С 530

Fuel UO2 or UO2-PuO2

235U enrichment, % 45-90

Maximum Pu concentration, % 40

Maximum volumetric power in the core, kW/l 1100

Maximum neutron flux density, cm-2·s-1 3.7·1015

Average neutron energy, MeV 0.4

Neutron fluence per 1 year, cm-2 3·1022

Damage dose accumulation rate, dpa/y Up to 25

Fuel burnup rate, %/y Up to 6

Power non-uniformity factors: AxialRadial

1.141.15

Fig. 1. Simplified schematic diagram of the BOR-60 reactor facilityFig. 1. Simplified schematic diagram of the BOR-60 reactor facility

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. 2. The BOR-60 reactor sectionFig. 2. The BOR-60 reactor section

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

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. 3. Pressure plenum

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. 4. Cartogram of the BOR-60 reactorFig. 4. Cartogram of the BOR-60 reactor

Reactor loading possibility

Cells quantity for S/A for absorbing rods instrumented cells

265156

73

State S/A quantity 85-124

Maximum quantity of the experimental non-fuel S/A in the core

12

Maximum quantity of the experimental fuel S/A in the core

156

cora S/A screen S/A

ZrH ronfuel S/A

control rod instrumented cell

Fig. 5. Radial distribution of average neutron energy (En), integral energy (Fn) and neutron flux density with Е>0.1 Mev (Fn(0.1))

0.0E+00

5.0E+14

1.0E+15

1.5E+15

2.0E+15

2.5E+15

3.0E+15

3.5E+15

0 1 2 3 4 5 6 7 8 9

Ряд

Пл

от

но

сть

по

то

ка

ней

тр

он

ов

,

см

-2с-1

.

0

50

100

150

200

250

300

350

Ср

едн

яя

эн

ерги

я н

ей

тр

он

ов

, к

эВ

Fn, MCU

Fn(0.1), MCU

En, M CU

Neu

tron

flu

x de

nsit

y, s

m-2

s-1

Ave

rage

neu

tron

ene

rgy,

keV

Layer

Fig. 6. Neutron spectrum of the BOR-60 reactor core - layer (cell number)

1.E-07

1.E-06

1.E-05

1.E-04

1.E-03

1.E-02

1.E-01

1.E+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

Е, эв

Зн

ач

ен

ие

, о

тн.е

д.

1 (Б31)

2 (Б39)

3 (Б43)

4 (В05)

5 (В11)Val

ue, r

elat

ive

unit

s

E, eV

Fig. 7. Neutron spectrum of the BOR-60 reactor reflector – layer

(cell number)

1.E-04

1.E-03

1.E-02

1.E-01

1.E+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, эВ

Зн

ач

ен

ие

, о

тн.е

д.

6 (В19)

7 (В24)

8 (В25)

9 (Г01+ZrH)

Val

ue, r

elat

ive

unit

s

E, eV

- 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 2Testing conditions of materials and products in cell D-23

Table 2Testing conditions of materials and products in cell D-23

Parameter Value

Neutron flux density, sm-2·s-1 2·1015

Specific radiation energy release in structural materials (with atomic number Z = 2630), W/g

4

Absorbed gamma-radiation dose rate, Gy/s 4.5·103

Coefficient of non-uniform radiation density distribution along the core height (450 mm):

for neutronsfor gamma-radiation

1.131.25

Sodium flow rate, m3/h:when fed from high pressure chamberwhen fed from low pressure chamber

up to 8up to 2

Table 3Neutron-physical characteristics of the BOR-60 instrumented

cells (Wreactor=55 MW)

Table 3Neutron-physical characteristics of the BOR-60 instrumented

cells (Wreactor=55 MW)

1 year of irradiation - WT≈ 250 000 MW×h, Kz and Kr – axial and radial non-uniformity coefficient.1 year of irradiation - WT≈ 250 000 MW×h, Kz and Kr – axial and radial non-uniformity coefficient.

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)3.42.8

3.12.5

2.52.0

1.20.6

Damage accumulation rate in steel (DPA), 10-6 d.p.a./s 1.4 1.3 1.0 0.2

Kz(AP), relative unit

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

Kr(CCP), relative unitF0 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

Neutron fluence, 1022 sm-2E>0.0 MeV 5.5 5.0 4.1 1.9

E>0.1 MeV 4.6 4.1 3.3 1.0

Steel damage dose, d.p.a. 24 21 17 4

Fig. 8. Neutron spectrum of cell G01 of the BOR-60 reflector Fig. 8. Neutron spectrum of cell G01 of the BOR-60 reflector

0.00

0.05

0.10

0.15

0.20

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, эВ

Зн

ачен

ие

, отн

.ед

.

(Г01+B4C)

(Г01)

(Г01+ZrH)

Val

ue, r

elat

ive

unit

s

E, eV

Fig. 9. BOR-60 HEC and VEC location schemeFig. 9. BOR-60 HEC and VEC location scheme Fig. 9. BOR-60 HEC and VEC location schemeFig. 9. BOR-60 HEC and VEC location scheme

1 - HEC,2 - sand,3 - oxide,4 – disperser drive,5 – cast iron,6 - graphite,7 - concrete,8 - VEC

1 - HEC,2 - sand,3 - oxide,4 – disperser drive,5 – cast iron,6 - graphite,7 - concrete,8 - VEC

Fig. 10. Neutron spectrum of the BOR-60 VEC (calculations were made on the basis of MMK and OKS-ROZ-6 programs)

Fig. 10. Neutron spectrum of the BOR-60 VEC (calculations were made on the basis of MMK and OKS-ROZ-6 programs)

1.E-05

1.E-04

1.E-03

1.E-02

1.E-01

1.E+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, эВ

Зн

ач

ен

ие

, отн

.ед

.

ММК

ОКС-РОЗ-6

ОКС-РОЗ-6 (3см-ZrH2)

Val

ue, r

elat

ive

unit

s

E, eV

0.00

0.10

0.20

0.30

0.40

0.50

1.0E-01 1.0E+00 1.0E+01 1.0E+02 1.0E+03 1.0E+04 1.0E+05 1.0E+06 1.0E+07

En, эВ

Отн

оси

тел

ьно

е зн

ачен

ие

Вход ГЭК

Выход ГЭК

Выход Pb

Val

ue, r

elat

ive

unit

s

E, eV

HEC inlet

HEC outlet

Pb yield

Fig.11. Neutron spectrum at the HEC inlet and outlet Fig.11. Neutron spectrum at the HEC inlet and outlet

Table 4Neutron flux and gamma-quantum density at the BOR-60 HEC outlet (sm-2s-1)

Table 4Neutron flux and gamma-quantum density at the BOR-60 HEC outlet (sm-2s-1)

HEC Calculated value

Experiment value

Fn F Fn

Without Pb-screen, Еn>0 MeV (0.841.2)1010 9.6108 (2.93.4)108

Without Pb-screen, Еn>1.2 MeV (6.28.6)107 - (5.76.5)107

With Pb-screen 3.6109 2.9106 -

Fig. 12. Typical diagram of the BOR-60 reactor operation

0 1 2 3 4 5 6 7 8 9 10 11 12 Month0 1 2 3 4 5 6 7 8 9 10 11 12 Month

0 1 2 3 4 5 6 7 8 9 10 11 12 Month

60 N,

МВт 50

40

30

20

10

Long term investigation of neutron physical, heat-Long term investigation of neutron physical, heat-hydraulic and dynamic reactor characteristics allowed hydraulic and dynamic reactor characteristics allowed detailed study of reactor behavior in different detailed study of reactor behavior in different operation modes, creating a complex of computation operation modes, creating a complex of computation programs for reactor operation and experiments programs for reactor operation and experiments performance. As a result, calculations authenticity performance. As a result, calculations authenticity increased to support experimental programs, reactor increased to support experimental programs, reactor operation and its safety substantiation. On the basis operation and its safety substantiation. On the basis of great experience of reactor characteristics of great experience of reactor characteristics investigations and a verified complex of computation investigations and a verified complex of computation programs different methods were developed that programs different methods were developed that enable high accuracy control of operation modes and enable high accuracy control of operation modes and parameters of materials irradiation in the non-parameters of materials irradiation in the non-instrumented reactor cells. instrumented reactor cells.

Table 5 Irradiation parameters errors, %

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.

Fig. 13. Flow experimental assemblies Fig. 14. Device with evaporativewith gas heat insulation: thermosiphon:

1,2 – outer and inner bodies;

3 – clearance;4 – shells;5 - specimens

1,2 – outer and inner bodies;

3 – clearance;4 – shells;5 - specimens

1 – specimens;2 – shell;3 – heater;4 – body

1 – specimens;2 – shell;3 – heater;4 – body

6

7

5

1575

310

761

2

9 300

8

10

3

3

1

4

Fig. 15. Dismountable assembly with a hot probe for irradiation of

structural materials

1 - thermometric probe; 6 - gas clearance;

2 - detachable head; 7 - inner pipe;3 - spacer tubes; 8 - capsule assembly;4 - probe thermocouples; 9 - core center;5 - wrapper; 10 - fixture

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.

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.

Fig.16. Cross-section of the experimental device with capsules for irradiation of vanadium in lithium medium

1, 2 – leak-tight capsules with different type specimens in lithium-4 medium;

3 – inner capsule cladding from Inconel-type 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

1, 2 – leak-tight capsules with different type specimens in lithium-4 medium;

3 – inner capsule cladding from Inconel-type 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. 17. Scheme of sodium boiling generator

1 - fixture 2 - throttling orifice 3 - filter 4 - block of tungsten rods 5 - gas clearance 6 - block of steel rods 7 - head

1 - fixture 2 - throttling orifice 3 - filter 4 - block of tungsten rods 5 - gas clearance 6 - block of steel rods 7 - head

Fig. 18. Scheme of the instrumented nozzle

1 – fuel assembly2 - nozzle body3 – tube with sensors4 – flow regulator5 – sodium vapor filter6 – electric engine

1 – fuel assembly2 - nozzle body3 – tube with sensors4 – flow regulator5 – sodium vapor filter6 – electric engine

Fig. 19. Scheme of the capsule loop with the MGD-pump

1 – sodium vapor catcher2- level gauges inside of the channel3 – maximum sodium level in the channel4- KGO pipe5 – sodium flow regulator6 – sodium yield from electromagnetic pump7 – MGD pump8 – fuel assembly body9 – sodium upflow in the channel 10 – sodium down flow in the channel 11- upflow of reactor sodium12 – heat insulating gas clearance of FA in the channel13 – channel body14 - neutron sensors15 – inner wrapper of the channel16 – fuel elements17 – membrane18 – tube for sodium channel filling19 – throttling orifice20 – channel tail21 – inlet of reactor sodium into the channel from the BOR-60 high pressure chamber22 – protective membrane1-8 – thermocouples

1 – sodium vapor catcher2- level gauges inside of the channel3 – maximum sodium level in the channel4- KGO pipe5 – sodium flow regulator6 – sodium yield from electromagnetic pump7 – MGD pump8 – fuel assembly body9 – sodium upflow in the channel 10 – sodium down flow in the channel 11- upflow of reactor sodium12 – heat insulating gas clearance of FA in the channel13 – channel body14 - neutron sensors15 – inner wrapper of the channel16 – fuel elements17 – membrane18 – tube for sodium channel filling19 – throttling orifice20 – channel tail21 – inlet of reactor sodium into the channel from the BOR-60 high pressure chamber22 – protective membrane1-8 – thermocouples

Cross-section of the loop channel core centerCross-section of the loop channel core center

Fig. 20. Scheme of the lead loopFig. 20. Scheme of the lead loop

7 8

3 4

1

d)

c)

b) b)

5

6

1

2

3

4

5

6

7

8 9 10

11

12

13 14

2 3

9

a)

15

16

ILCC cross-section in the core central plane

41*0,5

35*0,6

37,5*0,6

9,4*0,5 (4 шт.)

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 a) to gas rig b) lead level c) Na output d) Na input

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 6Reactor materials tested in the BOR-60 reactor

Material Type

Fuel

CeramicsUO2, UO2-PuO2, UC, UN, UPuN,

UPuCN

Metal U, UPu, UpuZrNb

Ceramal U-PuO2, UO2-U, UN-U

AbsorbingSamples

Ta, Hf, Dy, Sm, Gd, AlB6, AlB12,

EuO3

CPS rods CrB2, B4C, Eu2O3, Eu2O3+H2Zr

Structural

Stainless steelsOX18H9, X18H10T, ЭП-450, ЭП-823 03Х16Н9М2, ЭП-912, ЭИ-847, 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

GraphitesГРП-2-125, МП6-6, ГР-280, АРВ, IG-11, ПГИ

Material Type

Electrotechnical Insulation Al2О3, SiO2, Si, mica

Cables КТМС, КНМС(Н)

Magnets ЮНДК

Others Special ceramics ГБ-7, ИФ-46, ЦТС, LiNbO3

Biological shielding materials

Concretes

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,) 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.

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.

Plans for future reactor facility operationPlans for future reactor facility operation

Thank you for attention!