JSC “State Scientific Center-RIAR”, Dimitrovgrad, Russia, 2010

International EU-Russia / CIS Conference: ”On Technologies of the Future: The Collaboration of Spain and CIS “, ( Madrid, Spain ,22-23 April 2010).

General information on research reactors and

techniques for In-pile and PIE tests of materials in RIAR

Research reactors provide for solving problems which Research reactors provide for solving problems which are of practical importance for nuclear power engineering are of practical importance for nuclear power engineering development, its safety ensuring, environmental development, its safety ensuring, environmental acceptability, efficiency and effectiveness.acceptability, efficiency and effectiveness.

The RIAR research reactors are used for examination The RIAR research reactors are used for examination of radiation resistance properties of a wide range of reactor of radiation resistance properties of a wide range of reactor materials in support of new technologies and engineering materials in support of new technologies and engineering solutions for modern and advanced reactors of a new solutions for modern and advanced reactors of a new generation. generation.

Research reactorsResearch reactors

Main reactor characteristics for material Main reactor characteristics for material testing and irradiationtesting and irradiation

300/530300/530NaNa

45/6045/60waterwater

(30-60)/(40-70)water

40/83water

50/95water

Coolant temperature(inlet/outlet,°?

ThermalThermal--

Fast Fast (E>0,1Mev)

2,85*10E152,85*10E15

Thermal1,5*10E14Fast (E>0,1Mev)

6,9*10E13

Thermal1,4*10E14Fast (E>0,1Mev)

5,6*10E13

Thermal5*10E14Fast (E>0,1Mev)2*10E14

Thermal5*10E15Fast (E>0,1Mev)2*10E15

MaximumNeutron flux

Cm-2 S-1

606010106100100100ThermalPower,MW

BOR-60RBT-10/1RBT-10/2Pool-type

RBT-6Pool-type

MIRLoop-type

SMParameter

Main reactor characteristics for material Main reactor characteristics for material testing and irradiation (continuation).testing and irradiation (continuation).

Up to 30Up to 30

From From 320320Up to Up to

12001200°°CC

~1,4

Up to900°?

~1,0

Up to900°?

Up to 3

300°?

Up to 34

From 100Up to

1600°?

MaximumValue of radiation damage

(for V), dpaDiapason of temperature

irradiation

5800580030003000

8000800048004800

80004800

6600660024002400

6700Up to 300

Operation timehours:-annual

-continual

BOR-60RBT-10/1RBT-10/2Pool-type

RBT-6Pool-type

MIRLoop-type

SMParameter

Main reactor characteristics for material Main reactor characteristics for material testing and irradiation (continuation).testing and irradiation (continuation).

BOR-60RBT-10/1RBT-10/2

RBT-6MIRLoop-type

SMParameter

4040Behind Behind corecore--up up to 300to 300

6060Up to 200Up to 200--250250

12012060Maximum Maximum diameter of diameter of capsule,mmcapsule,mm

45045035035035035010001000350Height of Height of core,core,mmmm

Inert gas,

liquid metal

Water,Water,Inert gasInert gas

Liquid metalLiquid metal

Water,Water,Inert gasInert gasLiquid Liquid metalmetal

Water,Water,Inert gasInert gas

Liquid metalLiquid metal

Water,Inert gas,

liquid metal

Environment Environment of irradiationof irradiation

THE CENTRAL HALL OF THE CENTRAL HALL OF SM SM AND AND RBTRBT--66REACTORSREACTORS

s?

RBT-6

1-core, 2 - partitioning, 3 - shielding,4 - specimens,5 - supporting plate to install a row of capsules,6 - guides, along which supporting plates with capsules move, 7 -1st type capsules with specimens,8 - communication outlets from the 1st type capsules,9 - 2nd type capsules with specimens,10 - communication outlets from the 2nd type capsule11 - concrete partitioning with a cylindrical niche,12 - block for out-of-vessel space simulation,13 - mobile platform

SSC RF RIAR (Russian Federation) EdF (France)

Fontenvro, 23-27 September, 2002

1 3 1 2 1 1

12 1 2 3 4 5 6 7 8 9 1 0

RBT-6 reactor and KORPUS facility

KORPUS facility

The RBT-6 reactor is provided with the KORPUS facility that was put intooperation in 1993. It is designed for irradiation of a great number of vessel steel specimens in a wide range of irradiation test parameters. The capsule channels of the RBT-6 reactor core incorporate the devices intended for irradiation of the ITER reactor component parts and units and the GT-MHR reactor fuel compacts and coated fuel particles. The capsule channels of reflector are provided with the devices intended for radiation resistance testing of sensors and measuring equipment parts. A certified background neutron field is available and maintained in one of the reflector channels. It is meant for neutron flux unit transfer in the operating neutron fields of the RIAR research reactors.

Distribution of density of a flow of neutrons on lines of ampoules in the KORPUS facility.

KORPUS facility

capsules

Reactor core

3

1

? ?

? ?

Section ?-?

Section ?-?

R

L

Θ

R3. Base Metal

(GOST)

7. HAZ (GOST)

1. Base Metal(RCC-M)

2. Weld(RCC-M)

5.Weld(GOST)

6.Weldtensilespecimens

4.Base Metaltensile specimens

Cut off diagram of specimens for mechanical tests by the Russian and French standards

SSC RF RIAR (Russian Federation) EdF (France)

Fontenvro, 23-27 September, 2002

Weld No. 4

Investigations block

THE VESSEL OF KALININ NPP

4-TH UNIT (VVER-1000/320) with investigation

block

Note: after the specimen is fed to the support, it is tested for less than 2 sec. In this series of

tests there was no supercooling or overheating of specimens.

10×10×55 mm

Dimensions of the specimen

±1 0?Accuracy of temperature control

-190 +500 0?

Range of test temperatures

40.00 mmDistance between the supports (EN -0 + 0.25 ASTM ±0.05mm )

-0.15 mmSymmetry of the hammer edge about the supports

(EN ±0.5 ASTM ±0.25 mm )

5.21 m/sSpeed at the moment of impact450 JStored energy

Main characteristics of the Roell AmslerRKP-450 testing machine (in “hot”cell)

SSC RF RIAR (Russian Federation) EdF (France)

Fontenvro, 23-27 September, 2002

DBTT shift due irradiation depends on the choice of

criterion levels of absorbed energy for base metal

specimens,

- non-irradiated specimens,approximating curve

- irradiated specimens, approximating curve

?) absorbed energy b) ductile fracture fraction

0

50

100

150

200

0

20

40

60

80

100

-160 -120 -80 -40 0 40 80 120 160 200 240

En

erg

y(

J)S

hea

rF

ract

ure

Ap

pea

ran

ce(%

)

Temperature (0C)

a

b

SSC RF RIAR (Russian Federation) EdF (France)

Fontenvro, 23-27 September, 2002

∆T49J=38 0?

∆T50% =33 0?

Temperature (0C)

0

50

100

150

200

-160 -120 -80 -40 0 40 80 120 160 200 240

0

20

40

60

80

100

-160 -120 -80 -40 0 40 80 120 160 200 240

En

erg

y(

J)

a

b

Sh

ear

Fra

ctu

reap

pea

ran

ce(%

)SSC RF RIAR (Russian Federation) EdF (France)

Fontenvro, 23-27 September, 2002

Weld metal specimens cut off to

GOST 6996-66,- non-irradiated

specimens,approximating curve

- irradiated specimens, approximating curve

?) absorbed energy b) ductile fracture fraction

∆T49J=54 0?

∆T50% =45 0?

∆T38J=62 0?

THE SM REACTORTHE SM REACTOR

Various temperature conditions, fluence accumulation and damage dose rates are created in the loop facilities and capsule channels of the SM reactor reflector as well as in capsules to irradiate structural and fuel materials. The developed irradiation devices are used for irradiation tests of fuel materials and fuel rods of research and high-temperature gas-cooled reactors, structural materials of power reactors as well as parts and units of fusion reactor.

Diagram of the SM reactor core

? -3 ? -1

9 12

46

56

66

76

86

96

65

75

45

55

85

95

42

52

62

72

82

92

41

51

61

71

81

44

54

84

94

43

53

83

93

??

-4??

-3

??

-191??

-2

? -22

6

14

15

3

7

816

? -4

? -5

??

-217

? -6

? -10

? -9

13

? -8

??

-1

19

4

10 ? -7

5

20

11

21

18

central blockcentral block

? ?? ? insertinsert

reflector blocksreflector blocks

centralcentral shimshimrodrod

Channel and its number

Technical characteristics of the SM-3 water loop facilities

3,71093,7109Maximum coolant activity (under emergency conditions), Bq/kg

0,43,5Primary coolant volume, m3

530Maximum coolant flow rate through the channel, m3/h

320,473,933Flux density with the neutron density of 0.1-10.5 MeV, 1013 cm-2s-1

816,32889Flux density with the neutron energy of 0-0.465 eV, 1013 cm-2s-1

41562Numbers of the reflector cells used for channel location

260-280??31020-60/?? 100Coolant inlet\outlet temperature, ??

18,55Coolant pressure, MPa

33Number of loop channels

VPVP--33VPVP--1 1 Loop facility

Kinds of in-pile examination in reactor SMIrradiation devices and test facilities are developed for SM capsule channels in order

to use the following techniques for examination purposes:• - In-pile creep and plasticity properties of claddings made of new materials for

fuel rods of power and transport reactors, reactor graphite in a wide temperature range of the irradiation test;

• - In-pile creep, relaxation properties, plasticity and swelling of WWER high-burn-up fuel, new fuel materials at specific fission rates in a wide temperature-power range and specific stressed states;

• - In-pile creep of fuel rods specimens and claddings preliminary irradiated under loading conditions and without them at different types of stressed states with due consideration for annealing kinetics of radiation hardening; possible changes in anisotropy and occurrence of dimensional changes, which are not a result of loading effect;

Relaxation properties of anisotropic materials with allowance made for application of the obtained results for WWER cladding tubes and spacer grids to examine radiation creep and cladding ability for stress level reduction when it comes in contact with fuel;

•

REACTOR SM( continuation)• - Kinetics of relaxation processes in WWER fuel assembly springs to measure self-

created stresses in the course of operation;• Creation of mathematical models for in-pile strain processes in structural materials

of various crystallographic groups as well as in standard and new nuclear fuel;

Long-term strength of cladding tubes under static and cycling conditions and external environment effect.

• Hydride effect on in-pile creep of the used and advanced zirconium cladding alloys to verify the developed model of radiation-thermal creep;

• In-pile creep of irradiated specimens and claddings preliminary irradiated under loading conditions and without them at the longitudinal tension and pressurized states in the temperature range from 300 to 600 ?? to modify the creep model with regard to normal and abnormal “dry” storage conditions of spent WWER fuel assemblies;

• • Iodine stress corrosion cracking of cladding tubes preliminary irradiated under pressure with a grown crack and without it in the SM loop channels in high quality water when fast neutron flux density (? >0.1 MeV) is up to 3? 1014 cm-2s-1 to develop a model of WWER FR cladding damage.

Mock-up of the ITER reactor divertor target

The SM reactor was used for testing of fusion reactor component parts and units under the temperature cycling conditions. Next fig. demonstrates the irradiation device together with the mock-up of the ITER reactor divertortarget. It was used in the course of ~1000 cycles when the maximum surface thermal flux density was about 8 MW/m2.

Testing device for ITR reactor divertor target mock-up

11--gas tube; 2gas tube; 2--electric drive; 3electric drive; 3--reducer; 4reducer; 4--thermocouplt thermocouplt outputs; 5outputs; 5--IR flange; 6IR flange; 6--tightening unit; 7tightening unit; 7--coupling rod; 8coupling rod; 8--sealing unit; 9, 13sealing unit; 9, 13--cooling tubes and fastening components, cooling tubes and fastening components, 1010--mockmock--up with beryllium plates; 11up with beryllium plates; 11--simulator; 12simulator; 12--graphite mockgraphite mock--up; 14up; 14--reactivity compensator reactivity compensator

11--vessel; 2,6vessel; 2,6--cooling tubes;cooling tubes;33--tungsten heater;tungsten heater;4,74,7--fastening components; 5fastening components; 5--mockmock--upup

? enter plane of the reactor core

Support of irradiation tests

The RIAR develops and manufactures special-purpose sensors, detectors and converters of physical quantities such as: mini-thermocouples, inductive and mutual induction pressure transducers, linear movement pickup, direct-charge transducers, gamma thermometers etc. Nomenclature of standard and special-purpose sensors used in a specific experiment defines the configuration and structure of communication device that forms a medium level of DAMS.

Irradiation rigs of the RBTIrradiation rigs of the RBT--6 reactor for radiation 6 reactor for radiation stability tests of instability tests of in--core sensors and detectors of core sensors and detectors of

reactor emissionreactor emission

“Neutron-8” facility for creep testing in reactor CM .

1-system for automatic control; 2-PC; 3-system for movement measurement; 4- unit for stepping motor control; 5,7 and 14-passive and active traction; 6-reactor channel; 8-heater; 9-lower grip; 10-stepping motor; 11-reducer; 12-ball-and screw pair; 13-scrwe nut; 15-force sensor; 16-junction unit; 17-strain sensors; 18-upper grip; 19-specimen; 20-gas-vacuum system, 21-temperature control system

Creep testing ZrCreep testing Zr--1%Nb.1%Nb.

ln ε ???

1000 / T , K-1

-18

-14

-10

-6

-2

2

1,1 1,5

?110 σ??? = 100 ? ? ?

????????????

? ? ?-60???-6

-221,3 1,7

Irrediatedln ε ??

?-16

-12

-8

-4

0

4

1,0 1,5 2,0 1000 / T , K-1

?110 σ??? = 300 ? ? ?

????????????

Irrediated

? ? ?-60

???-6

-20

MIR reactor

Ø The MIR reactor is used for implementation of several fuel testing programs for all types of reactors. The reactor loop facilities provide for maintenance and control of cooling conditions for experimental fuel assemblies, coolant water chemistry, continuous leakage testing of fuel rods as well as data acquisition, processing and display from irradiation device sensors and instrumented fuel rods. It is possible to load 4 m long fuel rods of full-size power reactor fuel assemblies in loop channels. Testing of RR fuel rods is carried out in the loop and operating reactor channels. Two hot cells are used for preparation and dismantling of irradiation devices as well as intermediate non-destructive examination of experimental fuel assemblies and fuel rods under irradiation at the MIR reactor building. There are devices providing for irradiation of fresh, short-size (refabricated) and full-size fuel rods equipped with temperature sensors, pressure gauges, gas release sensors, strain and elongation gages in the loop and operating reactor channels under the steady-state, transient, power ramping and maneuvering conditions as well as design-basis accident conditions.

Diagram of the MIR reactor core

2-9 3-17

4-24

1

2

3

4

5

6

7

89

10

11 12

13

14

1516

17

18

19

21 22

4-17 4-18 4-19 4-20 4-21

4-16 3-13 3-14 3-15 3-16 4-22

4-15 3-12 2-10 2-11 4-23

4-14 3-11 2-8 1-5 1-6 2-12 3-18

4-13 3-10 2-7 1-4 0-1 1-1 2-1 3-1 4-1

4-12 3-9 2-6 1-3 1-2 2-2 3-2 4-2

4-11 3-8 2-5 2-4 2-3 3-3 4-3

4-10 3-7 3-6 3-5 3-4 4-4

4-9 4-8 4-7 4-6 4-5

20 23

27

??-II

??-I

,

- operating channel;

- channel with additional loading;

- loop channel;

- capsule channel;

-beryllium block ;

-control rods: 1÷6-emergency; 7÷27-

shim.

MAIN TECHNICAL CHARACTERISTICS OF THE MIR LOOP FACILITIES

3.7×10103.7×1073.7×1073.7×1073.7×1073.7×107Activity level, Bq/kg

1,00,67514141616Flow rate, t/h

500500340340340340Temperature, ??

208,520202020Pressure, MPa

Maximum coolant parametersMaximum coolant parameters

water,steam

water,steam

water,steam

water,steamwaterwaterCoolant

112222Number of channels

PVPPVP--22PVPPVP--11PVKPVK--22PVKPVK--11PVPV --22PVPV--11Loop facility

MIR reactor

The change in fuel rod parameters and characteristics in the course of irradiation tests can be examined by signal processing of the in-core sensors installed in instrumented fuel assemblies and using mathematical simulation of processes in fuel assembly and fuel rods as well as transformations in sensors and transducers:

· - Kinetics of fission gas release in FR gas plenum when the power is increased and decreased as well as a function of cycles number;

· - Liner power level appropriate to fuel-cladding cohesion;· - Processes of linear and diametrical strains of the cladding as a function

of linear power level and a number of cycles as well as residual strain;· Fuel temperature as a function of linear power and fuel burnup;· - Heat-hydraulic parameters of coolant in the experimental fuel assembly;- Thermal characteristics of fuel rod (time constant, heat rate, thermal

conductivity and heat-transfer factor referred to volumetric heat capacity of fuel meat).

Schemes of the fuel rod locations in the different types of irradiation rigs installed in the MIR reactor

? ? ? ?-100062.262.2

6060

RFRRFRRFRRFR

Central Central plane of plane of the core the core

(CPC(CPC))Lower RFR (active part)

Upper Refabricatedfuel rod (active part)

square square �� 4242××11VVERVVER--440 and VVER440 and VVER--10001000

Fuel rods with a gas pressure sensor (?), thermal probe (b) FP release sensor (c)

?) ?)?)

????

?????????? ?????-

????????

??????????????? ??????? ?

??????? ????

???????

??????????? 5/20 ?

???????????? ?????

????

?????????? ?????-

????????

???? ????????????

???????????

linear displacement converter

sylphon

fuel rod

thermocouple

steel-zirconium adapter

sealing unit

thermocouple

steel-zirconium adapter

fuel rod

Layout of sensor for cladding diameter ?)and fuel rod length b) measurement.

?) ?)

? ? ????????????? ??????? ?

??????? ????

????? ????? ??????

??????? ?????? ? ? ?

????????????????? ?

?? ???

? ??? ???????????

??????

? ??????? ? ??????

a) b)

DTT coil

Fuel rod

Lever

Flexible joint

Flexible

junction

Stable junction

1. SpringShell

Fuel rod

Zirconium tube

Coupling rod

DTT coil

Relation of changes of RFR No75 length and diameter during testing including 1st and 2nd cycles

0

5

10

15

20

25

30

35

0.4 0.6 0.8 1.0 1.2 1.4 1.6

Elongation, mm

Dia

met

erch

ange

, µm

Change

prior to ECP

drop

2nd cycle

1st cycle

Change of length (1) and diameter of (2) RFR No75 during testing

0,0

0,2

0,4

0,6

0,8

1,0

1,2

1,4

1,6

90 96 102 108 114 120 126 132?????, ?

???

????

?? ?

????

, ??

0

4

8

12

16

20

24

28

32

???

????

?? ?

????

???

, ???

2

1

Diam

eter change, µm

Time, h

Len

gth

chan

ge, m

m

Change of the fuel rod length under cycling Change of the fuel rod length under cycling conditions (cycles form conditions (cycles form ?? 1 to 80) power 1 to 80) power rampramp

0,0

0,1

0,2

0,3

0,4

0,5

0,60,7

0,8

0,9

700 800 900 1000 1100 1200 1300 1400 1500??????????? ???????, ??

???

????

??, ?

?

- ?????????? ??? ?????

- ???? ???? ??? ?????

- ?????? ??? ?????

41 ????

50 70

80

Fuel temperature, oC

Elo

ngat

ion,

mm

Power increase

Power decrease

Power ramp

41st cycle

Estimation of the fuel temperature increase in RFR 72 versus the average burn-up increase for two levels of the average

linear power (1 – 200…210 W/cm, 2 – 170…180 W/cm 19 –30 campaigns of 2002.)

600

700

800

900

1000

1100

6 7 8 9 10 11 12 13

? ????? ???? ???????? ?? ???????, ? ? ?·???/??U

????

????

???,

? ?

1

2

Average burn-up increase, MW*day/kgU

Tem

pera

ture

, o C

Change of the FP partial pressure in the fuel rod Change of the FP partial pressure in the fuel rod reduced to a temperature of 25reduced to a temperature of 25 ???? (1); 2(1); 2--approximationapproximation

0

0.1

0.2

0.3

0.4

0.5

0 1 2 3 4 5 6 7 8 9? ? ? ? ? , ? ? ?

???

????

????

???

????

?, ?

??

2

1

Time, day

Part

ialp

ress

ure,

MP

a

BOR-60 reactor

In order to irradiate a great number of materials and products under different irradiation conditions and parameters, a set of special-purpose experimental devices which are comprised of capsules and dismountable assemblies loaded in the reactor core cells or radial blanket are used in the BOR-60 reactor. The reactor has got a special-purpose thermometric channel providing for experimental device location in the reactor core and irradiation test data output for irradiated materials through 30-50 communication lines. A number of autonomous instrumented capsuleloops and instrumented fuel assemblies was specially developed for it. These devices make it possible to test various advanced types of fuel and structural materials at high thermal power (100 kW/m), temperatures (1000 ?? ), fuel burnup levels (33% h.a.) and neutron fluences (1.81023cm-2

s-1, ? > 0.1 MeV).

Diagram of the BORDiagram of the BOR--60 reactor core60 reactor core

? ?2? 28

? ?1? 07

? 34

? 35 ? 38

? 29

? 20

? 43? 32? 33

? 28? 22? 26 ? 23? 30? 38

? 36? 40

? 31? 15? 12? 13? 17? 21? 10

? 43

? 39

? 31

? 05 ? 31? 39? 11? 19

? 07? 16? 25

? 35 ? 24

? 34

? 27

? 14

? 09

? 06

? 07

? 11

? 17

? 20? 18

? 16

? 19

? 21

? 26

? 23

? 29? 25? 24? 27? 30? 36

? 42

? 41

? 44

? 32

? 11? 21? 30? 40

? 07? 19? 27

? 10? 03? 04? 08? 14? 20? 29

? 01? 35? 34? 38? 40? 02? 09? 18? 35 ? 25

? 06? 16? 24? 34

? 38

? 01

? 10

? 03

? 04

? 08

? 40

? 42? 41? 44? 13

? 22

? 37

? 05? 17? 26? 36? 02? 14

? 03

? 04

? 36? 33? 37 ? 32? 43? 12

? 28

? 43

? 13? 23? 33? 42? 09? 20

? 38? 30? 26? 23? 28 ? 22? 22 ? 59? 39

? 15

? 12? 32? 41? 06? 18? 29

? 01

? 03

? 04

? 44

? 09

? 41? 42? 02? 08

? 07? 06? 14

? 30 ? 06

? 41? 33? 20 ? 18 ? 26? 21? 19? 16

? 42? 36? 29 ? 27? 24? 25

? 01

? 02

? 01? 14

? 08

? 10

? 05

? 04

? 40

? 14? 02

? 08

? 38 ? 04? 27

? 16

? 19

? 03? 34? 24

? 35 ? 10? 25? 18

? 09

? 29? 18? 41

? 35? 24? 19? 11

? 07

? 05

? 01

? 25? 44

? 16

? 06

? 37

? 28

? 32

? 15

? 22

? 39

? 31

? 12

? 15? 39

? 12

? 36

? 26

? 20? 42

? 09? 23

? 33

? 13

? 17

? 05

? 37

? 43

? 22? 13? 05

? 40

? 30

? 21

? 11

? 44? 32? 23? 17? 11

0? 34? 21 ? 27? 17? 13? 12? 15? 31? 31

? 15? 28? 37? 44

? 43? ?1

? 10

? ?3

? ? 2

? ? 1? 37

? 08

? 02

? 05

? 33

? ? 2

CAPSULES FOR IRRADIATION METALLIC SUMPLES IN REACTORS BOR-60, SM, RBT-6.

Accumulation of a damaging doze and ? ? in Cr-Ni steels. (Tirr=3000C ).

? =320-340; ? =350-450; ? = 450-700).

BOR-60 SMRBT-6

SM

SWELLING STEEL 0,1? -13Cr-2Mo-V-Nb-B

0

2

4

6

8

10

12

14

16

18

70 90 110 130 140

????, ???

????

????

??,

? ?,

1 %

Dose, dpa

Swel

ling

The autonomous loop channel BOR-60.

The autonomous loop channel is the most sophisticated in its design because it has been constructed as two-loop system . The channel is made in the form of capsule and entirely loaded in the reactor with provision made for termination of all communication line. This cannel was used for irradiation experiment followed with flow rate blockage up to total damage of fuel rods. The channel was constructed in such a way that all fragments of damaged fuel rods were present inside it. The gained experience made it possible to develop a similar channel with lead coolant for irradiation test of dummy fuel rods of advanced fast reactor BREST-OD-300 .

The irradiation devices were developed to examine creep properties, corrosion resistance and swelling of cladding and wrapper materials and perform irradiation tests of specimens at 320-1200 ?? .

Scheme of lead capsule loop BOR-60.

∅41*0,5

∅ 35*0,6∅37,5*0,6

∅9,4

flow meter (for lead)

channel vessel

Fuelelectric heater

heat exchanger (lead-sodium)pump wheel

magnetic muffpump electric drive

Thermocouples

The information support of irradiation tests

The information support of irradiation tests carried out in the research reactors aims at reception of signals from sensors of both process reactor systems and instrumented irradiation devices, processing of experimental data, mapping of trends and current values for measured values with their deviations from the specified limits, calculation, on-line monitoring of the irradiation test parameters and obtained data logging.

Structure of the dataStructure of the data--measurement system measurement system for experimental examinationsfor experimental examinations..

????????????? ??????????, ????????????????? ???

temperature

? ? pressure

? Flow rate

????????? ???????

level

?????????? ???????

displacement ? ?

ionizing chambers Direct charge detectors

strain location

Sensors and detectors

????????????????? Experimental ?????????

Processors for data acquisition, industrial mounting units with expansion plates, cables

single-channel modules of normalization and galvanic

separation 5B Analog Devices, for different types of signals:

Server for data

acquisition Ethernet

software

Operating stations, Ethernet software

Lower level Middle level Upper level

???? ?????? ????????? ? ?? ? ?????? ?????????????? ?????????? ?????? ? ??????? WAGO ?????????? ?????????? ????? / ?? ???? ????????? ????????? ? ????????????? ????????? ?????? ??? / ???

?????? 5? DAP

CONCLUSIONCONCLUSION

Ø RIAR has got several research nuclear reactors provided with experimental devices and in-pile examination techniques to solve pressing tasks of nuclear power engineering and reactor material science with regard to the reactors of different application. The Institute has got staff and subdivisions to perform all required activities such as:

Ø • Calculations, designing, fabrication, pre-irradiation tests of experimental devices, in-core sensors, reactor radiation detectors, experimental specimens including fuel rods;

Ø • Preparation of full-size and refabricated fuel rods of power reactor reference and experimental FA for irradiation test in research reactors;

Ø • In-pile testing followed with calculations of neutron and physical and thermal-hydraulic characteristics;

Ø • Mathematical simulation of processes occurred in experimental devises and products under irradiation;

Ø • Metrological and information support in-pile tests and examinations;Ø • Post-irradiation examination including intermediate and materials science ones of

products;Ø • Radioactive waste disposal and storage.