State Scientific Center – Research Institute of Atomic Reactors

Capabilities and Capacities of RIAR Research Reactors

N. ArkhangelskyState Corporation"ROSATOM“, Ul. Bolshaya Ordinka 24, 119017 Moscow,

Russian Federation

A. IzhutovRIAR, Dimitrovgrad-10, Ulyanovsk region, 433510 Dimitrovgrad

Russian Federation

IAEA, Vienna, 10-12 June 2013

Introduction

State Scientific Center–Research Institute of Atomic Reactors(RIAR) comprises:

• Research reactors complex:

SМ-3, BOR-60, МIR.М1, RBT-6, RBT-10/2;

• Europe’s largest complex for PIE of nuclear reactor coreelements, irradiated materials and nuclear fuel;

• Facilities and technologies for R&D nuclear fuel cycle area;

• Radiochemical complex for research and productiontransuranium elements, radionuclide's of high specificactivity and sources;

• Radioactive waste treatment complex.2

1. The SM-3 reactor

3

SM-3 - high-flux reactor

operates since 1961

1.1 General information and technical data of the SM-3

Specification Value

Reactor typeVessel-type water-cooled with

a trap

Power, МW 100

Max neutron flux, s-1⋅сm-2 5·1015

Power operation days per year,

days230-240

Fuel UO2 90% enriched in U-235

Core arrangement Square with a central trap

Core dimensions, mm 420 ×420 ×350

Planned life-time More than 2020

4

1.1 General information and technical data of the SM-3

Cells for irradiation

up to 81

trap Up to 27 cells fortargets Ø 12mm

core up to 6 FAs with 4 cellsfor targets Ø12mm, up

to 4 FAs with 1 cell fortargets Ø24.5mm

reflector 30 cells for channels and devices Ø68mm,

including 20 cells for instrumented devices.

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Irradiation capabilities of the SM-3

1.1 General information and technical data of the SM-3

Number of cells for irradiation, including: Up to 81

• trapblock option: 27 cells Ø(12 – 25) mm;

channel option: channel Ø50 mm + 18 cells

• coreUp to 6 and up to 4 FAs with 1 cell for

targets Ø24.5mm

• reflector30 (of which 20 cells can be instrumented

or supply by separately coolant)

Irradiation positions:

Neutron flux, s-1·cm-2

≤ 0,67 eV 0,67÷100 eV ≥ 0,1 MeV

• trap 1,3⋅1015 1,7⋅ 1014 1,4⋅1015

• core 1,3⋅1014 9,3⋅1013 2,0⋅1015

• reflector ≤ 1,35⋅1015 – ≤ 3,3⋅1014

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1.2 General data of testing facilities of the SM-3

Characteristics VP-1 VP-3

Maximal operating pressure,

MPa5.0 18.5

Coolant temperature, °С 90 300

Flow rate, m3/h 30 5÷8

Thermal power 500 90

Coolant water water

7

Loop facilities for testing fuel, control rods, structural materials

of the water cooled reactors

1.2 General data of testing facilities of the SM-3

8

Design of Irradiation rig Medium

Testing parameters

φ, (Е>0.1 MeV)

cm-2s-1

φ, cm-2s-1 K, dpa/h Kt, dpa/year

Loop rig in the reflector

Water (300°C,

18.5 MPa)1013–4·1014 2·1013 –4·1014 3·10-5 –1,2·10-3 0,15–6,0

Loop rig in the core

Water (300°C,

18.5 MPa)1,5·1015 2·1014 ≤3·10-3 15–18

Ampoule rig in the

reflector

Boiling water

(up to 320°C),

supercritical

water, gas (He,

Ne, N2) - 400-

1150°C)

5·1012 – 4·1014 2·1013 –4·1014 1·10-5 –1,2·10-3 0,1–6,0

Ampoule rig in the core

Boiling water

(up to 320°C),

supercritical

water, gas(400-

1150°C)

(1,5–2)·1015 (2–3)·1015 ≤4·10-3 16–22

1.2 General data of testing facilities of the SM-3

In-pile testing facilities:

� Long-term strength and creep tests of steels and alloys under

longitudinal tension (facility “Neutron-8”) and internal gas

pressure test at 550÷800оС;

� In-pile investigation of relaxation resistance of structural

materials ;

� In-pile investigation of creep of nuclear fuel at temperatures

700÷1100°C, including pre-irradiated fuel samples to investigate

the burn up effect on the creep characteristics;

� In-pile tests of the core material for NPP and advanced

nuclear facilities at high damage rate of up to 25 dpa/year in the

wide range of temperatures and different environment.

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1.2 General data of testing facilities of the SM-3

Irradiation devices for testing different types samples up to 7-17dpa in

the near-core reflector cells and in the core cells:

� Tests in a high-temperature loop facility VP-3 in the near-core

reflector cells at 250-300°C and 16MPa up to 10dpa. The experimental

volume taken by the samples is Ø40mm and 350mm in height. The

neutron flux (Е≥0.1MeV) makes up (3-4)*1014n/cm2*s. The temperature

non-uniformity in the experimental volume is 10-20°C.

� Tests in the boiling water channels in the near-core reflector cell at

water temperature 160-300°C and 10MPa up to 10dpa. Samples should

be located in stainless steel containers to prevent corrosion. The

experimental volume taken by the samples is Ø60mm and 350mm in

height. The neutron flux (Е≥0.1MeV) makes up (3-4)*1014n/cm2*s. The

temperature non-uniformity in the experimental volume is 5-10°C.

� Tests in the core cells up to 15-17dpa at ≥ 600°С. The experimental

volume taken by the samples is Ø20mm and 350mm in height. The

neutron flux (Е≥0.1MeV) makes up (1.5-2.0)*1014n/cm2*s. The

temperature non-uniformity in the experimental volume is 100-150°C.

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1.2 General data of testing facilities of the SM-3

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Radionuclides accumulation devices:A wide range of neutron flux density change and its spectral characteristicsmake it possible to search and implement optimal schemes of radionuclidesaccumulation:

• the multi-stage accumulation of Cm and Cf-252 heavy isotopes isperformed in the neutron trap cells and two reflector cells closest to thecore;

• high specific activity Ni-63, Sn-113, Sn-119m, Fe-55, Fe-59, Cr-51, Se-75,Ir-192 etc.;

• the hard neutron spectrum demanding P-32, P-33; Gd-153 and Sn-117m;• large-scale accumulation of such radionuclides as Co-60 and Ir-192. The

annual output achieves 300 kCi ofCo-60 and 400 kCi of Ir-192.

2. The BOR-60 reactor

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BOR-60 – sodium fast experimentalreactor was put into operation in

1969.

2.1 General information and technical data of the BOR-60

Characteristics Value

Thermal capacity, MW до 60

Max neutron flux density, сm-2·s-1 3.7·1015

Sodium velocity in the core, m/s up to 8

Coolant temperature, ºС:

inlet / outlet up to 360 / 515

FuelUO2

or UO2-PuO2

Planned life-time 2020

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Loading cartogram of the BOR-60 reactor

1 - FA, 2 - reflector assemblies, 3 - vertical channels, 4 - control rod, 5 - instrumented cell (D23).

Reactor Load Capacity

Cells quantity

- for FA

- for absorbing rods

- instrumented cells

265

156

7

1

Max quantity of non-

fuel cells in the core

~(15-

22)

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2.1 General information and technical data of the BOR-60

0

5

10

15

20

0.0 4.5 9.0 13.5 18.0 22.5 27.0 31.5 36.0 40.5

R, cm

DPA

, dpa/

yea

r

15

2.1 General information and technical data of the BOR-60

Radial distribution of the damage dose

The experimental volume:hexagonal cross-section of irradiation device,

flat to flat size is 44 mm, the core height 450 mm

2.2 General data of testing facilities of the BOR-60

Schematic view of irradiation device

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2.2 General data of testing facilities of the BOR-60

Types of irradiation rigs (IR) Temperature,°°°°С

Heater

IR with sodium medium :•without heater•with metallic heater•with fuel heater•based on driver fuel design•with convection

320-360450-500500-700

up to 700450-600

self-heating,

W,UO2

IR:•with liquid metal medium (Pb, Pb-Bi)

•Gas medium (He, Ne)

350-650350-650

W,UO2,self-

heating

650-900350-1200

W,self-

heating

1717

--------Air+Ar---------

----------Na------------

----------He------------

-Capsule of testing

sampels--------

-----------Na-----------

The view of instrumented irradiation device for the cell D23

2.2 General data of testing facilities of the BOR-60

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The irradiation

conditions

The test

temperature and it

non-uniformity, oC

The doze rate

dpa/year

1 (350-450) ± 10 7-15

2 (450-650) ± 25 7-15

3 (650-800) ± 35 7-15

4(800-1100) ± 100

7-15

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2.2 General data of testing facilities of the BOR-60

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3. The MIR.M1 reactor

The MIR.M1 reactor designed for tests of fuel of different types in the loop test facilities,

in operation since 1967.

3.1 General information and technical data of the MIR.M1

Characteristics Value

Nominal thermal power, MW 100

Max thermal neutron flux density

in the loop channel, s-1⋅cm-2 5·1014

Power operation days per year ~ 240

Fuel UO2 90% enriched in U-235

Core height, mm 1000

No. of loop facilities 8

No. of loop channels. 11

Life-time More than 2020

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3.1 General information and technical data of the MIR.M1

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The MIR reactor is mainly

designed for testing of

different nuclear power

reactor fuel under normal

(steady-state) operating

conditions as well as

transient and emergency

ones in a certain project.

– operating FA channel

– experimental channel

– combined operating

FA with absorber

– control rod channel

3.1 General information and technical data of the MIR.M1

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Currently 8 loop facilities are

available in the MIR reactor. Each

of these facilities is connected

with 1-2 loop channels (the

maximum diameter - up to 148

mm). The channels are used for

setting up experimental devices

with experimental fuel.

Loop facilities equipment andsystems:Circulation circuit (pumps, heatexchangers, pressurizers, etc);Fuel cladding integrity control;Water condition, feeding andsampling;Emegency cooling;Measuring and registration .

3.2 General data of testing facilities of the MIR.M1

Loop facilities PV-1 PV-2 PVK-1 PVK-2 PVP-1 PVP-2 PM-4 PG-1

Max capacity, kW 2000 3000 2000 3000 500 2000 500 160

Coolant WaterWater, boiling

water

Water-superheated

steam

Lead-bismuth

N2, He

Number of channels

2 2 2 2 1 1 1 1

Flow rate, t/h 16.0 14.0 16.0 14.0 0.7 10.0 8.0 0.7

Pressure, МPа ≤ 17 ≤ 18 ≤ 7 ≤ 20 ≤ 1,5 ≤ 20

Temperature: inlet/outlet, ° С

≤ 300/350 325/350 300/500 550 550

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3.2 General data of testing facilities of the MIR.M1

25

Current test programmes:1. The tests for improving and upgrading the Russian PWR (WWER)fuel:

� long term tests of short-size rods with different modifications ofcladding and fuel pellets;

� reirradiation of NPP refabricated and full-size fuel rods up toachieving 80 MW·d/kg U;

� continuation of the RAMP type experiments at high burn-up offuel;

� experiments with leaking fuel rods at different burn-up andunder transient conditions;

� in-pile tests with simulation of LOCA and RIA type accidents.

2. Testing of the LEU research reactor fuel within the framework ofthe RERTR programme:

� irradiation of mini fuel elements;� Irradiation of full-size fuel assemblies.

3.2 General data of testing facilities of the MIR.M1

26

Types of irradiation devices for testing of the PWR fuel:

a) for testing of fuel under normal (steady-state) operating conditions:

� dismountable devices for testing short-size (≤ 250 mm) fuel rods,

up to 4 such rigs can be installed one over another in one loop

channel;

� dismountable and instrumented device for testing fuel rods

~1000 mm, containing up to 19 fuel rods;

� device for combined irradiation of refabricated (≤ 1000 mm) and

full-size fuel rods (≤ 3500 mm) of spent NPP fuel.

b) for testing under transient and emergency design based conditions:

� dismountable devices for power cycling and RAMP experiments of

instrumented fuel rods by displacement or rotation of the

absorbing screens in the experimental channel;

� instrumented device for testing under LOCA and RIA conditions.

3.2 General data of testing facilities of the MIR.M1

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Types of irradiation devices for sort size (dummy) fresh and

refabricated fuel rods