Experimental complexes

The research work is based on scientific-technical and production infrastructure and highly qualified scientific and technical personnel. Part of scientific research is supported by projects of the IAEA, ISTC, European Economical Community (EEC) and other contract projects providing for considerable re-equipment of NNC Institutes with devices and facilities. Experimental Complexes:

  • PROTOCOL on Trial Start-Up to Test KTM Tokamak

  • WWR-K Reactor

  • Critical Test Bench

  • IVG.1M Reactor

  • IGR Reactor

  • Liana Test Bench

  • Experimental complexes

PROTOCOL on Trial Start-Up to Test KTM Tokamak

On September 05, 2010 trial start-up of KTM Tokamak was carried out. The operations on preparation and performance of KTM Tokamak trial start-up are fulfilled within implementation of budget program 045 Creation of Kazakhstan Thermonuclear Material Studying Reactor Tokamak KTM.

Trial Start-Up Objective is working gas breakdown in vacuum chamber of KTM Tokamak and formation of plasma thread with current of 10-30 кА.

   To obtain plasma thread the setting of the following Tokamak systems are performed:

The diagnostics of first order plasma physical parameters;

  • Subsystem of data collection and registration;
  • Preionization system of working gas;
  • Power supply system of KTM magnetic windings;
  • Vacuum pumping system of KTM chamber.


Монтаж токамака КТМ
Монтаж токамака КТМ 2-е фото
Проверка герметичности вакуумной камеры КТМ


In the course of operations the following is carried out:

The conditions for stable operation of preionization system of working gas (argon) are found – pressure of working gas and induction of toroidal magnetic field;

  • Zone in vacuum chamber with low level of scattered magnetic field is found;
  • Realization of working gas breakdown and plasma current rise up to target level;
  • Registration of basic physical parameters is performed.

The following systems are used for breakdown and formation of plasma thread:

  • Preionization system for ionization of working gas;
  • Power supply of electromagnetic winding of toroidal field;
  • Condensing power supply of central solenoid and electromagnetic windings of poloidal field PF1, PF4;
  • Magnetic diagnostics, CCD video camera and photoelectric multiplier (PEM) for diagnostics of physical parameters.

Working gas breakdown and plasma current are registered with the help of Rogowski coil, voltage sensing device and photomultiplier.

KTM subsystems, participating in the experiment on plasma obtaining are shown in Figure.

For pre-ionization formation by electron cyclotron resonance magnetron source is used with frequency of 2.45 GHz, microwave emissive power is - 900 W, emission is supplied through waveguide from low magnetic field, microwave emission impulse duration - 300 ms. Pre-ionization system is presented in Figure 2.

Light is seen through vacuum chamber window due to gas ionization.

To create eddy electric field, current is supplied in poloidal field winding by way of capacitors bank discharging of 22 mF capacity and voltage of 5 kV.

The main parameters upon which breakdown and plasma thread are formed: working gas pressure (argon) - 10-4 torr, voltage near the breakdown is – 8.4 V, magnetic field on axis of toroidal winding is – 0.043 tesla, plasma discharge duration is ~ 40 ms.

Figure 3 demonstrates video shooting moment of vacuum chamber from vertical branch pipe during plasma thread gaining (approx. 20-th millisecond since current injection into central solenoid coil).


1. Scenario is developed for trial start-up of Tokamak KTM. Physical parameters system adjustment is performed for first order plasma, as well as adjustment of data registration system, system of working gas pre-ionization, KTM electromagnetic winding feed system, vacuum chamber preparation system KTM.

2. Trial start-up of KTM is carried out and maximum plasma flux of 25 kA is obtained in accordance with Tokamak KTM trial start-up tasks.

WWR-K Reactor

A thermal neutron pool-type reactor. Coolant, moderator and reflector - desalted water.

It was commissioned in 1967, operated at thermal power of 10MW till 1988 without deviations from standard modes.

In period 1988 to 1998 the work was performed to enhance safety under high seismicity conditions (calculations and validations, structure strengthening, doubling of systems responsible for safety, new documentation execution). Since the core configuration was changed the thermal power was reduced to 6MW without neutron flux loss.

Technical Parameters

Thermal power, MW




Loading 235U, kg


Enrichment 235U, %


Core height, mm


Experimental channel diameter, mm


Thermal neutron flux density:


in the central channel

1,4×1014 n/cm2·s

in two channels of the core

1,1×1014 n/cm2·s

in peripheral channels

1012 - 1013 n/cm2·s

Reactor life-time

14 days

The reactor is equipped with water-tube dispatch system, air-tube dispatch system, a universal loop facility, neutron radiography facility, the facility for analysis of uranium containing samples by delayed neutron method, in-pile facilities to test structure materials for long-term strength and creep, a number of hot cells for work involving high-level materials.

At this reactor, besides the fundamental nuclear-physical and material and structure investigations and in-pile tests, work on medical radioisotopes and gamma-source production, neutron alloying of silicon, neutron-activation analysis is conducted. The opportunity of core upgrading to make application of low-enriched uranium possible is studied. National and international workshops on physical protection of nuclear facilities, nuclear material accounting and control are conducted.

Critical Test Bench

The critical test bench is a low power thermal neutron physical reactor that includes light water moderator and reflector (water or beryllium). It is meant for studies of neutron-physical characteristics of WWR type water-water reactor cores and other reactor core elements, for experiments to validate reactor facility safety, as well as for simulation of conditions of loop channel and different in-pile device tests. The diameters of central experimental channels are 65, 96, 140 and 380 mm, diameters of peripheral channels are 65, 96 and 140 mm, thermal neutron flux density is 5×108 n/cm2·s.

IVG.1M Reactor

IVG.1M research reactor is an upgraded version of IVG.1 reactor used for tests of fuel assemblies (FA) and cores of high temperature gas-cooled reactors including nuclear jet propulsion (NJP) reactors and nuclear power propulsion plants (NPPP).

IVG.1M reactor can be used in tests providing solution of the following tasks:

  • testing of different type FA under operating conditions;
  • FA structure material reactor tests;
  • testing of FA structures and their elements;
  • study of possible accident conditions and testing of preventive measures.

Technical characteristics

Thermal power

72 MW

Core effective diameter

548 mm

Core height

800 mm

Uranium-235 content in the core

4,6 kg

Thermal neutron flux density

3,5×1014 n/cm2·s

Water rate through the reactor

up to 380 kg/s

Maximum water temperature


Горячая камера
Объект 300 - Реактор ИВГ.1М
Объект 300 - Пультовая ИВГ.1М
Защитный периметр
Стенд Ангара
Снятие крышки реактора ИВГ.1М


The most important works conducted recently at the reactor are:

  • studies on ITER (International Thermonuclear Experimental Reactor) structure material interaction with hydrogen and its isotopes under reactor irradiation condition;
  • study of reactor radiation scattering in the atmosphere to validate safety of nuclear power.

IGR Reactor

One of the oldest research reactors in the world, IGR reactor is a unique source of neutron and gamma radiation characterized by a high dynamics of power change.

Intense development of reactor engineering in 50-s of the 20th century predetermined creation of pulse reactor for experimental research of non-stationary physical processes occurring in the core when high reactivity is imparted.

IGR reactor history is started with CPSU Central Committee's and USSR Soviet of Ministers' Decree #518-246 of May 13, 1958. Under this Decree it was envisaged "...to build an experimental facility with high temperature homogeneous graphite reactor at site #905 (Semipalatinsk Test Site) of the Ministry of Defense".

IGR reactor was built in the shortest time possible and allowed to start experimental study of pulse reactor dynamics in 1961. In 1962 the studies of fuel and structure materials behavior in advanced reactor facilities including NJP (nuclear jet propulsion) began.

Technical Parameters

Neutron flux maximum density

7×1016 n/cm2·s

Thermal neutron maximum fluence

3,7×1016 n/cm2

Minimum half-width of impulse

0,12 s

Центральный зал ИГР
Реактор ИГР
Объект 100 - панорама


IGR research reactor is a pulse reactor on thermal neutrons with homogeneous uranium-graphite core of thermal capacity type. Graphite high thermal capacity enabled to do without the system of forced removal of heat released in the core during the reactor operation. Absence of standard coolant circuit considerably reduces the risk of radiation accident at the reactor.

IGR reactor nuclear safety is specified by a considerable level of negative reactivity coefficient that provides guaranteed shutdown of any physically possible power impulse, initiated by positive reactivity insertion through CPS (control and protection system) rods removal.

Among pulse reactors, IGR reactor has the highest fluency of thermal neutrons and integral gamma radiation dose in the experimental cavity of 228 mm in diameter and 3825 mm high.

Liana Test Bench

Liana test bench is designed for studies of high temperature processes of hydrogen isotope mass-transfer in structure materials during irradiation at IVG.1M reactor.

Studies are conducted with the help of differential method of hydrogen permeability in the mode of continuous pump-down. Recording of partial pressure of hydrogen isotopes penetrating through the sample studied is performed with computerized program-technical complex basing on mass-spectrometer IPDO-1 built in systems of reactor complex registration.

Hydrogen permeability studies were performed using samples of stainless steel Cr18Ni10Ti, nickel, vanadium alloy V-4Cr-4Ti, copper alloy CuCrZr in IVG.1M reactor at fast neutron flux of 1013 n/cm2s, thermal neutron flux of 1014 n/cm2s, at temperature range 523-1023K and hydrogen input pressure of 10-2 - 106 Pa.

Technical Parameters

Temperature interval of investigation

500...1300 K

Hydrogen isotope pressure range at the inlet side of the sample


Hydrogen isotope pressure range at the outlet side of the sample


Experimental complexes

The research work is based on scientific-technical and production infrastructure and highly qualified scientific and technical personnel. Part of scientific research is supported by projects of the IAEA, ISTC, European Economical Community (EEC) and other contract projects providing for considerable re-equipment of NNC Institutes with devices and facilities. Experimental Complexes:



Dedicated to the 25th Anniversary of Semipalatinsk Test Site Closure
VII International Conference SEMIPALATINSK

First announcement 
Second announcement 

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