The TRIGA Mark-II reactor was installed by General Atomic (San Diego, California, U.S.A.) in the years 1959 through 1962, and went critical for the first time on march 7, 1962. Operation of the reactor since that time has averaged 220 days per year, without any long outages. The TRIGA-reactor is purely a research reactor of the swimming-pool type that is used for training, research and isotope production (Training, Research, Isotope Production, General Atomic = TRIGA). Throughout the world there are more than 50 TRIGA-reactors in operation, Europe alone accounting for 10 of them.

The TRIGA-reactor Vienna has a maximum continuous power output of 250 kW (thermal). The heat produced is released into a channel of the river Danube via a primary coolant circuit (deionized, distilled water at emperatures between 20 and 40 *C) and a secondary coolant circuit (ground water at temperatures between 12 and 18 *C), the two circuits being separated by a heat exchanger.

The reactor core consists of some 80 fuel elements (3.75 cm in diameter and 72.24 cm in length), which are arranged in an annular lattice (Figs. 1 to 3). Two fuel elements have thermocouples implemented (Fig. 4) in the fuel meat which allow to measure the fuel temperature during reactor operation. At nominal power (250 kW), the center fuel temperature is about 200 *C. Because of the low reactor power level, the burn-up of the fuel is very small and most of the fuel elements loaded into the core in 1962 are still there. Should these fuel elements ever become unserviceable, they will be sent back to the United States.

Inside the fuel element cladding (aluminum or steel), the fuel is in the form of a uniform mixture of 8 wt% uranium, 1 wt% hydrogen and 91 wt% zirconium, the zirconium-hydride, being the main moderator. Since the moderator has the special property of moderating less efficiently at high temperatures, the TRIGA-reactor Vienna can also be operated in a pulsed mode (with a rapid power rise to 250 MW for roughly 40 milliseconds). The power rise is accompanied by an increase in the maximum neutron flux density from 1x1013 cm-2s-1 (at 250 kW) to 1x1016 cm-1 (at 250 MW). This negative temperature coefficient of reactivity, as it is called, brings the power level back to approximately 250 kW after the excursion, the maximal pulse rate is 12 per hour, since the temperature of the fuel elements rises to about 360 *C during the pulse and, therefore, the fuel is subjected to strong thermal stress.

The reactor is controlled by three control rods which contain boron carbide as absorber material. When these rods are fully inserted into the reactor core, the neutrons continuously emitted from a start-up source (Sb-Be photoneutron source) are absorbed by the rods and the reactor remains sub-critical. If the absorber rods are withdrawn from the core (two of them by an electric motor and one pneumatically, Figs. 5 and 6), the number of fissions in the core and the power level increases. The start-up process takes roughly one minute for the reactor to reach a power level of 250 kW from the sub-critical state. The reactor can be shut-down either manually or automatically by the safety system. It takes about 1/10 of a second for the control rods to fall into the core.

The reactor is controlled by four nuclear channels (Fig.7), their signals are displayed both at a colour graphic- monitor and at bar graph indicators (Figs.8a and b).

a) The auto-ranging wide-range channel NM-1000 controls the reactor power from the source level (around 5 
    mW) up to nominal power of 250 kW. It uses a Campbell fission chamber, the signal is controlled by a 

b) Two independent linear channels, NMP-Ch and NMP-Ph control the reactor power from the source level up to 
    nominal power. The signals pass over a range switch which selects the power range. If the signal of one of 
    these two channels exceeds the selected power range for more than 5%, the reactor is shut down automatically. 
    Both channels use compensated ionisation chambers as sensors.

c) For the control of reactor pulse operation an uncompensated ionisation chamber is used. This chamber 
    measures the shape of the reactor pulse which is displayed on the graphic monitor. Further pulse data like 
    integrated power are calculated from this signal.

In accordance with its purpose as a research reactor, the TRIGA Mark-II is equipped with a number of irradiation devices (Figs. 9 and 10):

  •  5 reflector irradiation tubes 1 central irradiation tube
  •  1 pneumatic transfer system (transfer time 3 s)
  •  1 fast pneumatic transfer system (transfer time 20 ms)
  •  4 neutron beam holes
  •  1 thermal column
  •  1 neutron radiography facility

In the reflector irradiation tubes 10 containers can be irradiated simultaneously. In the central irradiation tube samples up to 38.4 mm in diameter can be exposed to neutrons at a neutron flux density of 1013 cm-2s-1, while the pneumatic transfer system allows to transfer the materials to be activated into the reactor from a chemistry laboratory and back again after the required period of irradiation, without the experimentalist having to leave his working place. The four neutron beam tubes permit extraction of neutron beams of all energies into the reactor hall for the purpose of neutron and solid-state physics experiments. The thermal column is used to extract with a thermal spectrum into the reactor hall, unlike the beam holes, the space between the reactor core and the hall is in this case filled with graphite to slow down the neutrons.

The neutron radiography facility is used to investigate components by neutron irradiation similar to X-ray radiography. However, neutrons show especially hydrogen or neutron absorber material in solid matter.


fuel-moderator material 8 wt% uranium
91 wt% zirconium
1 wt% hydrogen
uranium enrichment 20% uranium-235
fuel element dimensions 3.75 cm in diameter
72.24 cm in length
cladding  0.76 mm aluminum or 0.51 mm steel
active core volume  max. 49.5 cm diameter, 35.56 cm high
core loading  3 kg of uranium-235

material graphite with aluminum cladding
radial thickness 30.5 cm
top and bottom thickness 10.2 cm

reactor shielding construction  heavy and standard concrete
6.55 m high, 6.19 m wide, 8.76 m long
reactor tank 1.98 m in diameter
6.40 m in depth

radial:  30.5 cm of graphite;
45.7 cm of water and at least
206 cm of heavy concrete
vertical: above the core 4.90 m of water and
10.2 cm graphite;
underneath the core 61.0 cm water,
10.2 cm graphite and at least
91 cm standard concrete


(1) four beam holes 15.2 cm in diameter
(2) one central irradiation tube (center of core)
(3) five reflector irradiation tubes
(4) one pneumatic transfer system (near core edge)
(5) a thermal column with cross section 1.22x1.22 m and length 1.68 m
(6) experimental tank with surface area 2.44x2.74 m and depth 3.66 m; connected to the 
     reactor by means of a neutron radiography collimator 0.61x0.61 m in cross section and 
     1.22 m long. 

  • Two boron carbide control rods with electric motor and rack and pinion  drive.
  • One boron carbide pulse rod with compressed air drive (5 bars).
  • Maximum reactivity insertion rate - time rate of change (excluding pulse operation ): 0.04% dk/k per second 
  • Total rod worth about 4.8% dk/k.

Thermal power output: 250 kW
Fuel element cooling: natural convection of the tank water
below 100 kW, pump circulation cooling above 100 kW
tank water cooling: heat exchanger
thermal flux:  1x1013 cm-2s-1 in the central irradiation tube
1.7x1012 cm-2s-1 in the irradiation tubes
prompt temperature coefficient: -1.2x10-4 dk/k°C
mean prompt neutron lifetime: 6.0x10-5 s.

peak power 250 MW
prompt pulse energy yield 10 MW s
prompt pulse lifetime 40 ms
total energy yield  16 MW s
minimal period 10 ms
maximum reactivity insertion 1.4% dk/k = 2$
maximum repetition frequency  12/h
number of fissions during a pulse  3x1017
maximum fuel temperature: during the pulse    240 °C
9 seconds after the pulse  360 °C

References and suggestions to