jlii:#. n e n e~tqb~5~ · reactor core by natural convection. during the long time of reactor...

Jlii:#. N E N E~tQb~5~ ~Jr ... ~~ PORTOROZ _ • SLOVENIA 25th International Conference Nuclear Energy for New Europe

Coolant Temperature Measurements in the Core of TRIGA Research Reactor

ABSTRACT

Romain Henry lozef Stefan Institute, Reactor engineering division R4

lamova 39 SI-IOOO Ljubljana, Slovenia

[email protected]

Marko Matkovic lozef Stefan Institute, Reactor engineering division R4

lamova 39 SI-IOOO Ljubljana, Slovenia

[email protected]

The TRIGA Mark II research reactor at the "lozef Stefan" Institute in Ljubljana is an open-pool type reactor cooled by demineralised light water. It has been used in various applications such as Neutron Activation Analysis, Neutron Radiography and Tomography and also for training personnel. Substantial studies have been done with regard to the neutron physics, however, only a few experiments related to reactor's thermal hydraulics have been performed so far, which makes the validation of neutron physics and reactor thermal hydraulics coupling attempt very difficult. In this light, two measurement campaigns were performed. Among them, one focuses on axial coolant temperature profile measurement within the reactor core. Axial coolant temperature profiles along the narrow water column confined with hot fuel elements were acquired during reactor operation. For this purpose, a specially tailored support structure was designed and built to accommodate 10 thermocouples in a vertical column within the core. Special attention was paid to: first, shield the temperature sensors from the hot surfaces of the fuel elements, second, keep the sensor's tips in contact with local coolant circulation, and third, generate reduced amount of activated material. The obtained experimental results were properly analysed and compared with CFD simulations. In fact, the only measurements of-a-kind were essential as they produced unique experimental data suitable for validation of the TRIGA's reactor core CFD model, which will later be used in an attempt to couple with the neutron physics code.

1 INTRODUCTION

The TRIGA Mark II research reactor at the lozef Stefan institute (lSI) is a typical 250 kW TRIGA reactor, which is used for various applications: such as Neutron Activation Analysis (NAA), Neutron Radiography and Tomography, education and training, radiation hardness studies and benchmark experiments for verification and validation of computer codes [1]. It is a light water reactor cooled by demineralised water that flows through the reactor core by natural convection. During the long time of reactor operation, an external cooling system operates, which cools the upper section of the pool through a forced convection process; however, natural convection remains the driving force for the flow through the core.

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Proceedings of the International Conference Nuclear Energy for New Europe, Portorož, Slovenia, September 5-8, 2016

Figure 1: Schematic view of the TRIGA Reactor

The core of the TRIGA reactor is placed at the bottom of an open tank with 5 m of

water column above it (Fig.1). The core has a cylindrical configuration with 91 designed

locations to accommodate fuel elements or other components such as control rods, a neutron

source and irradiation channels. Elements are arranged in six concentric rings: A, B, C, D, E

and F having 1, 6, 12, 18, 24 and 30 locations, respectively.

While the TRIGA Mark II research reactor at JSI was studied in great detail from

neutron physics point of view [2], very few experiments related to thermal hydraulics were

performed. As a result, the number of the available experimental results on velocity and

temperature fields in the pool is limited, which makes the validation of the TRIGA thermal-

hydraulics model impossible. Therefore, two experimental campaigns were performed to

measure the temperature of the coolant. The first one aimed to describe the natural convection

process in the pool of the reactor [3], while the second one was performed to better

understand the cooling process within the core. The experimental results are presented here

after.

First, checking the decay times of the chemical elements of which the available types of

thermocouples (TC) are made, has directed us toward the minimum output of radioactive

waste at the expense of acceptable decrease in accuracy of the measurements. With the

declared 0.36°C typical accuracy of the National Instrument module (NI PXIe-4353) [4] and

the terminal block (TB-4353) with strong cold-junction compensation, K-type thermocouples

(made of nickel, chrome and aluminium) were found a good compromise. Safe operation of

the reactor had to be taken into account; therefore, an experimental procedure was prepared in

agreement with the requirements imposed by the safety committee of the reactor.

A data acquisition application was developed in LabVIEW (National Instrument Corp.,

2003) running on a Windows 7 platform to acquire temperature readings from the

thermocouples and to perform the associated measurement statistics. PXI express chassis with

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integrated controller was used in combination with two NI PXIe-4353 modules and the

corresponding terminal blocks TB-4353 with powerful cold-junction compensation. The NI

measurement hardware was calibrated to ensure that the device meets its published

specifications. The thermocouples were calibrated together with the acquisition system in a

constant temperature reactor pool: measurements of all thermocouples were typically within

the 0.2 °C uncertainty band.

2 EXPERIMENTAL SET UP

To be able to measure axial coolant temperature profile within the reactor core, a

specially tailored support structure was designed and built to accommodate 10 thermocouples

in a vertical column within the core (Fig. 2). Special attention was paid to: first, shield the

temperature sensors from the hot surfaces of the fuel elements, and second, keep the sensor’s

tips in contact with local coolant circulation.

Figure 2: Tailored support structure before insertion into a measuring position.

Axial coolant temperature profiles along the narrow water column confined with hot

fuel elements were acquired during steady-state operation. The support was inserted at

different measuring positions (Fig. 3 left). Thermocouples were spread axially between

bottom and top grid as shown in figure 3 right.

Each measurement (at a different position) was done at full power (250 kW) with the

same average pool temperature (27ºC)

Tailored support structure

Axial locations of

thermocouples

Selected radial location

(measuring position)

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Figure 3: Measuring positions in the core of the TRIGA reactor (left) and schematic of the

axial position of the thermocouples (right).

3 CFD MODEL

The CFD simulations were performed with the commercial code ANSYS® CFX [5].

The geometry of the core inside the reflector was reproduced (Fig. 4 left).

Figure 4: Geometrical model of the core (left) and top view of the CFX mesh section of the

core: part of the fuel rod and coolant channel (approximately 2 cm x 1.5 cm) (right).

water cladding fuel rod

Active fuel

length

Control rods

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The core was modelled with 58 fuel rods, 4 control rods and 5 irradiation channels of

similar geometry (Fig. 4). All non-heated volumes were cut out from the computational

domain. Fuel elements were fully modelled (rod, fuel, cladding and graphite top and bottom).

The cross section of a single fuel channel in the CFD model is shown in figure 4 (right).

CFX solves conjugate heat transfer. Each fuel part (38.1 cm high cylinder of radius 1.87 cm)

is subdivided in 100 × 10 × 39 elements (respectively axial, radial and azimuthal). The grid in

the cooling channel is approximately 140 × 20 × 40 (respectively axial, radial and azimuthal).

At the inlet (bottom of the core) mass flow rate distribution and temperature was imposed

according to the steady state converged solution of the pool model considered previously [3].

Pressure was specified at the outlet (top of the core).

Power density distribution inside the fuel was obtained from Monte Carlo calculations

performed with the neutron transport code TRIPOLI [2].

4 RESULTS

Steady-state mode operation of the reactor was studied: the reactor is operated at

constant power 250 kW. The heat produced is removed by the external cooling system; mixed

convection is driving the flow within the pool. The pool is kept at an average temperature

around 27°C. Temperatures were recorded during 5 minute intervals, since numerous

oscillations were observed. Averaged values over the time intervals with the corresponding

average deviation are presented here after.

Fig. 5-8 present the temperature difference measured between the measuring point and

the bottom grid at measuring positions (MP) 5, 14, 16, 21 and the corresponding temperature

calculated by CFX. For measured values, four different curves are presented depending of the

orientation of the opening holes of the guiding tube.

Figure 5: Temperature difference measured and calculated inside the core at position MP 5.

Bottom grid is the reference point

ACTIVE

FUEL

LENGTH

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ACTIVE

FUEL

LENGTH

ACTIVE

FUEL

LENGTH

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It appears that in general, our calculations slightly underestimate the measured

temperature. Sensitivity analysis should be performed to determine which assumption of the

model could be responsible for this. One guess is that the inlet velocity is too high.

During operation, measurements have revealed that mixing is predominant in the

section corresponding to the active fuel length. Different orientations have been tested for

every position. Different temperature profiles inside the same channel were obtained. Some

positions like MP 14 (Fig. 6) are simple to explain: higher temperature gradients are observed

for orientation facing a fuel element than for the ones facing the reflector. For MP 16 and 21

(Fig. 7-8), temperature gradient are higher for the orientation facing the C ring than for those

facing the D rings suggesting that C rings is having hotter fuel due to higher fission sources.

The MP 5 (Fig. 5) is more difficult to explain, the presence of the transient rod (un-heating

rod) in the neighbourhood causes a strong heterogeneity in term of heat sources. As a result

the mixing in this area induced by large gradient of temperatures makes the flow chaotic.

Assuming that the flow direction is mainly vertical one can deduce the velocity of the

water inside the core by applying the energy balance for the core region. Detailed analysis of

the velocity field is still under investigation; however, first calculations have shown that

velocity of the flow is in the range of 10 to 20 cm s-1

.

5 CONCLUSION

An experimental campaign aiming to study the temperature profiles of the coolant

inside the core was performed and a CFD model of the reactor core was built. Preliminary

calculations have shown encouraging results. However, measurements revealed that a lot of

mixing occurs between the fuel rods since very different profiles can be observed at the same

position just by rotating the measuring device. In this respect, further studies would be

required to better asses the effect of coolant mixing in that region. An estimation of the bulk

velocity inside the core has also been done. Anyway, sensitivity analysis especially

concerning the inlet velocity should be performed.

ACTIVE

FUEL

LENGTH

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ACKNOWLEDGMENTS

The authors gratefully acknowledge the Research group P2-0026 and the valuable

contributions of the reactor operator staff at JSI, in particular from Anže Jazbec, Darko

Kavšek, Sebastjan Rupnik and Marko Rosman, for their help from the conception to the

realization of the experiment.

REFERENCES

[1] Ravnik M, Jeraj R (2003) Research reactor benchmarks. Nuclear Science and

Engineering 145 (1):145-152.

[2] Henry R, Tiselj I, Snoj L (2015) Analysis of JSI TRIGA MARK II reactor physical

parameters calculated with TRIPOLI and MCNP. Applied radiation and isotopes.

[3] Henry, R., Tiselj, I. & Matkovič, M. Heat Mass Transfer (2016). doi:10.1007/s00231-

016-1833-2.

[4] Corp. NI (2015). http://www.ni.com/pxi/.

[5] ANSYS, Inc. Release 17.1 User’s guide.

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