**4.2.4 Thermal hydraulic parameters of coolant channels**

The pertinent parameters required for the analysis of coolant channels are tabulated in Table 3. Figure 17 shows the power dissipate and the temperature increase in each channel at 265 kW reactor total power. This power was the results of the thermal power calibration (Mesquita et al. 2007). The profile of the mass flow rate and velocity in the core is shown in the graphs of Figure 18. Figure 19 compares experimental and theoretical profile of mass flux *G* in the core coolant channels. The theoretical values were calculated using PANTERA code (Veloso, 2005). As it can see by the Reynolds number the flow regime is turbulent in channels near the core centre.

Experimental Investigation of Thermal Hydraulics in the IPR-R1 TRIGA Nuclear Reactor 19

Figure 20 shows the water temperatures evolution at the reactor pool, and the inlet and outlet coolant temperature in the core's hottest channel until the beginning of steady state. The results showed that the thermocouples positioned 143 mm over the top grid plate (Inf 7) measure a temperature level higher than all the other thermocouples positioned over the reactor core. The temperature measurements above the core showed that thorough mixing of water occurs within the first centimeters above core top resulting in a uniform water temperature. It means that the chimney effect is not much high, less than 400 mm above the reactor core, in agreement with similar experiments reported by Rao et al. (1988). The chimney effect is considered as an unheated extension of the core. The chimney height is the

Fig. 18. Mass flow rate and velocity in coolant channels at 265 kW

Fig. 19. Mass flux in coolant channels at 265 kW

**4.3 Pool temperature** 


1Specific heat (*cp* ) = 4.1809 [kJ/kgK], water density (*ρ*) 995 kg/m3 and dynamic viscosity(*µ*) = 0.620 10-3 kg/ms at 45 oC.

Table 3. Properties of the coolant channel at the power of 265 kW1

As can be seen in Figure 18 and Figure 19 the velocity and mass flux in each channel are proportional to power dissipated in the channel.

Fig. 17. Power and temperature increase in coolant channels at 265 kW

Fig. 18. Mass flow rate and velocity in coolant channels at 265 kW

Fig. 19. Mass flux in coolant channels at 265 kW

#### **4.3 Pool temperature**

18 Nuclear Reactors

**Channel Channel Tout - Tin Flow Area Mass Velocity Reynolds Power Rate Flux Number** 

 **[kW] [oC] [kg/s] [cm2] [kg/m2s] [m/s] -**  0 2.65 15.5 0.041 1.574 260.48 0.26 3228 1 9.81 15.5 0.151 8.214 183.83 0.18 5285 2 5.70 17.1 0.080 5.779 138.44 0.14 5181 3 4.85 16.3 0.071 5.735 123.79 0.12 4184 4 3.00 12.1 0.059 5.694 103.62 0.10 2525 5 0.93 7.7 0.029 3.969 73.06 0.07 549

1Specific heat (*cp* ) = 4.1809 [kJ/kgK], water density (*ρ*) 995 kg/m3 and dynamic viscosity(*µ*) = 0.620 10-3

As can be seen in Figure 18 and Figure 19 the velocity and mass flux in each channel are

Table 3. Properties of the coolant channel at the power of 265 kW1

Fig. 17. Power and temperature increase in coolant channels at 265 kW

proportional to power dissipated in the channel.

kg/ms at 45 oC.

*q* **∆** *T m G u* **Re** 

Figure 20 shows the water temperatures evolution at the reactor pool, and the inlet and outlet coolant temperature in the core's hottest channel until the beginning of steady state. The results showed that the thermocouples positioned 143 mm over the top grid plate (Inf 7) measure a temperature level higher than all the other thermocouples positioned over the reactor core. The temperature measurements above the core showed that thorough mixing of water occurs within the first centimeters above core top resulting in a uniform water temperature. It means that the chimney effect is not much high, less than 400 mm above the reactor core, in agreement with similar experiments reported by Rao et al. (1988). The chimney effect is considered as an unheated extension of the core. The chimney height is the

Experimental Investigation of Thermal Hydraulics in the IPR-R1 TRIGA Nuclear Reactor 21

The heat generated by fission in the fuel material is conducted through the fuel, through the fuel-cladding interface, and across the cladding to the coolant. The thermal and hydrodynamic purpose of the design is to safely remove the heat generated in the fuel without producing excessive fuel temperatures or steam void formations and without closely approaching the hidrodynamic Critical Heat Flux (CHF) (Huda et al. 2001). As the IPR-R1 TRIGA reactor core power is increased, the heat transfer regime from the fuel cladding to the coolant changes from the single phase natural convection regime to subcooled nucleate boiling. The hottest temperature measured in the core channel was 65 oC (Channel 1'), below 111.4 oC, the water saturation temperature for the pressure of 1.5 bar. Therefore, the saturated nucleate boiling regime is not reached. Channel 1' is the closest channel to the centre of the reactor where it is possible to measure the water entrance and exit temperatures. The hottest channel is Channel 0, closer to the centre. With the measured temperature values in the Channel 1', the value of critical flow was evaluated in these two channels. The Bernath correlation was used (Eq. 5) for the calculation of the critical heat flux. With the reactor power of 265 kW operating in steady state, the core inlet temperature was 47oC. The critical flow for the Channel 0 is about 1.6 MW/m2, giving a Departure from Nucleate Boiling Ratio (DNBR2 of 8.5. Figure 22 and Figure 23 show the values of critical flow and DNBR for the two channels. The theoretical values for reactor TRIGA of the University of New York and calculated with the PANTERA code for the IPR-R1 are also shown (General Atomic, 1970) and (Veloso, 2005). The two theoretical calculations gave smaller results than the experiments. These differences are due to the core inlet temperature

**4.5 Critical Heat Flux and DNBR** 

used in the models.

2CHF/actual local heat flux

Fig. 22. Critical heat flux as a function of the coolant inlet temperature

The minimum DNBR for IPR-R1 TRIGA (DNBR=8.5) is much larger than other TRIGA reactors. The 2 MW McClellen TRIGA calculated by Jensen and Newell (1998) had a DNBR=2.5 and the 3 MW Bangladesh TRIGA has a DNBR=2.8 (Huda and Rahman, 2004).

distance between the channel exit and the fluid isotherm plan above the core and it depends of the reactor power.

Fig. 20. Temperatures patterns in the reactor pool at 265 kW thermal power

#### **4.4 Temperatures with the forced cooling system tuned off**

Figure 21 shows the behavior of fuel element, channel outlet, reactor pool, and specimen rack temperatures at various operation powers, with the forced cooling system turned off.

Fig. 21. Temperature evolution as a function of power with the forced cooling system off

#### **4.5 Critical Heat Flux and DNBR**

20 Nuclear Reactors

distance between the channel exit and the fluid isotherm plan above the core and it depends

Fig. 20. Temperatures patterns in the reactor pool at 265 kW thermal power

Figure 21 shows the behavior of fuel element, channel outlet, reactor pool, and specimen rack temperatures at various operation powers, with the forced cooling system turned off.

Fig. 21. Temperature evolution as a function of power with the forced cooling system off

**4.4 Temperatures with the forced cooling system tuned off** 

of the reactor power.

The heat generated by fission in the fuel material is conducted through the fuel, through the fuel-cladding interface, and across the cladding to the coolant. The thermal and hydrodynamic purpose of the design is to safely remove the heat generated in the fuel without producing excessive fuel temperatures or steam void formations and without closely approaching the hidrodynamic Critical Heat Flux (CHF) (Huda et al. 2001). As the IPR-R1 TRIGA reactor core power is increased, the heat transfer regime from the fuel cladding to the coolant changes from the single phase natural convection regime to subcooled nucleate boiling. The hottest temperature measured in the core channel was 65 oC (Channel 1'), below 111.4 oC, the water saturation temperature for the pressure of 1.5 bar. Therefore, the saturated nucleate boiling regime is not reached. Channel 1' is the closest channel to the centre of the reactor where it is possible to measure the water entrance and exit temperatures. The hottest channel is Channel 0, closer to the centre. With the measured temperature values in the Channel 1', the value of critical flow was evaluated in these two channels. The Bernath correlation was used (Eq. 5) for the calculation of the critical heat flux. With the reactor power of 265 kW operating in steady state, the core inlet temperature was 47oC. The critical flow for the Channel 0 is about 1.6 MW/m2, giving a Departure from Nucleate Boiling Ratio (DNBR2 of 8.5. Figure 22 and Figure 23 show the values of critical flow and DNBR for the two channels. The theoretical values for reactor TRIGA of the University of New York and calculated with the PANTERA code for the IPR-R1 are also shown (General Atomic, 1970) and (Veloso, 2005). The two theoretical calculations gave smaller results than the experiments. These differences are due to the core inlet temperature used in the models.

2CHF/actual local heat flux

Fig. 22. Critical heat flux as a function of the coolant inlet temperature

The minimum DNBR for IPR-R1 TRIGA (DNBR=8.5) is much larger than other TRIGA reactors. The 2 MW McClellen TRIGA calculated by Jensen and Newell (1998) had a DNBR=2.5 and the 3 MW Bangladesh TRIGA has a DNBR=2.8 (Huda and Rahman, 2004).

flux.

**6. Acknowledgment** 

**7. References** 

Filand.

Portuguese).

of Higher Education Personnel (CAPES).

Turquia, October 23-27, 2000.

No. 8, ( May 2002) , pp. 901–912, ISSN 0306-4549.

Mark I Reactor. (GA-9864). San Diego.

from: http://www.ga-esi.com/triga/.

Number TOS210J220. San Diego, CA.

Experimental Investigation of Thermal Hydraulics in the IPR-R1 TRIGA Nuclear Reactor 23

flow is turbulent in all channels near the centre. The temperature measurements above the IPR-R1 core showed that water mixing occurs within the first few centimeters above the top of the core, resulting in an almost uniform water temperature. The temperature at the primary loop suction point at the pool bottom, as shown in Fig. 3, has been found as the lowest temperature in the reactor pool. Pool temperature depends on reactor power, as well as on the external temperature because it affects the heat dissipation rate in the cooling tower. The results can be considered as typical of pool-type research reactor. Further research could be done in the area of boiling heat transfer by using a simulated fuel element heated by electrical current (mock-up). The mock fuel element would eliminate the radiation hazard and allow further thermocouple instrumentation. By using a thermocouple near the fuel element surface, the surface temperature could be measured as a function of the heat

It is suggested to repeat the experiments reported here, by placing a hollow cylinder over the core, with the same diameter of it, to verify the improvement of the mass flow rate by the chimney effect. These experiments can help the designers of the Brazilian research Multipurpose Reactor (RBM), which will be a pool reactor equipped with a chimney to

This research project is supported by the following Brazilian institutions: Nuclear Technology Development Centre (CDTN), Brazilian Nuclear Energy Commission (CNEN), Research Support Foundation of the State of Minas Gerais (FAPEMIG), Brazilian Council for Scientific and Technological Development (CNPq) and Coordination for the Improvement

Bärs, B. & Vaurio, J. (1966). Power Increasing Experiments on a TRIGA Reactor. Technical

Büke, T & Yavuz, H. (2000). Thermal-Hydraulic Analysis of the ITU TRIGA Mark-II Reactor.

CDTN/CNEN - Nuclear Technology Development Centre*/*Brazilian Nuclear Energy

Dalle, H.M., Pereira, C., Souza, R.M.G.P. (2002). Neutronic Calculation to the TRIGA IPR-R1

General Atomic. (1970). Safeguards Summary Report for the New York University TRIGA

General Atomics. (June 2011). TRIGA. In: *TRIGA® Nuclear Reactors,* 07.06.2011*.* Available

Gulf General Atomic. (1972). 15" SST Fuel Element Assembly Instrumented Core. Drawing

University of Helsinki, Department of Technical Physics. Report No. 445. Otaniemi

*Proceeding of 1st Eurasia Conference on Nuclear Science and its Application*. Izmir,

Commission. (2009). Brazilian Multipurpose Reactor (RMB), Preliminary Report of Reactor Engineering Group, General Characteristics and Reactors Reference". (in

reactor using the WIMSD4 and CITATION codes. *Annals of Nuclear Energy*, Vol. 29,

improve the heat removal from the core (CDTN/CNEN, 2009).

The power reactors are projected for a minimum DNBR of 1.3. In routine operation they operated with a DNBR close to 2. The IPR-R1 reactor operates with a great margin of safety at its present power of 250 kW, the maximum heat flux in the hottest fuel is about 8 times lesser than the critical heat flux that would take the hydrodynamic crisis in the fuel cladding. This investigation indicates that the reactor would have an appropriate heat transfer if the reactor operated at a power of about 1 MW.

Fig. 23. DNBR as a function of the coolant inlet temperature
