**3. Methodology**

6 Nuclear Reactors

aluminum-clad fuel elements and 5 stainless steel-clad fuel elements. One of these steel-clad fuel elements is instrumented with three thermocouples along its centreline, and was inserted in the reactor core in order to evaluate the thermal hydraulic performance of the IPR-R1 reactor (Mesquita, 2005). The fuel rod has about 3.5 cm diameter, the active length is about 37 cm closed by graphite slugs at the top and bottom ends which act as axial reflector. The moderating effects are carried out mainly by the zirconium hydride in the mixture, and on a smaller scale by light water coolant. The characteristic of the fuel elements gives a very high negative prompt temperature coefficient, is the main reason of the high inherent safety behavior of the TRIGA reactors. The power level of the reactor is controlled with three

independent control rods: a Regulating rod, a Shim rod, and a Safety rod.

Fig. 5. Core configuration with the rotary specimen rack

#### **3.1 Fuel and core coolant channel temperatures**

Before starting the experiments the thermal power released by the core was calibrated, according with the methodology developed by Mesquita et al. (2007). The calibration method used consisted of the steady-state energy balance of the primary cooling loop. For this balance, the inlet and outlet temperatures and the water flow in this primary cooling loop were measured. The heat transferred through the primary loop was added to the heat leakage from the reactor pool. The temperature measurements lines were calibrated as a whole, including sensors, cables, data acquisition cards and computer. The uncertainties for the temperature measurement circuit were ±0.4 oC for resistance temperature detectors, and ±1.0 oC for thermocouples circuits. The adjusted equations were added to the program of the data acquisition system (DAS). The sensor signs were sent to an amplifier and multiplexing board of the DAS, which also makes the temperature compensation for the thermocouples. The temperatures were monitored in real time on the DAS computer screen. All data were obtained as the average of 120 readings and were recorded together with their standard deviations. The system was developed to monitor and to register the operational parameters once a second in a historical database (Mesquita & Souza, 2010).

The original fuel element at the reactor core position B1 was removed and replaced by an instrumented fuel element. Position B1 is the hottest location in the core (largest thermal power production), according to the neutronic calculation (Dalle et al., 2002). The instrumented fuel element is in all aspects identical to standard fuel elements, except that it is equipped with three chromel-alumel thermocouples (K type), embedded in the fuel meat. The sensitive tips of the thermocouples are located along the fuel centreline. Their axial position is one at the half-height of the fuel meat and the other two 2.54 mm above and 2.54 mm below. Figure 6 shows the diagram of the instrumented fuel element and the Table I presents its main characteristics (Gulf General Atomic, 1972). Figure 7 shows the instrumented fuel element and one thermocouple inside a core channel.

Fig. 6. Diagram of the instrumented fuel element

Experimental Investigation of Thermal Hydraulics in the IPR-R1 TRIGA Nuclear Reactor 9

and outlet temperatures. Two type K (chromel–alumel) thermocouples fixed in two rigid aluminum probes (7.9 mm of diameter), were inserted into the core in two channels close to position B1 (Channel 1 and 1' in Fig. 4 and Fig. 5) and measured the inlet and outlet coolant channel temperatures. The probes penetrated axially the channels through small holes in the core upper grid plate. The probes were positioned in diametrically opposite channels, so that when a probe measured the channel entrance temperature, the other one registered the channel exit temperature. In a subsequent run, the probe positions were inverted. This procedure was used also for the Channels 1', 2', 3', 4' and 5' (Fig. 5). There is no hole in the top grid plate in the direction of the Channel 0; so it was not possible to measure its temperature. The inlet and outlet temperatures in Channel 0 were considered as being the same of Channel 1. For the other channels there are holes in the top grid plate where it was possible to insert the temperature probes. To found the bulk coolant temperature axial profile at hot channel, with the reactor operating at 250 kW, the probe that measures the channel inlet temperature was raised in steps of 10 cm and the temperature was monitored. The same procedure was done with the reactor operating at 100 kW, but the probe was

The mass flow rate through the core coolant channels was determined indirectly from the heat balance across each channel using measurements of the water entrance and exit temperatures. Although the channels are laterally open, in this work cross flow or mass transfer between adjacent channels was ignored. Inlet and outlet coolant temperatures in channels were measured with two rigid aluminum probes with thermocouples. They were inserted in the upper grid plate holes (Fig. 5). Figure 8 illustrates schematically the general natural convection process established by the fuel elements bounding one flow channel in the core. The core coolant channels extend from the bottom grid plate to the top grid plate. The cooling water flows through the holes in the bottom grid plate, passes through the lower unheated region of the element, flows upwards through the active region, passes through the upper unheated region, and finally leaving the channel through the differential area between a triangular spacer block on the top of the fuel element and a round hole in the grid. As mentioned, in natural convection the driving force is supplied by the buoyancy of

In a typical TRIGA flow channel entire fuel element is cooled by single phase convection as long as the maximum wall temperature is kept below that required to initiate boiling. However, at higher power levels the inlet and outlet regions of the core, where the heat fluxes are the lowest, the channels are cooled by single phase convection. In the central region, where the axial heat flux is highest, the mode of heat transfer is predominantly

The channel heating process is the result of the thermal fraction contributions of the perimeter of each fuel around the channel. So there was an average power of 4.518 kW dissipated in each stainless steel cladding fuel element and 4.176 kW dissipated in each aluminum cladding fuel element at 265 kW core total power. The values are multiplied by the fuel element axial power distribution and core radial power distribution factors as

raised in steps of 5 cm.

**3.2 Hydraulic parameters of the coolant** 

the heated water in the core channels.

shown in profiles of Fig. 9.

subcooled boiling (Rao et al., 1988 and Mesquita et al. 2011).


Table 1. Instrumented fuel element features

Fig. 7. IPR-R1 core top view with the instrumented fuel element in ring B and one probe with thermocouple inside the core.

The instrumented fuel element was replaced to new positions and measures the fuel temperature in each one of the core fuel rings (from B to F). At the same way, two thermocouples were replaced to channels close to the instrumented fuel element to measure the coolant channel temperature. Experiments were carried out with the power changing from about 50 kW to 250 kW in 50 kW steps for each position of the instrumented fuel element. The fuel and coolant temperatures were monitored as function of the thermal power and position in the core.

In the TRIGA type reactors the buoyancy force induced by the density differential across the core maintains the water circulation through the core. Countering this buoyancy force are the pressure losses due to the contraction and expansion at the entrance and exit of the core as well as the acceleration and friction pressure losses in the flow channels. Direct measurement of the flow rate in a coolant channel is difficult because of the bulky size and low accuracy of flow meters. The flow rate through the channel may be determined indirectly from the heat balance across the channel using measurements of the water inlet and outlet temperatures. Two type K (chromel–alumel) thermocouples fixed in two rigid aluminum probes (7.9 mm of diameter), were inserted into the core in two channels close to position B1 (Channel 1 and 1' in Fig. 4 and Fig. 5) and measured the inlet and outlet coolant channel temperatures. The probes penetrated axially the channels through small holes in the core upper grid plate. The probes were positioned in diametrically opposite channels, so that when a probe measured the channel entrance temperature, the other one registered the channel exit temperature. In a subsequent run, the probe positions were inverted. This procedure was used also for the Channels 1', 2', 3', 4' and 5' (Fig. 5). There is no hole in the top grid plate in the direction of the Channel 0; so it was not possible to measure its temperature. The inlet and outlet temperatures in Channel 0 were considered as being the same of Channel 1. For the other channels there are holes in the top grid plate where it was possible to insert the temperature probes. To found the bulk coolant temperature axial profile at hot channel, with the reactor operating at 250 kW, the probe that measures the channel inlet temperature was raised in steps of 10 cm and the temperature was monitored. The same procedure was done with the reactor operating at 100 kW, but the probe was raised in steps of 5 cm.

#### **3.2 Hydraulic parameters of the coolant**

8 Nuclear Reactors

Fig. 7. IPR-R1 core top view with the instrumented fuel element in ring B and one probe

The instrumented fuel element was replaced to new positions and measures the fuel temperature in each one of the core fuel rings (from B to F). At the same way, two thermocouples were replaced to channels close to the instrumented fuel element to measure the coolant channel temperature. Experiments were carried out with the power changing from about 50 kW to 250 kW in 50 kW steps for each position of the instrumented fuel element. The fuel and coolant temperatures were monitored as function of the thermal

In the TRIGA type reactors the buoyancy force induced by the density differential across the core maintains the water circulation through the core. Countering this buoyancy force are the pressure losses due to the contraction and expansion at the entrance and exit of the core as well as the acceleration and friction pressure losses in the flow channels. Direct measurement of the flow rate in a coolant channel is difficult because of the bulky size and low accuracy of flow meters. The flow rate through the channel may be determined indirectly from the heat balance across the channel using measurements of the water inlet

Table 1. Instrumented fuel element features

with thermocouple inside the core.

power and position in the core.

**Parameter Value**  Heated length 38.1 cm Outside diameter 3.76 cm Active outside area 450.05 cm2 Fuel outside area (U-ZrH1.6) 434.49 cm2 Fuel element active volume 423.05 cm3 Fuel volume (U-ZrH1.6) 394.30 cm3 Power (total of the core = 265 kW) 4.518 kW

> The mass flow rate through the core coolant channels was determined indirectly from the heat balance across each channel using measurements of the water entrance and exit temperatures. Although the channels are laterally open, in this work cross flow or mass transfer between adjacent channels was ignored. Inlet and outlet coolant temperatures in channels were measured with two rigid aluminum probes with thermocouples. They were inserted in the upper grid plate holes (Fig. 5). Figure 8 illustrates schematically the general natural convection process established by the fuel elements bounding one flow channel in the core. The core coolant channels extend from the bottom grid plate to the top grid plate. The cooling water flows through the holes in the bottom grid plate, passes through the lower unheated region of the element, flows upwards through the active region, passes through the upper unheated region, and finally leaving the channel through the differential area between a triangular spacer block on the top of the fuel element and a round hole in the grid. As mentioned, in natural convection the driving force is supplied by the buoyancy of the heated water in the core channels.

> In a typical TRIGA flow channel entire fuel element is cooled by single phase convection as long as the maximum wall temperature is kept below that required to initiate boiling. However, at higher power levels the inlet and outlet regions of the core, where the heat fluxes are the lowest, the channels are cooled by single phase convection. In the central region, where the axial heat flux is highest, the mode of heat transfer is predominantly subcooled boiling (Rao et al., 1988 and Mesquita et al. 2011).

> The channel heating process is the result of the thermal fraction contributions of the perimeter of each fuel around the channel. So there was an average power of 4.518 kW dissipated in each stainless steel cladding fuel element and 4.176 kW dissipated in each aluminum cladding fuel element at 265 kW core total power. The values are multiplied by the fuel element axial power distribution and core radial power distribution factors as shown in profiles of Fig. 9.

Experimental Investigation of Thermal Hydraulics in the IPR-R1 TRIGA Nuclear Reactor 11

codes. The products are multiplied by the fractions of the perimeters of each fuel in contact with the coolant in each channel. The two hottest channels in the core are Channel 0 and Channel 1'. Channel 0 is located closer to the core centre, where density of neutron flux is larger, but there is no hole in the top grid plate in the direction of this channel. Table 1 gives the geometric data of the coolant channels and the percentage of contribution relative to

each fuel element to the channels power (Veloso, 2005 and Mesquita, 2005).

Fig. 10. Core coolant channels geometry and radial power distribution

**Wetted Perimeter [cm]** 

**Heated Perimeter [cm]** 

0 1.5740 5.9010 3.9060 1.0669 1.00 1' 8.2139 17.6427 15.1556 1.8623 3.70 2' 5.7786 11.7456 11.7456 1.9679 2.15 3' 5.7354 11.7181 11.7181 1.9578 1.83 4' 5.6938 11.7181 8.6005 1.9436 1.13 5' 3.9693 10.8678 3.1248 1.4609 0.35

Table 2. Channel geometry and hydraulic parameters (Veloso, 2005; Mesquita, 2005)

**Hydraulic Diameter [cm]** 

**Channel Power [%]** 

**Channel Number** 

**Area [cm2]** 

Fig. 8. A scheme of one flow channel in the TRIGA core

Fig. 9. Core radial and fuel element axial power profiles

The power axial distribution factor in the fuel is 1.25, according with Marcum (2008). Figure 10 shows in detail the coolant channels geometry. The core radial power distribution factors, shown in Fig. 10, were calculated by Dalle et al. (2002) using WIMS-D4 and CITATION

Fig. 8. A scheme of one flow channel in the TRIGA core

Fig. 9. Core radial and fuel element axial power profiles

The power axial distribution factor in the fuel is 1.25, according with Marcum (2008). Figure 10 shows in detail the coolant channels geometry. The core radial power distribution factors, shown in Fig. 10, were calculated by Dalle et al. (2002) using WIMS-D4 and CITATION codes. The products are multiplied by the fractions of the perimeters of each fuel in contact with the coolant in each channel. The two hottest channels in the core are Channel 0 and Channel 1'. Channel 0 is located closer to the core centre, where density of neutron flux is larger, but there is no hole in the top grid plate in the direction of this channel. Table 1 gives the geometric data of the coolant channels and the percentage of contribution relative to each fuel element to the channels power (Veloso, 2005 and Mesquita, 2005).

Fig. 10. Core coolant channels geometry and radial power distribution


Table 2. Channel geometry and hydraulic parameters (Veloso, 2005; Mesquita, 2005)

Experimental Investigation of Thermal Hydraulics in the IPR-R1 TRIGA Nuclear Reactor 13

As the power in the IPR-R1 TRIGA core is increased, nucleation begins to occur on the fuel rod surfaces. The typical pool boiling curve (Fig. 11) is represented on a log-log plot of heat flux versus wall superheat (*Tsur – Tsat*). At low values of Δ*Tsat* the curve is fairly linear, hence the convective heat transfer coefficient (*h*) is relatively constant. There is no bubble formation and the heat transfer occurs by liquid natural convection. At about ten to twenty degrees above saturation the heat flux increases rapidly with the increasing of the wall temperature. The increase in heat transfer is due to nucleate boiling. The formation of vapor bubbles increases the turbulence near the heated surface and allows mixing of the coolant fluid in the film region, thus enhancing the heat transfer rate (Haag, 1971). From the shape of the curve, it can be seen that the heat transfer coefficient increases dramatically in the

Fig. 11. Typical pool boiling curve for water under atmospheric pressure

Whenever the surface temperature of a solid exceeds the saturation temperature, local boiling may occur even if the bulk water temperature is below the saturation temperature. The water temperature in the boundary layer on the heated surface can become sufficiently high so that subcooled pool boiling takes place. The bubbles will be condensed upon leaving this boundary layer region because the bulk water is below the saturation temperature. By increasing the surface temperature, the heat flux can reach the critical heat flux where the film boiling occurs. At this point the bubbles become so numerous that they form an insulating layer of steam around the fuel element and the heat flux is reduced significantly.

**3.5 Critical heat flux and DNBR** 

boiling regime.

The mass flow rate in the hydraulic channel ( *m* ) in [kg/s] is given indirectly from the thermal balance along the channel using measurements of the water inlet and outlet temperatures:

$$
\dot{m} = \frac{q\_c}{c\_p \Delta T} \tag{1}
$$

Where *qc* is the power supplied to the channel [kW], *cp* is the isobaric specific heat of the water [J/kgK] and Δ*T* is the temperature difference along the channel [oC]. The mass flux *G* is given by: *G* = *m* / *channel area*. The velocity *u* is given by *u = G / ρ*, where *ρ* is the water density (995 kg/m3). The values of the water thermodynamic properties were obtained as function of the bulk water temperature at the channel for the pressure 1.5 bar (Wagner & Kruse, 1988) Reynolds number (Re), used to characterize the flow regime, is given by:

$$\text{Re} = \frac{GD\_w}{\mu} \tag{2}$$

Where *G* is the mass flux in [kg/m2s], *Dw* is the hydraulic diameter in [m] and *μ* is the dynamic viscosity [kg/ms].

#### **3.3 Pool temperatures**

Nine thermocouples and one platinum resistance thermometer (PT-100) were used to monitoring the reactor pool temperature. The thermocouples were positioned in a vertical aluminum probe and the first thermocouple was 143 mm above the core top grid plate. The reactor operated during a period of about eight hours at a thermal power of 265 kW before the steady state was obtained. The forced cooling system was turned on during the operation. This experiment is important to understand the behavior of the water temperature in the pool and evaluate the height of the chimney effect.

#### **3.4 Temperatures with the forced cooling system turned off**

The power of the IPR-R1 TRIGA was raised in steps of about 25 kW until to reach 265 kW. The forced cooling system of the reactor pool was turned off during the tests. The increase of the power was allowed only when all the desired quantities had been measured and the given limits were not exceeded. After the reactor power level was reached, the reactor was maintained at that power for about 15 min, so the entire steady-state conditions were not reached in the core and coolant. The fuel temperature data was obtained by using the instrumented fuel element. The fuel temperature measurements were taken at location B1 of the core (hottest position). The outlet temperature in the channel was measured with thermocouple inserted near the B1 position. One platinum resistance thermometer measured the water temperature in the upper part of the reactor tank. Two thermocouples measured the ambient temperatures around the reactor pool. The IPR-R1 reactor has a rotary specimen rack outside the reactor core for sample irradiation. It is composed by forty irradiation channels in a cylindrical geometry. One type K thermocouple was put during the experiment in Position 40 of the rotary specimen rack (Fig. 5).

#### **3.5 Critical heat flux and DNBR**

12 Nuclear Reactors

The mass flow rate in the hydraulic channel ( *m* ) in [kg/s] is given indirectly from the thermal balance along the channel using measurements of the water inlet and outlet

> *c p <sup>q</sup> <sup>m</sup>*

Where *qc* is the power supplied to the channel [kW], *cp* is the isobaric specific heat of the water [J/kgK] and Δ*T* is the temperature difference along the channel [oC]. The mass flux *G* is given by: *G* = *m* / *channel area*. The velocity *u* is given by *u = G / ρ*, where *ρ* is the water density (995 kg/m3). The values of the water thermodynamic properties were obtained as function of the bulk water temperature at the channel for the pressure 1.5 bar (Wagner &

Kruse, 1988) Reynolds number (Re), used to characterize the flow regime, is given by:

Re *GDw* 

Where *G* is the mass flux in [kg/m2s], *Dw* is the hydraulic diameter in [m] and *μ* is the

Nine thermocouples and one platinum resistance thermometer (PT-100) were used to monitoring the reactor pool temperature. The thermocouples were positioned in a vertical aluminum probe and the first thermocouple was 143 mm above the core top grid plate. The reactor operated during a period of about eight hours at a thermal power of 265 kW before the steady state was obtained. The forced cooling system was turned on during the operation. This experiment is important to understand the behavior of the water

The power of the IPR-R1 TRIGA was raised in steps of about 25 kW until to reach 265 kW. The forced cooling system of the reactor pool was turned off during the tests. The increase of the power was allowed only when all the desired quantities had been measured and the given limits were not exceeded. After the reactor power level was reached, the reactor was maintained at that power for about 15 min, so the entire steady-state conditions were not reached in the core and coolant. The fuel temperature data was obtained by using the instrumented fuel element. The fuel temperature measurements were taken at location B1 of the core (hottest position). The outlet temperature in the channel was measured with thermocouple inserted near the B1 position. One platinum resistance thermometer measured the water temperature in the upper part of the reactor tank. Two thermocouples measured the ambient temperatures around the reactor pool. The IPR-R1 reactor has a rotary specimen rack outside the reactor core for sample irradiation. It is composed by forty irradiation channels in a cylindrical geometry. One type K thermocouple was put during the

temperature in the pool and evaluate the height of the chimney effect.

**3.4 Temperatures with the forced cooling system turned off** 

experiment in Position 40 of the rotary specimen rack (Fig. 5).

*c T* (1)

(2)

temperatures:

dynamic viscosity [kg/ms].

**3.3 Pool temperatures** 

As the power in the IPR-R1 TRIGA core is increased, nucleation begins to occur on the fuel rod surfaces. The typical pool boiling curve (Fig. 11) is represented on a log-log plot of heat flux versus wall superheat (*Tsur – Tsat*). At low values of Δ*Tsat* the curve is fairly linear, hence the convective heat transfer coefficient (*h*) is relatively constant. There is no bubble formation and the heat transfer occurs by liquid natural convection. At about ten to twenty degrees above saturation the heat flux increases rapidly with the increasing of the wall temperature. The increase in heat transfer is due to nucleate boiling. The formation of vapor bubbles increases the turbulence near the heated surface and allows mixing of the coolant fluid in the film region, thus enhancing the heat transfer rate (Haag, 1971). From the shape of the curve, it can be seen that the heat transfer coefficient increases dramatically in the boiling regime.

Fig. 11. Typical pool boiling curve for water under atmospheric pressure

Whenever the surface temperature of a solid exceeds the saturation temperature, local boiling may occur even if the bulk water temperature is below the saturation temperature. The water temperature in the boundary layer on the heated surface can become sufficiently high so that subcooled pool boiling takes place. The bubbles will be condensed upon leaving this boundary layer region because the bulk water is below the saturation temperature. By increasing the surface temperature, the heat flux can reach the critical heat flux where the film boiling occurs. At this point the bubbles become so numerous that they form an insulating layer of steam around the fuel element and the heat flux is reduced significantly.

Experimental Investigation of Thermal Hydraulics in the IPR-R1 TRIGA Nuclear Reactor 15

Figure 13 shows the results of fuel temperature versus reactor thermal power. In the

Fig. 12. Core temperature radial profile at 265 kW thermal power

experiment the instrumented fuel element was positioned in each core ring.

Fig. 13. Fuel temperature as function of the reactor power in all core rings

**4.2.1 Outlet coolant temperature as function of the thermal power** 

The experimental coolant exit temperature for each core ring is shown in Fig 14 as a function of the reactor power. The aluminum probe with thermocouple was inserted in each hole at top grid plate, and the coolant inlet temperature was about 38 oC in all measurements.

**4.2 Core temperature** 

The critical heat flux is the maximum heat flux that a saturated fluid can absorb before acquiring more enthalpy than can be dissipated into its surroundings.

In the fully developed nucleate boiling regime, it is possible to increase the heat flux without an appreciable change in the surface temperature until the point of Departure from Nucleate Boiling (DNB). At this point, the bubble motion on the surface becomes so violent that a hydrodynamic crisis occurs with the formation of a continuous vapor film in the surface and the Critical Heat Flux (CHF) is reached. In subcooled boiling the CHF is a function of the coolant velocity, the degree of subcooling, and the pressure. There are a lot of correlations to predict the CHF. The correlation done by Bernath found in Lamarsh and Baratta (2001) was used to predicts CHF in the subcooled boiling region and is based on the critical wall superheat condition at burnout and turbulent mixing convective heat transfer. Bernath's equation gives the minimum results so it is the most conservative. It is given by:

$$\boldsymbol{q\_{crit}} = \boldsymbol{h\_{crit}}(\boldsymbol{T\_{crit}} - \boldsymbol{T\_{f}}) \,\tag{3}$$

where,

$$h\_{crit} = 61.84 \frac{D\_w}{D\_w + D\_i} + 0.01863 \frac{23.53}{D\_w^{0.6}} \mu \tag{4}$$

and,

$$T\_{crit} = 57 \ln(p - 54) \frac{p}{p + 0.1034} + 283.7 - \frac{u}{1.219} \tag{5}$$

'' *crit q* is the critical heat flux [W/m2], *hcrit* is the critical coefficient of heat transfer [W/m2K], *Tcrit* is the critical surface temperature [oC], *Tf* is the bulk fluid temperature [oC], *p* is the pressure [MPa], *u* is the fluid velocity [m/s] (*u* = *m* /*channel area/water density*), *Dw* is the wet hydraulic diameter [m], *Di* is the diameter of heat source [m]. This correlation is valid for circular, rectangular and annular channels, pressure of 0.1 to 20.6 MPa, velocity between 1 to 16 m/s and hydraulic diameter of 0.36 to 1.7 cm.

#### **4. Results**

#### **4.1 Fuel temperature**

Before beginning the experiments, the calibration of the thermal power released by the core were performed, and a power of 265 kW was found when the neutronic linear channel was indicating the power of 250 kW.

Figure 12 shows the radial power profile (neutron flux) calculated by Dalle et al. (2002) using the TRIGPOW code, the experimental fuel radial temperature profile, and the inlet/outlet coolant temperatures in the channel closest to the instrumented element. The theoretical results, for the IPR-R1 TRIGA, calculated by Veloso (2005) using the PANTERA code and the experimental results found in the ITU TRIGA Mark II reactor at the Istanbul University were also plotted (Özkul & Durmayaz, 2000).. All data are for the power of 265 kW.

Fig. 12. Core temperature radial profile at 265 kW thermal power

Figure 13 shows the results of fuel temperature versus reactor thermal power. In the experiment the instrumented fuel element was positioned in each core ring.

Fig. 13. Fuel temperature as function of the reactor power in all core rings

#### **4.2 Core temperature**

14 Nuclear Reactors

The critical heat flux is the maximum heat flux that a saturated fluid can absorb before

In the fully developed nucleate boiling regime, it is possible to increase the heat flux without an appreciable change in the surface temperature until the point of Departure from Nucleate Boiling (DNB). At this point, the bubble motion on the surface becomes so violent that a hydrodynamic crisis occurs with the formation of a continuous vapor film in the surface and the Critical Heat Flux (CHF) is reached. In subcooled boiling the CHF is a function of the coolant velocity, the degree of subcooling, and the pressure. There are a lot of correlations to predict the CHF. The correlation done by Bernath found in Lamarsh and Baratta (2001) was used to predicts CHF in the subcooled boiling region and is based on the critical wall superheat condition at burnout and turbulent mixing convective heat transfer. Bernath's

( ) *crit crit crit <sup>f</sup> q hT T* , (3)

(4)

(5)

0.6

acquiring more enthalpy than can be dissipated into its surroundings.

equation gives the minimum results so it is the most conservative. It is given by:

23.53 61.84 0.01863 *<sup>w</sup> crit*

57 ln( 54) 283.7 0.1034 1.219 *crit <sup>p</sup> <sup>u</sup> T p p* 

*crit q* is the critical heat flux [W/m2], *hcrit* is the critical coefficient of heat transfer [W/m2K], *Tcrit* is the critical surface temperature [oC], *Tf* is the bulk fluid temperature [oC], *p* is the pressure [MPa], *u* is the fluid velocity [m/s] (*u* = *m* /*channel area/water density*), *Dw* is the wet hydraulic diameter [m], *Di* is the diameter of heat source [m]. This correlation is valid for circular, rectangular and annular channels, pressure of 0.1 to 20.6 MPa, velocity between 1 to

Before beginning the experiments, the calibration of the thermal power released by the core were performed, and a power of 265 kW was found when the neutronic linear channel was

Figure 12 shows the radial power profile (neutron flux) calculated by Dalle et al. (2002) using the TRIGPOW code, the experimental fuel radial temperature profile, and the inlet/outlet coolant temperatures in the channel closest to the instrumented element. The theoretical results, for the IPR-R1 TRIGA, calculated by Veloso (2005) using the PANTERA code and the experimental results found in the ITU TRIGA Mark II reactor at the Istanbul University were also plotted (Özkul & Durmayaz, 2000).. All data are for the power of

*<sup>D</sup> h u D D <sup>D</sup>*

*w i w*

''

16 m/s and hydraulic diameter of 0.36 to 1.7 cm.

where,

and,

''

**4. Results** 

265 kW.

**4.1 Fuel temperature** 

indicating the power of 250 kW.

#### **4.2.1 Outlet coolant temperature as function of the thermal power**

The experimental coolant exit temperature for each core ring is shown in Fig 14 as a function of the reactor power. The aluminum probe with thermocouple was inserted in each hole at top grid plate, and the coolant inlet temperature was about 38 oC in all measurements.

Experimental Investigation of Thermal Hydraulics in the IPR-R1 TRIGA Nuclear Reactor 17

The experimental bulk coolant temperatures profile in Channel 1 is shown in Fig. 16 as a function of the axial position, for the powers of 265 kW and 106 kW. Figure 16 shows also the curve predicted from the theoretical model using the PANTERA code at 265 kW (Veloso, 2005). The figure shows also the experimental results for other TRIGA reactors Bärs &

The experimental temperature profile along the coolant is different from that predicted from the theoretical model. Ideally, the coolant temperature would increase along the entire length of the channel, because heat is being added to the water by all fuel regions in the channel. Experimentally, the water temperature reaches a maximum near the middle length and then decreases along the remaining channel. The shape of the experimental curves is similar to the axial power distribution within the fuel rod as shown in Fig. 9. Although Channel 1 is located beside the control rod, the axial temperature profile was not influenced by a possible deformation of the neutron flux caused by this rod, because it was in its upper position, i.e. outside the core. The actual coolant flow is quite different probably, because of

**4.2.3 Axial temperature profile in the hot channel** 

Vaurio, 1966; Haag, 1971) and (Büke & Yavuz, 2000).

the inflow of water from the core sides (colder than its centre).

Fig. 16. Axial bulk coolant temperature profile along the Channel 1

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

**4.2.4 Thermal hydraulic parameters of coolant channels** 

channels near the core centre.

Fig. 14. Outlet coolant temperature as function of the thermal power

#### **4.2.2 Radial temperature profile along the core coolant channels**

Figure 15 shows the radial core coolant temperature profiles (inlet/outlet channel temperatures) at 265 kW. Theoretical results using the PANTERA code are also shown in the figure (Veloso, 2005).

Fig. 15. Radial temperature profile in the core coolant channels at 265 kW thermal power
