**1. Introduction**

Rising concerns about global warming and energy security have spurred a revival of interest in nuclear energy, leading to a "nuclear power renaissance" in countries the world over. In Brazil, the nuclear renaissance can be seen in the completion of construction of its third nuclear power plant and in the government's decision to design and build the Brazilian Multipurpose research Reactor (RMB). The role of nuclear energy in Brazil is complementary to others sources. Presently two Nuclear Power Plants are in operation (Angra 1 and 2) with a total of 2000 MWe that accounts for the generation of approximately 3% of electric power consumed in Brazil. A third unity (Angra 3) is under construction. Even though with such relatively small nuclear park, Brazil has one of the biggest world nuclear resources, being the sixth natural uranium resource in the world and has a fuel cycle industry capable to provide fuel elements. Brazil has four research reactors in operation: the MB-01, a 0.1 kW critical facility; the IEA-R1, a 5 MW pool type reactor; the Argonauta, a 500 W Argonaut type reactor and the IPR-R1, a 100 kW TRIGA Mark I type reactor. They were constructed mainly for using in education, radioisotope production and nuclear research.

Understanding the behavior of the operational parameters of nuclear reactors allow the development of improved analytical models to predict the fuel temperature, and contributing to their safety. The recent natural disaster that caused damage in four reactors at the Fukushima nuclear power plant shows the importance of studies and experiments on natural convection to remove heat from the residual remaining after the shutdown. Experiments, developments and innovations used for research reactors can be later applied to larger power reactors. Their relatively low cost allows research reactors to provide an excellent testing ground for the reactors of tomorrow.

The IPR-R1 TRIGA Mark-I research reactor is located at the Nuclear Technology Development Centre - CDTN (Belo Horizonte/Brazil*)*, a research institute of the Brazilian Nuclear Energy Commission - CNEN. The IPR-R1 reached its first criticality on November

Experimental Investigation of Thermal Hydraulics in the IPR-R1 TRIGA Nuclear Reactor 3

(General Atomics, 2011). Figure 1 shows two photographs of the pool and the core with the

The IPR-R1 TRIGA reactor core is placed at the bottom of an open tank of about 6m height and 2m diameter. The tank is filled with approximately 18 m2 of water able to assure an adequate radioactive shielding, as shown in Fig. 2. The reactor is licensed to operate at a maximum steady-state thermal power level of 100 kW, but the core and the instrumentation are configured to 250 kW, and waiting the definitive license to operate in this new power. Some of the experiments reported here were performed at power operation of 250 kW. For these experiments was obtained a provisional license for operation to this new power.

The reactor core is cooled by water natural circulation. Cooling water passage through the top plate is provided by the differential area between a triangular spacer block on top of fuel element and the round hole in the grid. A heat removal system is provided for removing heat from the reactor pool water. The water is pumped through a heat exchanger, where the

IPR-R1 TRIGA reactor in operation.

Fig. 1. IPR-R1 TRIGA reactor pool and core

1960 with a core configuration containing 56 aluminum clad standard TRIGA fuel elements, and a maximum thermal power of 30 kW. In order to upgrade the IPR-R1 reactor power, nine stainless steel clad fuel elements were purchased in 1971. One of these fuel elements was instrumented in the centreline with three type K thermocouples. On December 2000, four of these stainless steel clad fuel elements were placed into the core allowing to upgrading the nominal power to 250 kW. In 2004 the instrumented fuel element (IF) was inserted into all core rings and monitored the fuel temperature, allowing heat transfer investigations at several operating powers, including the maximum power of 250 kW (Mesquita, 2005). The basic safety limit for the TRIGA reactor system is the fuel temperature, both in steady-state and pulse mode operation. The time-dependence of temperature was not considered here, hence only the steady-state temperature profile was studied.

This chapter presents the experiments performed in the IPR-R1 reactor for monitoring some thermal hydraulic parameters in the fuel, pool and core coolant channels. The fuel temperature as a function of reactor power was monitored in all core rings. The radial and axial temperature profile, coolant velocity, mass flow rate and Reynolds's number in coolant channels were monitored in all core channels. It also presents a prediction for the critical heat flux (CHF) in the fuel surface at hot channel. Data from the instrumented fuel element, pool, and bulk coolant temperature distribution were compared with the theoretical model and results from other TRIGA reactors. A data acquisition system was developed to provide a friendly interface for monitoring all operational parameters. The system performs the temperature compensation for the thermocouples. Information displayed in real-time was recorded on hard disk in a historical database (Mesquita & Souza, 2008). The data obtained during the experiments provide an excellent picture of the IPR-R1 reactor's thermal performance. The experiments confirm the efficiency of natural circulation in removing the heat produced in the reactor core by nuclear fission (Mesquita & Rezende, 2010).

### **2. The IPR-R1 reactor**

The IPR-R1 TRIGA (*Instituto de Pesquisas Radiativas* - Reactor 1, Training Research Isotope production, General Atomic) is a typical TRIGA Mark I light-water and open pool type reactor. The fuel elements in the reactor core are cooled by water natural circulation. The basic parameter which allows TRIGA reactors to operate safely during either steady-state or transient conditions is the prompt negative temperature coefficient associated with the TRIGA fuel and core design. This temperature coefficient allows great freedom in steady state and transient operations. TRIGA reactors are the most widely used research reactor in the world. There is an installed base of over sixty-five facilities in twenty-four countries on five continents. General Atomics (GA), the supplier of TRIGA research reactors, since late 50's continues to design and install TRIGA reactors around the world, and has built TRIGA reactors in a variety of configurations and capabilities, with steady state thermal power levels ranging from 100 kW to 16 MW. TRIGA reactors are used in many diverse applications, including production of radioisotopes for medicine and industry, treatment of tumors, nondestructive testing, basic research on the properties of matter, and for education and training. The TRIGA reactor is the only nuclear reactor in this category that offers true "inherent safety", rather than relying on "engineered safety". It is possible due to the unique properties of GA's uranium-zirconium hydride fuel, which provides incomparable safety characteristics, which also permit flexibility in sitting, with minimal environmental effects

1960 with a core configuration containing 56 aluminum clad standard TRIGA fuel elements, and a maximum thermal power of 30 kW. In order to upgrade the IPR-R1 reactor power, nine stainless steel clad fuel elements were purchased in 1971. One of these fuel elements was instrumented in the centreline with three type K thermocouples. On December 2000, four of these stainless steel clad fuel elements were placed into the core allowing to upgrading the nominal power to 250 kW. In 2004 the instrumented fuel element (IF) was inserted into all core rings and monitored the fuel temperature, allowing heat transfer investigations at several operating powers, including the maximum power of 250 kW (Mesquita, 2005). The basic safety limit for the TRIGA reactor system is the fuel temperature, both in steady-state and pulse mode operation. The time-dependence of temperature was

not considered here, hence only the steady-state temperature profile was studied.

heat produced in the reactor core by nuclear fission (Mesquita & Rezende, 2010).

The IPR-R1 TRIGA (*Instituto de Pesquisas Radiativas* - Reactor 1, Training Research Isotope production, General Atomic) is a typical TRIGA Mark I light-water and open pool type reactor. The fuel elements in the reactor core are cooled by water natural circulation. The basic parameter which allows TRIGA reactors to operate safely during either steady-state or transient conditions is the prompt negative temperature coefficient associated with the TRIGA fuel and core design. This temperature coefficient allows great freedom in steady state and transient operations. TRIGA reactors are the most widely used research reactor in the world. There is an installed base of over sixty-five facilities in twenty-four countries on five continents. General Atomics (GA), the supplier of TRIGA research reactors, since late 50's continues to design and install TRIGA reactors around the world, and has built TRIGA reactors in a variety of configurations and capabilities, with steady state thermal power levels ranging from 100 kW to 16 MW. TRIGA reactors are used in many diverse applications, including production of radioisotopes for medicine and industry, treatment of tumors, nondestructive testing, basic research on the properties of matter, and for education and training. The TRIGA reactor is the only nuclear reactor in this category that offers true "inherent safety", rather than relying on "engineered safety". It is possible due to the unique properties of GA's uranium-zirconium hydride fuel, which provides incomparable safety characteristics, which also permit flexibility in sitting, with minimal environmental effects

**2. The IPR-R1 reactor** 

This chapter presents the experiments performed in the IPR-R1 reactor for monitoring some thermal hydraulic parameters in the fuel, pool and core coolant channels. The fuel temperature as a function of reactor power was monitored in all core rings. The radial and axial temperature profile, coolant velocity, mass flow rate and Reynolds's number in coolant channels were monitored in all core channels. It also presents a prediction for the critical heat flux (CHF) in the fuel surface at hot channel. Data from the instrumented fuel element, pool, and bulk coolant temperature distribution were compared with the theoretical model and results from other TRIGA reactors. A data acquisition system was developed to provide a friendly interface for monitoring all operational parameters. The system performs the temperature compensation for the thermocouples. Information displayed in real-time was recorded on hard disk in a historical database (Mesquita & Souza, 2008). The data obtained during the experiments provide an excellent picture of the IPR-R1 reactor's thermal performance. The experiments confirm the efficiency of natural circulation in removing the (General Atomics, 2011). Figure 1 shows two photographs of the pool and the core with the IPR-R1 TRIGA reactor in operation.

Fig. 1. IPR-R1 TRIGA reactor pool and core

The IPR-R1 TRIGA reactor core is placed at the bottom of an open tank of about 6m height and 2m diameter. The tank is filled with approximately 18 m2 of water able to assure an adequate radioactive shielding, as shown in Fig. 2. The reactor is licensed to operate at a maximum steady-state thermal power level of 100 kW, but the core and the instrumentation are configured to 250 kW, and waiting the definitive license to operate in this new power. Some of the experiments reported here were performed at power operation of 250 kW. For these experiments was obtained a provisional license for operation to this new power.

The reactor core is cooled by water natural circulation. Cooling water passage through the top plate is provided by the differential area between a triangular spacer block on top of fuel element and the round hole in the grid. A heat removal system is provided for removing heat from the reactor pool water. The water is pumped through a heat exchanger, where the

Experimental Investigation of Thermal Hydraulics in the IPR-R1 TRIGA Nuclear Reactor 5

Fig. 3. IPR-R1 TRIGA reactor cooling system and instrumentation distribution

The prototypical cylindrical fuel elements are a homogeneous alloy of zirconium hydride (neutron moderator) and uranium enriched at 20% in 235U. The reactor core has 58

Fig. 4. Simplified core diagram

heat is transferred from the primary to the secondary loop. The secondary loop water is cooled in an external cooling tower. Figure 3 shows the forced cooling system, which transfers the heat generated in the reactor core to a water-to-water heat exchanger. The secondary cooling system transfers the reactor core heat from the heat exchanger to a cooling tower. In the diagram is shown also the instrumentation distribution and the forced and natural circulation paths in the pool.

Fig. 2. IPR-R1 TRIGA reactor pool and core

A simplified view of the IPR R1 TRIGA core configuration is shown in the Fig. 4. As shown in the diagram there are small holes in the core upper grid plate. These holes were used to insert thermocouples to monitor the coolant channel temperatures. The core has a cylindrical configuration of six rings (A, B, C, D, E and F) having 1, 6, 12, 18, 24 and 30 locations respectively. These 91 positions are able to host either fuel rods or other components like control rods, a neutron source, graphite dummies (mobile reflector), irradiating and measurement channels (e.g. central thimble or A ring). Each location corresponds to a role in the aluminum upper grid plate of the reactor core. The core is surrounded by an annular graphite reflector and water. Inside the reflector there is a rotary specimen rack with 40 positions for placement of samples to be activated by neutron flux. The top view of the reactor core and the rotary specimen rack are presented in Fig. 5. There is a very high number of reactor loading configurations, so that it is possible to obtain the sub-critical level required simply loading/unloading fuel rods from the core.

heat is transferred from the primary to the secondary loop. The secondary loop water is cooled in an external cooling tower. Figure 3 shows the forced cooling system, which transfers the heat generated in the reactor core to a water-to-water heat exchanger. The secondary cooling system transfers the reactor core heat from the heat exchanger to a cooling tower. In the diagram is shown also the instrumentation distribution and the forced

A simplified view of the IPR R1 TRIGA core configuration is shown in the Fig. 4. As shown in the diagram there are small holes in the core upper grid plate. These holes were used to insert thermocouples to monitor the coolant channel temperatures. The core has a cylindrical configuration of six rings (A, B, C, D, E and F) having 1, 6, 12, 18, 24 and 30 locations respectively. These 91 positions are able to host either fuel rods or other components like control rods, a neutron source, graphite dummies (mobile reflector), irradiating and measurement channels (e.g. central thimble or A ring). Each location corresponds to a role in the aluminum upper grid plate of the reactor core. The core is surrounded by an annular graphite reflector and water. Inside the reflector there is a rotary specimen rack with 40 positions for placement of samples to be activated by neutron flux. The top view of the reactor core and the rotary specimen rack are presented in Fig. 5. There is a very high number of reactor loading configurations, so that it is possible to obtain the

sub-critical level required simply loading/unloading fuel rods from the core.

and natural circulation paths in the pool.

Fig. 2. IPR-R1 TRIGA reactor pool and core

Fig. 3. IPR-R1 TRIGA reactor cooling system and instrumentation distribution

Fig. 4. Simplified core diagram

The prototypical cylindrical fuel elements are a homogeneous alloy of zirconium hydride (neutron moderator) and uranium enriched at 20% in 235U. The reactor core has 58

Experimental Investigation of Thermal Hydraulics in the IPR-R1 TRIGA Nuclear Reactor 7

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

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

**3. Methodology** 

**3.1 Fuel and core coolant channel temperatures** 

once a second in a historical database (Mesquita & Souza, 2010).

instrumented fuel element and one thermocouple inside a core channel.

Fig. 6. Diagram of the instrumented fuel element

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
