An experimental testing of coolant flow rate and velocity in the core of Nigeria Research Reactor-1

Transcription

1 Int. J. Nuclear Energy Science and Technology, Vol. 9, No. 2, An experimental testing of coolant flow rate and velocity in the core of Nigeria Research Reactor-1 S.A. Agbo* Physics Department, Ahmadu Bello University, Zaria, Nigeria *Corresponding author Y.A. Ahmed and I.O.B. Ewa Centre for Energy Research and Training, Ahmadu Bello University, Zaria, Nigeria M. Abubakar Physics Department, Ahmadu Bello University, Zaria, Nigeria M.S. Anas CAAS Mando, Division of Agricultural Colleges, Ahmadu Bello University, Zaria, Nigeria Abstract: This paper describes experiments to understand the behaviour of nuclear reactors thermo-hydraulic parameters allow improved model predictions, contributing to their safety. The Nigeria Miniature Neutron Source Reactor called Nigeria Research Reactor-1 (NIRR-1) is a tank-in-pool type reactor with 90.2% enriched uranium as fuel, and light water as moderator and coolant. The core is cooled by light water natural convection. The reactor core assembly, surrounded by beryllium reflectors, is located at the bottom of the reactor vessel. The NIRR-1 has no installed device to measure the core mass flow rate. In this work, experiments were performed at different power levels to monitor some thermo-hydraulic parameters like core mass flow rate, coolant velocity, mass flux, density and Reynolds number in the core of NIRR-1. These experiments confirm the efficiency of natural circulation in removing the heat produced in the reactor core by nuclear fission. Copyright 2015 Inderscience Enterprises Ltd.

2 172 S.A. Agbo et al. Keywords: NIRR-1; reactor; core; flow rate; thermo-hydraulic. Reference to this paper should be made as follows: Agbo, S.A., Ahmed, Y.A., Ewa, I.O.B., Abubakar, M. and Anas, M.S. (2015) An experimental testing of coolant flow rate and velocity in the core of Nigeria Research Reactor-1, Int. J. Nuclear Energy Science and Technology, Vol. 9, No. 2, pp Biographical notes: S.A. Agbo obtained a Master of Science degree in Nuclear Physics from Ahmadu Bello University, Zaria, Nigeria, in 2015 and a Bachelor of Science degree in Physics from Ahmadu Bello University, Zaria, Nigeria, in He has teaching experience of ten years and research experience of three years. His research interest includes thermo-fluid dynamics, flux measurements, nuclear instruments calibration and reactor physics. Y.A. Ahmed is Senior Research Fellow/Senior Lecturer at Centre for Energy Research and Training, Ahmadu Bello University, Zaria. He was awarded PhD degree in Physics (Nuclear Analytical Techniques) from Ahmadu Bello University, Zaria, Nigeria in He has published over 50 papers in reputed national and international journals. His research interest includes flux measurements, characterisation and dynamic behaviour of neutron sources, nuclear instruments calibration and radiation measurements, analytical techniques for food and nutritional studies, analytical techniques for geological and environmental studies, nuclear structure and decay data evaluation. I.O.B. Ewa is the former Minister of Science and Technology, Federal Republic of Nigeria and Professor at Centre for Energy Research and Training, Ahmadu Bello University, Zaria, Nigeria. He was awarded PhD degree in Physics from Ahmadu Bello University, Zaria, Nigeria in He has published over 100 papers in reputed national and international journals. M. Abubakar is a postgraduate student at Ahmadu Bello University, Zaria, Nigeria. M.S. Anas obtained a Master of Science degree in Nuclear Physics from Ahmadu Bello University, Zaria, Nigeria, in He is a Lecturer at Division of Agricultural Colleges, Ahmadu Bello University, Zaria. He has research experience of three years. 1 Introduction Mesquita et al. (2011) have shown that understanding the behaviour of the operational parameters of nuclear reactors allows the development of improved analytical models to predict the fuel temperature, so contributing to their safety. It has also been established by Aksan (2010) that in all light water reactors (LWRs), natural circulation is an important passive heat removal system. The importance of studies and experiments on heat removal by natural convection to remove heat from the residual remaining after the shutdown was established following the natural disaster at the Fukushima nuclear power plant.

3 An experimental testing of coolant flow rate and velocity 173 Nigeria Research Reactor-1 (NIRR-1) is a low-power, tank-in-pool reactor (schematic diagram and sectional view shown in Figures 1 and 2 respectively) with nominal thermal power of 30 kw under steady state condition (Jonah et al., 2012). It is designed by China Institute of Atomic Energy (CIAE) (Yongmao, 1985). The reactor s first criticality was accomplished on 3rd February 2004 and has been working safely for Neutron Activation Analysis (NAA) and limited radioisotope production (Balogun et al., 2005; Jonah et al., 2005; Jonah et al., 2006). The current core of the reactor is a mm square cylinder and fuelled by U-Al 4 enriched to 90% in Al-alloy cladding (Jonah et al., 2005). It has a total number of 347 fuel pins and three Al dummies in the fuel lattice. The length of the fuel element is 248 mm, the active length being 230 mm with 9 mm Al-alloy plug at each end. The diameter of the fuel meat is 4.3 mm and the 235 U loading in each fuel element is about 2.88 grams. The cladding is Al-alloy, whose thickness is 0.6 mm (Jonah et al., 2006). NIRR-1 has only one central control rod with an active length of 230 mm serving as shim rod, regulation rod and safety rod. The functions of reactor start-up, steady-state operation, and shutdown are achieved by moving the control rod which is made up of a Cd absorber (Jonah et al., 2006). NIRR-1 is operated using the control console, the microcomputer control system and the two rabbit systems all connected to a power source and during normal operation the reactor water temperature varies between 23 C and 46 C. However, the temperature difference rises rapidly and attains a stable value due to insufficient natural circulation phenomenon. The most important component of NIRR-1 facility is the reactor core, which is located at the bottom of the lower section of the reactor vessel (CERT, 2005). The reactor core consists of the fuel cage ( birdcage ), the control rod guide tube, fuel elements, dummy rods and four tie rods. The core is located 4.7 m under water close to the bottom of the water light reactor vessel (see Figure 1). The quantity of water is 1.5 m 3 in the vessel, which serves the purpose of radiation shielding, moderation and as primary heat transfer medium (CERT, 2005). The water-filled reactor vessel is in turn immersed in a water-filled pool 30 m 3. The cold water is drawn through the inlet orifice. The water flow past the hot fuel elements and comes out through the core outlet orifice. The hot water rises to mix with the large volume of water in the reactor vessel and to the cooling coil. Heat passes through the walls of the container to the pool water. The thermal-hydraulic design of NIRR-1, intended as with no reactivity insertion, is closely dependent on the structural design of the reactor. The core is cooled by natural convection which is established through the heat generated by fission occurring in the core. The reactor coolant is drawn through the inlet orifice by natural convection flow through the channels within the fuel elements. The coolant moves up through the core and exits through the core outlet orifice at temperature T 2 to the upper part of the tank where the temperature is T 3. The coolant inside the core passes through the aperture surrounding the upper ends of the fuel. New colder coolant is substituting the hotter one in the core causing the coolant in the downcomer (with temperature T 1 ) to move downwards and maintain its temperature so that the coolant enters the reactor core at temperature T 1. In the simplified model, the velocity of the coolant in the downcomer is supposed to be equal to the velocity of the coolant in the core. Heat transfer from core to the reflector is by conduction and convection and also heat transfer from the tank to the pool (T 0 ) across the wall of the tank is again by conduction and convection mechanisms. A diagrammatic representation of this heat transfer mechanism described above is represented in Figure 3 (CERT, 2005).

4 174 S.A. Agbo et al. Figure 1 Schematic diagram of NIRR-1 Figure 2 Sectional view of NIRR-1

5 An experimental testing of coolant flow rate and velocity 175 Figure 3 Schematic diagram of the coolant flow pattern of NIRR-1 CERT (2005) pointed out that the decay heat in NIRR-1 is removed by natural circulation of the reactor vessel water in a manner similar to that during operation. It has also been established in CERT (2005) that a number of thermal hydraulic tests and calculations, especially from the dynamic experiments, have shown that the natural circulation of the prototype Miniature Neutron Source Reactor (MNSR) (which is comparable to NIRR-1) has the following characteristics: 1 Negative feedback effect: This occurs when the temperature difference between the inlet and outlet coolant of MNSR increases; the floating force and circulating head will increase to make the flow velocity high, which in turn will limit the rise in temperature. 2 Insufficient circulation: Because of the small size of the core, the distance from the inlet orifice to the outlet orifice is small. The water, after being heated in the core goes out through the upper part of the core. Part of the water does not get sufficiently cooled before it sinks down, resulting in part of outlet water being carried back into the core due to siphoning effect. This direct re-circulation of the part of hot water causes a rise of the inlet water temperature. This phenomenon is called insufficient circulation, which speeds up the rise of the coolant temperature in the core and shortens the function time of the temperature effect. It is thus not possible to cause the inlet water temperature to rise in such a short time by heating the core only, but by the coupling action between the inlet and the outlet coolant. Consequently this offers some benefit to the reactor safety.

6 176 S.A. Agbo et al. Heat removal from NIRR-1 is accomplished by natural convective flow of coolant during steady state operation (CERT, 2005). Since measurements of some of the steady fluid flow variables such as velocity and mass flow rate are not amenable to direct measurements, an indirect means was employed in this study to monitor these thermohydraulic parameters. The research of core natural circulation ability has great meaning on reducing core degradation risk, as such these experiments confirm the efficiency of natural circulation in removing the heat produced in the reactor core by nuclear fission. The data taken during these experiments provide an excellent picture of the thermal performance of the NIRR-1 reactor core. 2 Mathematic physics model Core mass flow rate m (in kg/s) is given indirectly from the heat balance across the core using measurements of the water inlet and outlet temperatures (Mesquita and Rezende, 2010; Mesquita et al., 2007; Mesquita et al., 2009; Mesquita et al., 2011): q m (1) TC p where q is the power supplied to the core (kw), m is the core mass flow rate of water (kg/s), C p is the specific heat of the coolant passing through the core and ΔT is the difference between the inlet and outlet temperatures (in C). The mass flux G is given by (Mesquita and Rezende, 2010; Mesquita et al., 2011): m G (2) A where A is core area (0.249 m 2 for NIRR-1), G is the mass flux (kg/m 2 s) and m is the core mass flow rate of water (kg/s). The velocity of the coolant passing through the core is obtained by (Mesquita and Rezende, 2010; Mesquita et al., 2011): G v (3) where ρ is the coolant density. The equation that relates the coolant temperature and coolant density of NIRR-1 is given by (Mansir et al., 2012): y E X X X (4) where y = coolant density, X = coolant temperature. Reynolds number (R e ) used to characterise the flow regime is given by: R GD h e (5) where G is the mass flux (kg/m 2 s), D h is the hydraulic diameter (0.23 m for NIRR-1) and μ is the dynamic viscosity.

7 An experimental testing of coolant flow rate and velocity Experimental facility NIRR-1 description Prior to the starting of these experiments, the thermal power released by the core of NIRR-1 was calibrated by heat balance and calorimetric methods (Agbo, 2015). The heat balance methodology for monitoring the core mass flow rate consists of the measurement of the inlet temperature, outlet temperature and coolant temperature difference. In order to cover the whole range of the reactor, these measurements were performed at six different power levels: 3.6 kw, 6 kw, half power (15 kw), 18 kw, 27 kw and full power (30 kw) respectively. All measurements were done at 20 minutes interval. High precision temperature detectors (thermocouples) were used to measure the inlet and outlet temperature. One of them was positioned at the outside of the side beryllium annulus near the core inlet orifice to measure the inlet temperature. The other was positioned at the upper part of the side beryllium annulus near the core outlet orifice to measure the outlet temperature. The combination of these two pairs of thermocouples monitored the temperature difference of the reactor coolant. The locations of the thermocouples are shown in Figure 4. The thermocouples were calibrated to obtain measurements within the experimental resolution of ±0.5 C. Figure 4 A layout core configuration showing the various components A miniature fission chamber, made using stainless steel walls and electrodes, with the operating voltage that varies from about 50 to 300 V was employed as neutron flux detector. Walls of the chambers are lined with highly enriched uranium to enhance the

8 178 S.A. Agbo et al. ionisation current with argon as a common choice for the chamber fill gas used at a pressure of several atmospheres. There are two small vertical holes each 10 mm in diameter and 190 mm deep on the side annular beryllium reflector on the same circle as the inner irradiation sites (i.e. on a circle of radius 165 mm and at angles of 144 to each other). A current-type miniature fission chamber (LB 1120) is in each hole to monitor neutron flux at each irradiation site and provide control signals. The locations of the neutron flux detectors are shown in Figure 4. The sensitivity, linearity, follow-up features and lifetime of the selected miniature fission chambers are adequate to meet the requirements of NIRR-1 control. 4 Results and discussions At a neutron flux of n/cm 2 s corresponding to 3.6 kw, the reactor operated for 4 h and during this period, the inlet temperature, outlet temperature and coolant temperature difference were recorded for every 20 min. The average values of the inlet temperature, outlet temperature, coolant temperature difference, core mass flow rate, coolant density, mass flux, coolant velocity and Reynolds number obtained at this power level were C, C, C, kg/s, kg/m 3, kg/m 2 s, m/s and 105 respectively. The high values recorded for the inlet temperatures, outlet temperatures and coolant temperature difference (as presented in Table 1) at this low power level were because the reactor was operated for some hours for the purpose of sample irradiation before the of the experiment begins at this power level. Table 1 Time (h) Data obtained from experiment performed at 3.6 kw T inlet T outlet ΔT ρ (kg/m 3 ) m (kg/s) G (kg/m 2 s) v (10 4 m/s) 11: : : : : : : : : : : : : At a flux of n/cm 2 s corresponding to 6 kw, the reactor operated for over 5 h. The inlet temperature, outlet temperate and coolant temperature difference were recorded for every 20 min. The results as presented in Table 2 show that the coolant temperature difference ΔT recorded is approximately steady with an average value of 7.29 C. A real steady state is reached after some hours of operation. As can be seen in Table 2, the R e

9 An experimental testing of coolant flow rate and velocity 179 measured temperatures were observed to be stable for over 1 h (from 14:30 h to 15:50 h). The average inlet and outlet temperatures were C and C respectively. The value of the average core mass flow rate obtained at this power level was kg/s. The values of the average coolant density, mass flux coolant velocity and Reynolds number obtained at this power level were kg/m 3, kg/m 2 s, m/s and 294 respectively. Table 2 Time (h) Data obtained from experiment performed at 6 kw T inlet T outlet ΔT ρ (kg/m 3 ) m (kg/s) G (kg/m 2 s) v (10 4 m/s) 10: : : : : : : : : : : : : : : : : For experiment carried out at half power, the reactor was critical at 15 kw as indicated on the control console and microcomputer system. The reactor operated during a period of about 5 h. The inlet temperature and outlet temperature were measured and recorded at 20 min interval and the coolant temperature difference for each of the interval was obtained. The coolant temperature recorded during the operation was found to be steady at approximately C for the whole period of operation at this power level (as indicated in Table 3). A real steady state was reached after about 4 h of operation. The measured temperatures were observed to be stable for over 40 min. The measurements of punctual temperatures and time were done with considerable accuracy and precision. The value of the average core flow rate obtained kg/s compare well with the value obtained by Jonah et al. (2012) using PLTEMP/ANL code version 4.1. The values of the average inlet temperature, outlet temperature, coolant density, mass flux coolant velocity and Reynolds number obtained at this power level were C, C, kg/m 3, kg/m 2 s, m/s and 431 respectively. R e

10 180 S.A. Agbo et al. Table 3 Data obtained from experiment performed at half power (15 kw) Time (h) T inlet T outlet ΔT ρ (kg/m 3 ) m (kg/s) G (kg/m 2 s) v (10 3 m/s) 11: : : : : : : : : : : : : : : : At a flux of n/cm 2 s corresponding to 18 kw, the reactor was operated for 4 h. The inlet temperature and outlet temperature were measured and recorded at 20 min interval and the temperature difference corresponding to the inlet and outlet temperature at each of this interval was obtained (as indicated in Table 4). The average value of the: inlet temperature, outlet temperature, coolant temperature difference, core mass flow rate, coolant density, mass flux, coolant velocity and Reynolds number obtained at this power level were C, C, C, kg/s, kg/m 3, kg/m 2 s, m/s and 510 respectively. Table 4 Time (h) Data obtained from experiment performed at 18 kw T inlet T outlet ΔT ρ (kg/m 3 ) m (kg/s) G (kg/m 2 s) v (10 3 m/s) 10: : : : : : : : : : R e R e

11 An experimental testing of coolant flow rate and velocity 181 For experiment carried out at a flux of n/cm 2 s, the reactor was critical at 27 kw indicated on the control console and microcomputer system. The reactor operated for over 5 h, during this period, the inlet and outlet temperatures as well as the coolant temperature difference were measured and recorded at 20 min. As can be seen in Table 5, the measured temperatures were observed to be stable from 16:20 h to 17:00 h. The average value of the inlet temperature, outlet temperature, coolant temperature difference, core mass flow rate, coolant density, mass flux coolant velocity and Reynolds number obtained at this power level were C, C, 17 C, kg/s, kg/m 3, kg/m 2 s, m/s and 558 respectively. The reactor neutron flux was preset at its licensed full power value of ncm 2 s 1 and was operated for over 3 h. The inlet temperature and outlet temperature were measured and recorded at 20 min interval and the temperature difference corresponding to the inlet and outlet temperatures at each interval was obtained (as indicated in Table 6). Steady state was reached after 2 h of operation. The temperatures were observed to approach a constant value from 13:10 h to 13:50 h. The average value of the inlet temperature, outlet temperature, coolant temperature difference, core mass flow rate coolant density, mass flux coolant velocity and Reynolds number obtained at this power level were C, C, C, kg/s, kg/m 3, kg/m 2 s, m/s and 576 respectively. Table 5 Data obtained from experiment performed at 27 kw Time (h) T inlet T outlet ΔT ρ (kg/m 3 ) m (kg/s) G (kg/m 2 s) v (10 3 m/s) 11: : : : : : : : : : : : : : : : : : R e

12 182 S.A. Agbo et al. Table 6 Data obtained from experiment performed at full power (30 kw) Time (h) T inlet T outlet ΔT ρ (kg/m 3 ) m (kg/s) G (kg/m 2 s) v (10 3 m/s) 10: : : : : : : : : : : : R e Figure 5 Behaviour between core mass flow rate versus with power (see online version for colours) From the results presented in Tables 1 6 and Figures 5 7, it can be seen that the core mass flow rate, velocity, mass flux and Reynolds number increase with power. From Reynolds numbers (105, 294, 431, 510, 596, and 576) obtained at 3.6 kw, 6 kw, 15 kw, 18 kw, 27 kw and 30 kw respectively, it can be said that the coolant flow across the core is laminar and this is in agreement with the thermal and hydraulic design of NIRR-1 as reported in the CERT (2005). The coolant flow can transit from laminar to turbulent when onset nucleate boiling occurs, but this is not the case in these experiments. Fission energy produced in the core was removed by natural circulation during the test period and local boiling at the surface of fuel element did not occur. As such the flow rate is expected to be laminar. The inlet and outlet temperatures for all the experiments were

13 An experimental testing of coolant flow rate and velocity 183 observed at some points to increase with time. This is due to the compact nature of the core, which was designed to cause insufficient thermal circulation of coolant in the core. The phenomenon (insufficient circulation) speeds up the rise of the coolant temperature in the core and shortens the function time of the temperature effect. It is thus not possible to cause the inlet water temperature to increase in such a short time by heating the core only, but by the coupling action between the inlet and the outlet coolant. Consequently this offers some benefit to the reactor safety. The specific heat capacity (C p ) values were corrected as a function of the coolant temperature (Miller, 1989). The measurements of punctual temperatures and times for all the experiments were done with considerable accuracy and precision. Figure 6 Behaviour between mass flux versus power (see online version for colours) Figure 7 Behaviour between coolant velocity versus power (see online version for colours)

14 184 S.A. Agbo et al. 5 Conclusion The NIRR-1 has no installed device to measure the core flow rate. In this work, experiments were performed at different power levels to monitor some thermo-hydraulic parameters like core mass flow rate, coolant velocity, coolant density, mass flux and Reynolds number in the core of NIRR-1. The mass flow rate through the core for each of the experiments was determined indirectly from the heat balance across the core using measurements of water inlet and outlet temperatures. In order to cover the whole range of the reactor, the experiments were performed at six different power levels: (3.6 kw), (6 kw), half power (15 kw), (18 kw), (27 kw) and full (30 kw). The thermo-hydraulic parameters for each of the power levels were evaluated. The values of the Reynolds numbers obtained for the experiments carried out show that the coolant flow is laminar. From the results obtained from the experiments, it was observed that the thermohydraulic parameters of NIRR-1 increase with the core power which is in agreement with the reactor design as reported in the Safety Analysis Report (CERT, 2005). The values of the core mass flow rate obtained in this experiment compare well with the values obtained by Jonah et al. (2012) using PLTEMP/ANL code version 4.1. The inlet and outlet temperatures for all the experiments were observed at some points to increase with time. This is due to the compact nature of the core, which was designed to cause insufficient thermal circulation of coolant in the core. References Agbo, S.A. (2015) Thermal Power Calibration of Nigeria Research Reactor-1 (NIRR-1) by Calorimetric and Heat Balance Methods, MSc Thesis, Ahmadu Bello University, Zaria, Nigeria. Aksan, N. (2010) IAEA Training Course on Natural Circulation in Water-Cooled Nuclear Power Plants, International Centre for Theoretical Physics, Trieste, Italy, Paper ID T21b. Balogun, G.I., Jonah, S.A. and Umar, I.M. (2005) Measures aimed at enhancing safe operation of the Nigeria Research Reactor-1 (NIRR-1), International Conference on Operational Safety Performance in Nuclear Installations, 30 November 2 December, IAEA, Vienna Austria, pp Centre for Energy Research and Training (CERT) (2005) Final Safety Analysis Report of Nigeria Research Reactor-1 (NIRR-1/ SAR), CERT Technical Report-CERT/NIRR-1/FSAR-01. Jonah, S.A., Balogun, G.I., Umar, I.M. and Mayaki, M.C. (2005) Neutron spectrum parameters in irradiation channels of the Nigeria Research Reactor-1 (NIRR-1) for the k 0 -NAA standardization, Journal of Radioanalytical and Nuclear Chemistry, Vol. 266, No. 1, pp Jonah, S.A., Ibrahim, Y.V., Kalimulla, M. and Matos, J.E. (2012) Steady state thermal hydraulic operational parameters and safety margins of NIRR-1 with LEU fuel using PLTEMP-ANL CODE, European Nuclear Conference (ENC), 9 12 December, Manchester, UK. Jonah, S.A., Umar, I.M., Oladipo, M.O.A., Balogun, G.I. and Adeyemo, D.J. (2006) Standardization of NIRR-1 irradiation and counting facilities for instrumental neutron activation analysis, Applied Radiation and Isotopes, Vol. 64, No. 7, pp Mansir, I.B., Pam, G.Y. and Folayan, C.O. (2012) Effects of density variation on the cooling of the Nigeria Research Reactor-1 (NIRR-1), Advances in Applied Science Research, Vol. 3, No. 1, pp Mesquita, A.Z. and Rezende, H.C. (2010) Monitoring of coolant flow rate and velocity in the hot channel of the IPR-R1 TRIGA Reactor core, Científica, Vol. 14, No. 2, pp

15 An experimental testing of coolant flow rate and velocity 185 Mesquita, A.Z. and Rezende, H.C. (2010) Thermal methods for on-line power monitoring of the IPR-R1 TRIGA Reactor, Progress in Nuclear Energy, Vol. 52, pp Mesquita, A.Z., Rezende, H.C. and Souza, R.M.G.P. (2011) Thermal power calibrations of the IPR-R1 TRIGA Reactor by the calorimetric and the heat balance methods, Progress in Nuclear Energy, Vol. 53, pp Mesquita, A.Z., Rezende, H.C. and Tambourgi, E.B. (2007) Power calibration of the TRIGA Mark I Nuclear Research Reactor, Journal of the Brazilian Society of Mechanical Sciences and Engineering, Vol. XXIX, No. 3, pp Miller, R.W. (1989) Flow Measurement Engineering Handbook, 2nd ed., McGraw-Hill Publishing Company, New York, pp.e19 E21. Yongmao, Z. (1985) IAEA-TECDOC-384 Technology and Use of Low Power Research Reactors, Report of IAEA Consultants Meeting, 30 April 3 May Beijing, China.