Reactivity Power and Temperature Coefficients Determination of the TRR

Transcription

1 Reactivity and Temperature Coefficients Determination of the TRR ABSTRACT Ahmad Lashkari Nuclear Science and Technology Research Institute (NSTRI), Atomic Energy Organization of Iran Tehran , Iran The aim of this paper is to present the experimental results of the power and temperature coefficient of reactivity of the Tehran Research Reactor (TRR) at the Nuclear Science and Technology Research Institute (NSTRI) of Iran. In this work in addition to the previous method, new methods were used to measure the reactivity coefficient of TRR. The experiments were performed in the TRR reactor with 33 MTR fuel elements in the core. At the first method, we determined the isothermal coefficient of TRR and then calculated power and temperature coefficient of reactivity. This method is very similar to the method that is used to determine the power coefficient of IPR-R1 TRIGA reactor. One of the new methods used in this study is comparing the situation of control rod positions in two cooling modes (natural and force) at the same power of TRR. The difference between two control rod configurations is caused by the temperature difference in coolant in two modes. With measuring the reactivity difference and coolant temperature, we can calculate reactivity coefficient. The last new method is much more efficient than the above methods, using the dynamic behaviour of reactor power due to change of reactor core temperature. The main advantage of this method is that we can measure the reactivity coefficient of reactor very fast and independent of the control rods worth and positions. The average values of the temperature and power reactivity coefficient of the fuel and the coolant in TRR are: α_t (F)=1.95 / C, α_t (m)=13.57 / O C α_p (F)=.16 /kw, α_p (m)=.89 /kw 1 INTRODUCTION The portion of reactivity change arising from the effect of energy production is called reactivity feedback, which includes temperature and void coefficient of reactivity. Temperature coefficients of reactivity due to fuel, coolant and moderator component of a reactor core are defined as the change in reactivity per unit change in average temperature of that component. If αt,j represents the temperature reactivity coefficient of a component j then they can be written as: α T, j ρ (/ C) (1) T j k eff 1 5 ρ 1 () (2) k eff 416.1

2 416.2 Where keff is effective multiplication factor corresponding to average temperature T of the core component j. This paper reports the results of a set of experiments to determine the power and temperature coefficient of reactivity of the Tehran Research Reactor (TRR). For calculation of these parameters the values of isothermal temperature coefficient are needed. The isothermal temperature coefficient was measured by observing the reactivity change with core temperature. In this work in addition to the previous method, new methods were used to measure the reactivity coefficient of TRR. The experiments were performed in the TRR reactor with 33 MTR fuel elements in the core. At the first method, we determined the isothermal coefficient of TRR and then calculated power and temperature coefficient of reactivity. This method is very similar to the method that is used to determine the power coefficient of IPR-R1 TRIGA reactor [1]. One of the new methods used in this study is comparing the situation of control rod positions in two cooling modes (natural and force) at the same power of TRR. The difference between two control rod configurations is caused by the temperature difference in coolant in two modes. With measuring the difference reactivity and coolant temperature, we can calculate reactivity coefficient. The last new method that is much more efficient than the above methods is using the dynamic behavior of reactor power due to change of reactor core temperature. The main advantage of this method is that we can measure the reactivity coefficient of reactor very fast and independent of the control rods worth and positions. 2 METHODOLOGY One of the operating conditions that effect on the reactivity of a reactor core is temperature. Changes in temperature will cause changes in reactivity. The direction of the changes and its magnitude are great importance in the reactor safety and control. If the temperature change is uniform throughout the core, as would be in a homogeneous reactor, the temperature effect on the reactivity can be expressed only by a simple temperature coefficient, αiso, defined as the change in reactivity per degree change in temperature [2]: α ρ (3) Where and ΔT are the changes in reactivity and temperature, respectively. The negative reactivity feedback,, produced by a temperature increase ΔT is then: ISO T Assuming that ISO is constant over the range of temperature T. This ISO is sometimes called the isothermal temperature coefficient or the zero-power temperature coefficient [2]. Heterogeneous reactor changes in temperature during operation are not uniform, that is, they are not the same in the moderator as in the fuel. In such a reactor we have to distinguish between the reactivity arising in the cooling, or moderator, and that arising in the fuel, and, accordingly, define a coolant temperature coefficient, T(M), and a fuel temperature coefficient, T(F). These coefficients in general are different in magnitude and in response time. Effects on the fuel, for instance, resonance absorption (Doppler Effect) or thermal distortion of fuel elements are regarded as prompt, while effects on the moderator or coolant are delayed. The power coefficient of reactivity is defined as α ρ (4) Where P is the change in power. Proceedings of the International Conference Nuclear Energy for New Europe, Portorož, Slovenia, September 14 17, 215

3 416.3 To obtain the contribution of the fuel to P and thus the fuel power coefficient, P(F), in a reactor, we have to subtract T(M), the effect arising in the moderator due to the change in moderator temperature, T(M). An approximate value for T(M) would be: T(M)= αiso ΔT(M) (5) Then, the fuel and moderator power coefficient of reactivity are given by T(F)= - T((M) (6) α F ρ α M ρ (7) (8) In this work, the temperature of coolant was measured easily but we haven t any facility to measure the temperature of the fuel plate. In this research we used CONVECT code to determine the fuel plat temperatures. CONVECT is a steady state code used to analyses of natural convection of MTR type reactors [3]. With measuring the coolant temperature and calculating the fuel temperature, the temperature coefficients are defined as: α T ρ α T ρ (9) (1) Temperature coefficients are the main safety parameters that are used in two natural and force cooling system. 2.1 The Tehran Research Reactor The TRR is a pool type research reactor, in which light water serves as coolant, radiological shielding as well as neutron moderating medium and reflector. The reactor is designed and licensed to operate at a maximum thermal power level of 5 MW. The reactor core assembly is located in a two-section pool and may be operated in either pool. One of the sections contains experimental facilities, like beam tubes, rabbit system, and thermal column. The other section is an open area for bulk irradiation studies. The major components of TRR are the pool (including embedment and accessories), bridge and support structure, core, cooling system, control and instrumentation, ventilation system, and the experimental facilities. Elements of the reactor core are arranged in a 9 by 6 grid plate structure. Details of reactor description and core parameters are given in TRR- Safety Analysis Reports (SAR). 3 RESULT AND DISCUSSION 3.1 Isothermal Temperature Coefficient Measurement In these experiments, the reactor was shut down for a one week, so the reactor was xenon poisoning free with low background signal. Reactor was critical at 1kW and the positions of SRs were written. The inlet and outlet temperature of the coolant was measured by thermocouple. In 1kW the difference between inlet and outlet was about 1.5 C. The power of reactor was increased by positive insertion by RR and set power to 2, 5, 8 and 1 kw. In Proceedings of the International Conference Nuclear Energy for New Europe, Portorož, Slovenia, September 14 17, 215

4 416.4 (kw) each power the rod positions and the inlet and outlet coolant temperature were measured. Table 1 shows all data information at all power, only the position of RR changed due to temperature increment. The average temperature of coolant is the average of inlet and outlet coolant temperature and easily measured by two thermocouple. Also the CONVECT code was used to calculate the fuel and coolant temperature in each scenarios. The results show a good agreement between experimental and simulation results of coolant temperatures. Because we have not any facility to measure the fuel temperature, we have to rely on calculation results. Calculations show that the temperature of the fuel and the coolant are the same in low powers and we can show the both temperature with a one temperature that is named isothermal temperature. After 2 kw the difference between the coolant and fuel temperature increased. So we can calculate the isothermal temperature coefficient at 2 kw with a good approximate. Table 1: Results of isothermal temperature coefficient of reactivity Tin Tout CONVECT Tave(F) TPool Rods position SRs RR () / At the first and before doing any experiments the RR was calibrated and the worth of RR determined. Fig.1 shows the RR integral worth of reactivity. The ratio of the reactivity changes to the average coolant temperature at the 2 KW is about (/ O C) and named isothermal temperature coefficient. The ratio of the reactivity changes to the coolant temperature at higher powers increases. The reason of this increment is that at the higher powers the average temperature of the fuel plate is higher than the average temperature of coolant. For example at the power 1 kw the fuel average temperature is higher than coolant about 2 C. The incremental trend of this parameter at higher power shows the accuracy of method. /kw Reactivity () Rod position ( %) Figure1: RR reactivity worth verse rod position. Proceedings of the International Conference Nuclear Energy for New Europe, Portorož, Slovenia, September 14 17, 215

5 416.5 At the next power with measuring the average coolant temperature and total reactivity changes, we can calculate the reactivity changes due to fuel temperature increment. Column 8 of table 2 shows the reactivity changes due to the fuel temperature. In the last two columns of the table 2 power reactivity coefficients of the fuel and coolant were reported for 3 powers 5, 8 and 1 kw. The average of these coefficients is about.15 and.75 (/kw) respectively. In this paper only the power reactivity coefficient report experimentally. In the next step we used CONVECT code to calculate Tf and Tm at each powers. The difference between Tf and Tm at three powers 5, 8 and 1 kw were calculated and shown in the table 2. The ratio of (F)/ ΔT (F) gives the temperature reactivity of the fuel. The average value for three powers 5, 8 and 1 KW is 1.95 (/ C). Also is easily calculated for three powers and the average value of this parameter is (/ C). The last four columns of table 2 shows the power and temperature reactivity coefficients. Table 2: Temperature and power coefficient of reactivity for the power intervals measured (KW) Unit CONVECT C Tave(F). C C / / C (m) (F) T(F)= TF-Tm C α F / C / C Comparing the Control Rods Position in Two Natural and Force Cooling System Modes In this method the control rods position was compared in two natural and force cooling modes at 8 kw. The reactivity difference between two modes is distributed to the fuel and moderator temperature. Table 3 shows all temperature measurements and control rods position. According to CONVECT calculation in the natural circulation mod, the temperature of the fuel and the coolant are the same with a good approximation. With using α F obtained from the previous method, we calculated the value of with this new method. As can be seen, the new value is (14.2 / C) a little higher than the previous value. Table 3: Coolant temperature coefficient of reactivity for the power 8 kw. (KW) 8 F 8 N T in C T out C CONVECT T ave (m) C T ave (F) C T ave (m) C T Pool C Rods position SRs RR () 83.3 ΔT(m) C 5 ΔT(F) C 6.5 (F) Measurement of Reactivity Coefficient According to Inhour Equation (m) 68 α F Pcm/ kw / C In this method a new technique was used to measure the temperature reactivity coefficient. Increasing the temperature of the reactor core causes applying negative reactivity feedback and vice versa decreasing the temperature makes positive reactivity. In this method, reactor power was set at specific power in natural cooling system. The inlet and outlet temperature of coolant were measured with thermocouple. With changing the cooling system mode from natural to force, a positive reactivity inserted to the reactor and the power of the reactor was beginning to increase. With measuring doubling time and using the inhour equation, the reactivity worth of temperature changing in the reactor component is calculated directly 14.2 Pcm/ kw Proceedings of the International Conference Nuclear Energy for New Europe, Portorož, Slovenia, September 14 17, 215

6 416.6 from equation The advantage of this method is the measuring of the reactivity worth without control rods dependence. (11) (12) Where: : Reactivity : delayed neutron fraction for group i : Prompt neutron lifetime (s) : Reactor period / T1/2: doubling time : Decay constant of delayed neutron group i In this work one group delayed neutron approximation was used in inhour equation. In table 4 the results of this method are shown only for 6 kw power. Doubling time was measured about 12.47s and the reactivity was calculated about 68 (). Similar to the previous methods, the average temperature of the fuel was calculated by CONVECT code. With using the value of α F the value of was obtained from the new method (12 / C). Table 4: Coolant temperature coefficient of reactivity for the power 6 kw. (kw) Tin Tout Cal. Tave(F). TPool T1/2 ()) ΔT(m) ΔT(F) (F) (m) / C CONCLUSIONS The experiments were performed in the TRR reactor with 33 fuel elements in the core. At first, it was determined the isothermal coefficient of (/ C). As it was shown, most of the negative reactivity change with increasing power must be attributed to the change in the coolant temperature. The coefficient due to the fuel heating was very small. Then, we can conclude that the power coefficient of reactivity of the coolant is the main contributor to the power coefficient of reactivity in TRR. The aim of this paper is to present the experimental results of the power and temperature coefficient of reactivity. In this work in addition to the previous method, two new methods were used to measure the reactivity coefficient of TRR. At the first method, we determined the isothermal coefficient of TRR and then calculated power and temperature coefficient of reactivity. The first new method used in this study is comparing the situation of control rod positions in two cooling modes (natural and force) at the same power of TRR. The difference between two control rod configurations is caused by the temperature difference in coolant in two modes. With measuring the difference reactivity and coolant temperature, we can calculate reactivity coefficient. The last new method is much more efficient than the above methods. The main advantage of this method is that we can measure the reactivity coefficient of reactor very Proceedings of the International Conference Nuclear Energy for New Europe, Portorož, Slovenia, September 14 17, 215

7 416.7 fast and independent of the control rods worth and positions. In each section reactivity coefficient are calculated and finally the average values of the temperature and power reactivity coefficient of the fuel and the coolant in TRR are: αt (F)=1.95 / C, αt (m)=13.57 / C αp (F)=.16 /kw, αp (m)=.89 /kw REFERENCES [1] GENERAL ATOMIC, Safeguards summary report for the New York University TRIGA Mark I Reactor. San Diego, 197. (GA-9864). [2] DUDERSTADT, J.J.; HAMILTON, L.J. Nuclear Reactor Analysis. New York, N.Y.: J. Wiley & Sons (1976). [3] 29. Abatte., P., CONVEC V 3.4, A program for Thermal-hydraulic Analysis of a MTR core in Natural Convection (22). Proceedings of the International Conference Nuclear Energy for New Europe, Portorož, Slovenia, September 14 17, 215