Safety Analysis of Loss of Flow Transients in a Typical Research Reactor by RELAP5/MOD3.3

Transcription

1 International Conference Nuclear Energy for New Europe 23 Portorož, Slovenia, September 8-11, 23 Safety Analysis of Loss of Flow Transients in a Typical Research Reactor by RELAP5/MOD3.3 ABSTRACT B. Di Maro, F. Pierro, M. Adorni, A. Bousbia Salah, and F. D Auria Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione Università di Pisa (Via Diotisalvi, Pisa, Italy) francopierro@tiscali.it, minodimaro@tiscali.it The main aim of the following study is to assess the RELAP5/MOD3.3 code capability in simulating transient dynamic behaviour in nuclear research reactors. For this purpose typical loss of flow transient in a representative MTR (Metal Test Reactor) fuel type Research Reactor is considered. The transient herein considered is a sudden pump trip followed by the opening of a safety valve in order to allow passive decay heat removal by natural convection. During such transient the coolant flow decay, originally downward, leads to a flow reversal and the cooling process of the core passes from forced, mixed and finally to natural circulation. This fact makes it suitable for evaluating the new features of RELAP5 to simulate such specific operating conditions. The instantaneous reactor power is derived through the point kinetic calculation, both protected and unprotected cases are considered (with and without Scram). The results obtained from this analysis were also compared with previous results obtained by old version RELAP5/MOD2 code. 1 INTRODUCTION The main purpose of the present work is to investigate the new features of the RELAP5/MOD3.3 code in simulating transients under Research Reactors operating conditions[1]. In this framework two Loss Of Flow Transient (LOFT) cases are considered. The first case is LOFT with and without Scram and it is based upon a realistic hypothesis which takes into account the power feedback using the point kinetic approach. In the second case the Scram is disabled and the power is imposed as input [2]. 2 MODELLING DESCRIPTION 2.1 Reactor description The reactor is a typical pool-type with an open water surface. The core is constituted by several assemblies of MTR fuel and graphite reflector blocs. The core of 1 MW power is cooled by a downward flow as shown in Fig. 1. The main operating as well as thermalhydraulic and kinetic characteristics of the core are outlined in Table 1[3]

2 Nuclear fuel Fuel element Coolant Moderator Reflector Control rod Table 1 : Main reactor operating data CORE MATERIAL MTR Plate-type clad in Al Light water (downward forced flow) Light water Graphite-Light water Absorbing rods in Ag-In-Cd CORE THERMALHYDRAULICS Fuel thermal Conductivity (W/cm o K).5 Fuel Density (g/cm 3 ) 4.45 Fuel Heat Capacity (J/g o K).34 Cladding thermal Conductivity (W/cm 1.8 o K) Cladding Density (g/cm 3 ) 2.7 Cladding Heat Capacity (J/g o K).892 Inlet coolant temperature ( C) 38. Operating pressure (bar) 1.7 CORE KINETICS Effective delayed neutron Fraction 7.67E-3 Prompt neutron generation time (µsec) Void feedback coefficient ($/% void).3257 Doppler feedback coefficient ( $/ o C) 3.6 E-5 Coolant temperature feedback ($/ o C) E-3 Figure 1 : Forced downward circulation in the pool during the steady-state 2.2 Coolant loop nodalisation Proceedings of the International Conference Nuclear Energy for New Europe, Portorož, Slovenia, Sept. 8-11, 23

3 218.3 The Fig.2 shows the coolant loop nadalisation for simulate the pump trip using the RELAP5 code,. Table 2 shows the correspondence between the main plant component and their equivalent node in the nodalisation scheme. [4] Table 2 : Main components of the nodalisation. COMPONENT EQUIVALENT ELEMENT Core 1 Natural convection valve 245 Primary side heat exchanger 33 Secondary side heat exchanger 34 Reactor pool Core bypass 15 Holdup tank 28 Pump 31 Figure 2 : Primary system nodalisation Proceedings of the International Conference Nuclear Energy for New Europe, Portorož, Slovenia, Sept. 8-11, 23

4 TRANSIENT RESULTS ANALYSIS Results obtained from RELAP5/MOD3.3 calculations are described in this section. Relevant output data are sketched in Figs. 3 to 11. In particular the Figs 4 to 6 show also the comparison between the RELAP5/MOD2 and MOD3.3 versions code [5]. The graphics obtained from the MOD2 analysis are been digitalized and for this reason they have an uncertainty. Case 1: power by point kinetic With SCRAM At 1 s the main coolant pump trip occurs and a rapid lowering of the flow of coolant in the core takes place. Therefore, the temperature of the fuel starts to grow as well as the coolant temperature, although the power produced stays constant (Fig.3). At 1,2 s the reactor SCRAM occurs due to the low core mass flow rate signal. Consequently, the fuel and the outlet coolant temperatures decrease due to the core power reduction (See Figs 4 and 5). When the coolant mass flow reaches about 3% of its nominal value, the natural convection valve opens. Shortly after, the buoyancy forces due to difference in fluid density begin to act (Figs 6 and 7). Mixed convection until 125 seconds: the node 1 temperature decreases due to the minor core power production (see Fig. 4). The node 1 temperature is nearly constant, because the mass flow is not inverted yet and the temperature is the same as in the pool. At 125 s the mass flow inverts and natural convection starts (Fig. 6). This causes the increases of the node 1 temperature (Fig.4 ). The coolant mass arrived at the node 1 is stopped and then its direction is inverted. The second peak is due to the time which the coolant mass remain stationary (Fig. 4). At 225 s when the warm mass inside the core is expelled, the natural convection is completely stabilized (Figs. 4 and 7) Power (MW) Figure 3 : Reactor power

5 218.5 temp outcore mod3.3 temp incore mod3.3 temp outcore mod2 temp incore mod Figure 4 : Coolant temperature mod3.3 mod temperature (K) time (s) Figure 5 : Fuel temperature. mod3.3 mod mass flow (kg/s) time (s) Figure 6 : Core mass flow

6 218.6 Figure 7 : Particle motion during the flow inversion Without SCRAM At 1 s the main coolant pump trip occurs and a rapid lowering of the flow of coolant in the core take place. Therefore, the temperature of the fuel starts to grow as well as the coolant temperature. In this case, the scram does not take place and the natural convection valve opens after about 1,5 s. As can be seen in Fig. 8, even thought the Scram is disabled, the feedback relative to fuel temperature grow and decrease of coolant density make the reactor power to decrease rapidly. After about 13,4 s the inversion of flow takes place and natural convection starts. Around 5 s there is a relative maximum of the power, due to the mass flow increase even if coolant and fuel temperature stay constant as shown in Figs 8,9,1 and 11. At 1. s calculations stopped POWER (MW) Figure 8 : Reactor power

7 218.7 Mass Flow (kg/s) Element 245 Volume Figure 9 : Mass flow in natural convective valve and in the core 36 Temperature (K) Volume 1 node 1 Volume 1 node Figure 1 : Coolant temperature at node 1 and 1

8 Temperature (K) Volume 1 BAF Volume 1 middle Figure11: Fuel temperature Case 2 : power imposed in time Without SCRAM Until 13,4 s the results are the same of case 1 with scram. After about 13,4 s the first inversion of flow takes place and natural convection starts. After about s there is a relative minimum in the fuel and coolant temperature, as well as a relative maximum of the heat transfer coefficient. This is due to the void generation in the core ( see Figs. 12,13, 14,15 and 16). After about 15. s the above mentioned situation repeats. After about s boiling starts from node 1 until it gets steady at node 1 in 13s. From s no more natural convection takes place Mass Flow (kg/s) Figure 12 : Coolant flow in the core.

9 Temperature volume 1 node 1 Temperature volume 1 node 1 Temperature (K) Figure 13 : Temperature of the coolant in the core inlet and outlet 42 4 Temperature (K) Temperature BAF Temperature middle Figure 14 : Fuel temperature.

10 Void Fraction Volume 1 node 1 Volume 1 node Figure 15 : Void fraction in the core. Heat Transfer Coefficient (W/m^2k) Volume 1 BAF Volume 1 MIDDLE Volume 1 TAF Figure 16 : Heat transfer coefficient

11 CONCLUSION Given what so far has been said, it is now possible to draw the following conclusions: In the steady-state no appreciably differences between the two versions of the RELAP5 code have been noted. In the case of transient, differences between the two versions have emerged. As boiling starts, natural convection does not occur. Using the point kinetic approach, the reactor operates safely under LOFT conditions even under extreme scenario without Scram. 5 BIBLIOGRAPHY [1] IAEA Safety guide on the Safety Assessment of Research Reactors and Preparation of the Safety Analysis Report, Safety Series 35-G1. [2] RELAP5/MOD3.3Beta, Code Manual volume II: Appendix A input requirements, ISL, Idaho, 21. [3] Todreas, N. E., Kazimi, S. M., Nuclear system, Taylor & Francis, 1993 [4] Boulheouchat, M. E., D Auria, F., d accidents dans un reacteur de recherche de type piscine», DIMNP, Pisa,1994. [5] Di Maro, B., Pierro, F., D Auria, F., Bousbia-Salah, A., Analysis of a Pump Trip in a Typical Research Reactor by RELAP5/MOD3.3, Proc. Int. Congress on Advances in Nuclear Power Plants,Cordoba, Spain, May 4-7, 23 [6] Incropera, P. F., Dewitt, P. D., Fundamentals of heat and mass transfer, Wiley & Sons, New York,1981