SMALL BREAK LOSS OF COOLANT ACCIDENT OF 200 cm2 IN COLD LEG OF PRIMARY LOOP OF ANGRA 2 NUCLEAR POWER REACTOR EVALUATION

The aim of this paper is evaluated the consequences to ANGRA 2 nuclear power reactor and to identify the flow regimes, the heat transfer modes, and the correlations used by RELAP5/MOD3.2.gama code in ANGRA 2 during the Small-Break Loss-of-Coolant Accident (SBLOCA) with a 200cm of rupture area in the cold leg of primary loop. The Chapter 15 of the Final Safety Analysis Report of ANGRA 2 (FSAR-A2) reports this specific kind of accident. The results from this work demonstrated the several flow regimes and heat transfer modes that can be present in the core of ANGRA 2 during the postulated accident. The results obtained for ANGRA 2 nuclear reactor core during the postulated accident were satisfactory when compared with the FSAR-A2. Additionally, the results showed the correct actuation of the ECCS guaranteeing the integrity of the reactor core.


INTRODUCTION
The evaluated accident consists of the partial break of the cold leg of the ANGRA 2 nuclear power plant. The rupture is the 200 cm2 and the efficiency of the Emergency Core Coolant System (ECCS) is verified for this accident.
The aim of this paper is to identify the flow regimes, the heat transfer modes, and the correlations used in the RELAP5/MOD3.2.gama [1] code in ANGRA 2 during the Small-Break Loss-of-Coolant Accident (SBLOCA) with a 200cm2 of rupture area in the cold leg of primary loop how described in detail in the Chapter 15 of the Final Safety Analysis Report of ANGRA 2 (FSAR-A2) [2].
Results presented in this paper showed the correct actuation of the ECCS guaranteeing the integrity of the ANGRA 2 reactor core. This plant has a PWR built by Siemens-KWU (now Areva NPP), resulting from an agreement between Brazil and Germany in 1975. ANGRA 2 is a reactor with 1,350MWe capacity providing energy to a 2-million-inhabitant city. ANGRA 2 has four pumps to control of water flow, four loops with two ECCS (Emergency Core Cooling System) for each loop (one Hot and one Cold ECCS). Figure 1 shows the arrangement of the components of ANGRA 2 nuclear power plant [3]. Source: Mitsubishi Heavy Industries.

RELAP5 code
The RELAP5 code was developed by the Idaho National Laboratory. This code was originally designed for the analysis of thermal hydraulic transients in Pressurized Water Reactors (PWR). The RELAP5 can model the primary and secondary cooling systems of experimental facilities and of Nuclear Reactors with geometric details. The program uses the non-homogeneous non-equilibrium two-fluid model, and considers the conservation equations of mass, momentum and energy for the liquid and gas phases. One-dimensional model is used to treat the fluid flow and the heat conduction in the structures; however, in some special cases such as the cross flow in the reactor core and the rewetting region in flooding model, the two-dimensional model is used [1].
The RELAP5 code uses and is capable to identify fifteen different flow regimes, which are presented in Table 1. Each one associated to an integer number. Those numbers are obtained from RELAP5 code output file to specify the fluid behavior for each control volume during the accident simulation [1]. Tables 2 and 3 show the mode numbers and the wall convection heat transfer used in RELAP5 code, respectively [1]. They were accessed during the execution of the program to this case, and the results are presented in the next item of this paper.  Single-phase vapor convection or supercritical pressure with the void fraction greater than zero 10 Condensation when the void is less than one 11 Condensation when the void equals one Souce: Idaho Lab. Scientech Inc.

RESULTS AND DISCUSSION
The accident started after 100 seconds of the steady state simulation time, when the valve 951 was opened. Valve 951 is connected to the branch 255 (primary cold leg), which is connected to the volume 960 (containment). The area of the valve opening is 200 cm2. This is the size of the simple rupture considered in this case.
The input file was based in the work performed by the Technical Cooperation among Instituto de Pesquisas Energética e Nucleares (IPEN), Centro de Desenvolvimento Tecnológico Nuclear (CDTN), and Comissão Nacional de Energia Nuclear (CNEN) [19,20]. Reference [21] shows details about simulations with the RELAP5 code of SBLOCAs in ANGRA 2 reactor.  The main boundary conditions used in this simulation were obtained from the FSAR-A2 [2] and are presented as the following: • reactor power -106% nominal power; • reactor trip from Reactor Coolant System (RCS) pressure < 132 bar; • 100 k/h secondary-side cooldown (PRCS < 132 bar and containment pressure > 1.03 bar); • ECC criteria met (PRCS < 110 bar and containment pressure > 1.03 bar).
For each postulated LOCA, the ECCS performance is different. The Chapter 15 of the Final Safety Analysis Report of ANGRA 2 (FSAR-A2) reports the ECCS actuation for each accident [2].
In this case failure and repair criteria for the ECCS components were adopted as specified to this event in the FSAR-A2 in order to verify the system operation, preserving the integrity of the reactor core and to guarantee its cooling, as presented in Table 4. SBLOCA accidents are characterized by a slow blow down in the primary circuit to values that the high pressure injection system is activated. The thermal-hydraulic processes inherent to the accident phenomenon, such as hot leg of ECCS vaporization and consequently core vaporization causing an inappropriate flow distribution in the reactor core, can lead to a reduction in the core liquid level, until the ECCS is capable to refill it. These are the principal reasons to identifying and understand the flow regimes and the heat transfer modes used by RELAP5 code in the nuclear core of ANGRA 2 during accident simulation. Table 5 provides a summary of the analyzed accident, the temporal sequence of operation and evaluation of ANGRA 2 nuclear reactor ECCS performance. According to some results provided by RELAP5 and FSAR-A2 it was possible to observe the differences related to onset time of some phenomena. Figures 3 to 12 show the results obtained from SBLOCA of ANGRA 2 analysis using RELAP5 code. Some of these results were compared with the results found in the FSAR-A2. Some results obtained using RELAP5 were similar to the results of the FSAR-A2 [2]. Figure 3 shows the pressures in the primary and secondary loops to RELAP5 and FSAR-A2. It is note that in RELAP5 code simulation the primary pressure decreases faster than FSAR-A2 one. The end of Simulation 1300 1300       only vapor in the rupture was observed to RELAP5 simulation. Note that after 630 seconds the void fraction to RELAP5 simulation is higher than FSAR-A2 one.  Between 360 and 515 seconds, only vapor as observed to RELAP5 simulation. There aren´t FSAR-A2 core void fraction data.    Figure 12 shows three points of hot rod core cladding temperature of ANGRA 2 to RELAP5 simulation and FSAR-A2. These RELAP5 data are higher than FSAR-A2 one. But the hottest rod core cladding temperature is lower than 800 °C.

CONCLUSION
In this work the flow regimes, the heat transfer modes, and the correlation used by RELAP5/MOD3.2.gama code, during the SBLOCA with 200cm2 of rupture area in the cold leg of primary loop were identified.
The evaluation of the most important variables in this simulated accident with RELAP5 code, when compared to their FSAR-A2 data one, showed that the analysis of RELAP5 was more conservative than the FSAR-A2.
Results presented in this paper showed the correct actuation of the ECCS guaranteeing the integrity of ANGRA 2 reactor core.