Thermal Hydraulic Simulations of the Angra 2 PWR

Transcription

1 Thermal Hydraulic Simulations of the Angra 2 PWR Javier González-Mantecón, Antonella Lombardi Costa, Maria Auxiliadora Fortini Veloso, Claubia Pereira, Patrícia Amélia de Lima Reis, Adolfo Romero Hamers, Maria Elizabeth Scari Departamento de Engenharia Nuclear, Universidade Federal de Minas Gerais Av. Antônio Carlos, 6627, Escola de Engenharia, Pampulha CEP , Belo Horizonte, Brazil mantecon1987@gmail.com, antonella@nuclear.ufmg.br, dora@nuclear.ufmg.br, claubia@nuclear.ufmg.br, patricialire@yahoo.com.br, adolforomerohamers@hotmail.com, melizabethscari@yahoo.com Abstract Angra 2, the second Brazilian nuclear power plant, began the commercial operation in The plant is a Pressurized Water Reactor (PWR) type with electrical output of about 1350 MW. In the present work, the thermal hydraulic RELAP5-3D code was used to develop a model of this reactor. The model was performed using geometrical and material data from the Angra 2 Final Safety Analysis Report (FSAR). Simulations of the reactor behavior during steadystate and loss of coolant accident were performed. Results of temperature distribution within the core, inlet and outlet coolant temperatures, coolant mass flow, and others parameters have been compared with the reference data and demonstrated to be in good agreement. This study demonstrates that the RELAP5-3D model is capable to reproduce the thermal hydraulic behavior of the Angra 2 PWR and it can contribute for the process of the plant safety analysis. I. INTRODUCTION As the global population increases, also grows up demand for energy and the benefits that it provides. With the worldwide concern over global warming, it is necessary to use clean sources which not cause the greenhouse effect. Nuclear energy is increasingly considered as an attractive energy source that can bring an answer to this increasing demands but, safety of nuclear power reactors, is one of the most important public worry. For many years, the main research area in the nuclear field has focused on the performance of the Nuclear Power Plants (NPP) during accident conditions. In order to simulate the behavior of water cooled reactors, the nuclear engineering community developed several complex thermal hydraulic codes systems. RELAP5-3D 1, developed by Idaho National Laboratory, is one of the most used bestestimate thermal hydraulic codes. It is capable performing steady-state, transient and postulated accident simulations including Loss of Coolant Accidents (LOCA) and a several types of transients in Light Water Reactors (LWRs). The aim of this work is to present simulations of the Angra 2 nuclear reactor behavior during steady-state and for a postulated loss of coolant accident in the primary circuit, Small Break Loss of Coolant Accident (SBLOCA), using the thermal hydraulic computer code RELAP5-3D 2-4. The accident simulated consists of a total break (310 cm 2 ) in the cold-leg piping of one of the loops of Angra 2, which is a PWR reactor with four primary loops. A variety of break sizes in the cold-leg and hot-leg piping and other parts of the reactor coolant system representing a typical range and locations of small- and medium-break LOCAs are described in the Final Safety Analysis Report of Angra 2 FSAR 5. II. MODEL DESCRIPTION II.A. Angra 2 Plant Description In June 1975, it was signed a cooperation agreement for the peaceful uses of nuclear energy between Brazil and the Federal Republic of Germany. Under this agreement Brazil acquired two nuclear plants, Angra 2 and 3, from the German company Siemens / KWU, currently Areva ANP. The Almirante Álvaro Alberto NPP - Unit 2 (Angra 2) is located on the Atlantic Coast in a bay at the western extremity of the Brazilian state of Rio de Janeiro. The second Brazilian Nuclear Power Plant (NPP) began commercial operation in The plant is equipped with a Pressurized Water Reactor with an electrical output of about 1350 MWe which utilizes light water as both reactor coolant and moderator. The PWR is designed as a four loop plant which is based on the proven technology of other four loop plants. Some main data of the plant are listed in Table I 5.

2 TABLE I Some Data of the Angra 2 NPP Parameter Thermal reactor output Gross electrical output Net electrical output Thermal yield Number of fuel assemblies Number of fuel rods per assemblies Value 3771 MWt 1350 MWe 1280 MWe 35.8 % of thermal hydraulic channels, and heat structures were associated to each one. A not heated channel represents the bypass (550). The thermal hydraulic channels and heat structures were subdivided axially in 34 volumes. II.B. RELAP5-3D Nodalization The RELAP5-3D nodalization diagram of the system adopted for the simulation is shown in Fig. 1 and Fig. 2, for the vessel and the loop, respectively. Fig. 2. RELA5-3D nodalization for the Angra 2 PWR: Loop Model. The reactor coolant pumps (RCPs) included in this system (N14) are identical for each loop and contain realistic characteristics. The coolant flow through the reactor is 4700 kg/s. The effective flow rate for heat transfer in the core is kg/s. III. RESULTS Fig. 1. RELA5-3D nodalization for the Angra 2 PWR: Vessel Model. The model contains 84 hydraulic components and 10 heat structures. It was performed using reference data from the Angra 2 FSAR 5. The primary system was modeled using one-dimensional components for these simulations. All four coolant loops are independently modeled. The loops were modeled symmetrically except for the differences due to the location of the pressurizer. Component numbers for the loop with pressurizer were between 102 and 133; while for the loops without pressurizer (2, 3 and 4) component numbers are indicated in the figure with an N. The model doesn t include steam generators and then Time-Dependent Volumes are used for simulating the thermodynamic characteristics of the fluid that enters and leaves the U-tube arrangement. The coolant flow area through the core was divided into ten regions (600 to 609) representing the same number The RELAP5-3D steady-state and transient calculations have been performed for Angra 2 PWR operating at 3771 MWt. The point reactor kinetics model was used to compute the transient behavior of the neutron fission power in the nuclear reactor for both simulations, steady-state and transient. III.A. Steady-State Simulation Table 2 shows the values of the thermal hydraulic parameters calculated by RELAP5-3D code during steadystate, and they are compared with the data provided in FSAR. The results show a good agreement with the reference data and errors calculated are in correspondence with the usual criteria for quantification of steady-state prediction quality that have been adopted 6. This means that the calculation model reproduces with good approximation the thermal behavior of the reactor.

3 TABLE 2 Comparison between the steady-state thermal hydraulic parameters calculated by RELAP5-3D code and FSAR data Parameter FSAR RELAP5-3D Error (%) T inlet ( C) T outlet ( C) P inlet (MPa) P outlet (MPa) Mass-flow (kg/s) The coolant mass flow has been calculated (Fig. 5). As it was expected, such value remains constant in kg/s. The error or difference between simulation results (RELAP-3D) and nominal values of Angra 2 included in FSAR was calculated by Eq. (1) Error = 100 (RELAP5-3D FSAR)/FSAR (1) The inlet and outlet temperatures of the coolant flow in the Reactor Pressure Vessel (RPV) are shown in Fig. 3. The water enters to the RPV vessel at temperature of C and pressure of MPa. When the coolant leaves the core its temperature has increased of approximately C. Fig. 4 also shows the inlet and outlet pressure. The pressure drop was 0.19 MPa. Fig. 5. Coolant mass flow through the reactor. III.B. Transient Analysis The break in the cold-leg of loop 2 was simulated by the addition of a trip valve (225) with 310 cm 2 of area in the pump discharge piping. A Time-Dependent Volume component (TDV), connected with the valve, was used to simulate the atmospheric pressure. Fig. 6 shows the nodalization to simulate the break. Fig. 3 Coolant temperature at the RPV inlet and outlet. Fig. 4. Coolant pressure at the RPV inlet and outlet. Fig. 6. RELA5-3D nodalization diagram for the break. The transient was initiated by an instantaneous opening of the break. The trip valve (225) was opened in the time of 100 seconds of calculation and remained in this way up to the final of the calculation. This is an extreme case of transient because safety systems are not being activated after the beginning of the transient and the reactor scram was not considered. The transient analysis is used to illustrate the capabilities of the model and the software. The results obtained cannot be compared because there are not available data. Future works will include a more detailed nodalization of the system and transient analysis will be compared with results from the FSAR. Fig. 7 shows the time evolution of the coolant mass flow in the core inlet (component 530). It decreases rapidly. However, as only one loop was closed, the others three maintained the coolant core mass flow with enough

4 quantity for the reactor operation. As a consequence of the transient, the coolant temperature increases reaching C at the core outlet (570) as it can be observed in the Fig. 8. The core pressure practically does not change. At such conditions of pressure and temperature, there is not void formation. conditions. In the transient simulation, one coolant loop was closed and the analyses demonstrate that the safe core operation was assured. The next step of this work will be to insert the safety dispositive and observe how they can mitigate consequences after a severe transient. Moreover, the neutronic parameters will be inserted in the RELAP5-3D model to realistically reproduce transient conditions with coupled thermal hydraulic/neutron kinetics effects. ACKNOWLEDGMENTS The authors are grateful to CAPES, FAPEMIG and CNPq for the support. Thanks also to Idaho National Laboratory for the license to use the RELAP5-3D computer software. NOMENCLATURE Fig. 7. Mass flow coolant time evolution in the core inlet after an extreme LOCA transient. The maximum fuel temperature remained with the value of C during this extreme transient. Therefore, the fuel temperature does not reach the fuel fusion point (about C) and the safe core operation is assured. FSAR Final Safety Analysis Report NPP Nuclear Power Plant LOCA Loss of Coolant Accident SBLOCA Small Break Loss of Coolant Accident RCP Reactor Coolant Pump RPV Reactor Pressure Vessel TDV Time-Dependent Volume REFERENCES 1. THE RELAP5-3D CODE DEVELOPMENT TEAM, RELAP5-3D User s Manual, Idaho National Laboratory, Idaho Falls (2009). 2. G. SABUNDJIAN, D. A. Andrade, A. Belchior Jr. et al., The Behavior of Angra 2 Nuclear Power Plant Core for a Small Break LOCA Simulated with RELAP5 Code, AIP Conference Proceedings, 1529, 151 (2013). Fig. 8. Coolant temperature time evolution in the core outlet after an extreme LOCA transient. IV. CONCLUSIONS The model was performed using geometrical and material data from the Angra 2 Final Safety Analysis Report. Simulations of the reactor behavior during steadystate and loss of coolant accident were performed. The analyzed parameters for the simulated cases demonstrated that the model is capable to reproduce the behavior of the reactor as in steady-state as in transient operation 3. M. S. ROCHA, G. Sabundjian, A. Belchior Jr. et al., Angra 2 Small Break LOCA Flow Regime Identification Through RELAP5 Code, Proceedings of ENCIT 2012, 14th Brazilian Congress of Thermal Sciences and Engineering, Rio de Janeiro, Brazil (2012). 4. T. CROOK, R. Vaghetto, A. Vanni and Y. A. Hassan, Sensitivity Analysis of a PWR Response During a Loss of Coolant Accident under a Hypothetical Core Blockage Scenario using RELAP5-3D, Proc. of the nd International Conference on Nuclear Engineering, ICONE22, Prague, Czech Republic (2014). 5. ELETROBRÁS ELETRONUCLEAR, Final Safety Analysis Report Central Nuclear Almirante Álvaro

5 Alberto Unit 2, Eletrobrás Termonuclear S.A., Rev. 13 (2013). 6. T. BAJS, D. Grgić, V. Sêgon, L. Oriani and L. E. Conway, Development of a RELAP5 Nodalization for IRIS Non-LOCA Transient Analyses, Nuclear Mathematical and Computational Sciences: A Century in Review, A Century Anew, Gatlinburg, United States (2003).