Today considering the world energy demand increase, the use of advanced nuclear power plants, have an important role in the environment and economic sustainability of country energy strategy mix considering the capacity of nuclear reactors of producing energy in safe and stable way contributing in cutting the CO2 emission (Bertel & Morrison, 2001; World Energy Outlook-Executive Summary, 2009; Wolde-Rufael & Menyah, 2010; Mascari et al., 2011d). According to the information’s provided by the “Power Reactor Information System” of the International Atomic Energy Agency (IAEA), today 433 nuclear power reactors are in operation in the world providing a total power installed capacity of 366.610 GWe, 5 nuclear reactors are in long term shutdown and 65 units are under construction (IAEA PRIS, 2011). In the last 20 years, the international community, taking into account the operational experience of the nuclear reactors, starts the development of new advanced reactor designs, to satisfy the demands of the people to improve the safety of nuclear power plants and the demands of the utilities to improve the economic efficiency and reduce the capital costs (D'Auria et al., 1993; Mascari et al., 2011c). Design simplifications and increased design margins are included in the advanced Light Water Reactors (LWR) (Aksan, 2005). In this framework, the project of some advanced reactors considers the use of emergency systems based entirely on natural circulation for the removal of the decay power in transient condition and in some reactors for the removal of core power during normal operating conditions (IAEA-TECDOC-1624, 2009; Mascari et al., 2010a; Mascari et al., 2011d). For example, if the normal heat sink is not available, the decay heat can be removed by using a passive connection between the primary system and heat exchangers (Aksan, 2005; Mascari et al., 2010a, Mascari, 2010b). The AP600/1000 (Advanced Plant 600/1000 MWe) design, for example, includes a Passive Residual Heat Removal (PRHR) system consisting of a C-Tube type heat exchanger immersed in the In-containment Refueling Water Storage Tank (IRWST) and connected to one of the Hot Legs (HL) (IAEA-TECDOC-1391, 2004; Reyes, 2005c; Gou et al., 2009; Mascari et al., 2010a). A PRHR from the core via Steam Generators (SG) to the atmosphere, considered in the WWER-1000/V-392 (Water Moderated, Water Cooled Energy Reactor) design, consists of heat exchangers cooled by atmospheric air, while the PRHR via SGs, considered in the WWER-640/V-407 design, consists of heat exchangers immersed in emergency heat removal tanks installed outside the containment (Kurakov et al., 2002; IAEA-TECDOC-1391, 2004; Gou et al., 2009; Mascari et al., 2010a). In the AC-600 (Advanced Chinese PWR) the PRHR heat exchangers are cooled by atmospheric air (IAEATECDOC 1281, 2002; Zejun et al., 2003; IAEA-TECDOC-1391, 2004; Gou et al., 2009; Mascari et al., 2010a) and in the System Integrated Modular Advanced Reactor (SMART) the PRHR heat exchangers are submerged in an in-containment refuelling water tank (IAEA-TECDOC- 1391, 2004; Lee & Kim, 2008; Gou et al., 2009; Mascari et al., 2010a). The International Reactor Innovative and Secure (IRIS) design includes a passive Emergency Heat Removal System (EHRS) consisting of an heat exchanger immersed in the Refueling Water Storage Tank (RWST). The EHRS is connected to a separate SG feed and steam line and the RWST is installed outside the containment structure (Carelli et al., 2004; Carelli et al., 2009; Mascari, 2010b; Chiovaro et al., 2011). In the advanced BWR designs the core water evaporates, removing the core decay heat, and condenses in a heat exchanger placed in a pool. Then the condensate comes back to the core (Hicken & Jaegers, 2002; Mascari et al., 2010a). For example, the SWR-1000 (Siede Wasser Reaktor, 1000 MWe) design has emergency condensers immersed in a core flooding pool and connected to the core, while the ESBWR (Economic Simplified Boiling Water Reactor) design uses isolation condensers connected to the Reactor Pressure Vessel (RPV) and immersed in external pools (IAEA-TECDOC-1391, 2004; Aksan, 2005; Mascari et al., 2010a). The designs of some advanced reactors rely on natural circulation for the removing of the core power during normal operation. Examples of these reactors are the MASLWR (Multi- Application Small Light Water Reactor), the ESBWR, the SMART and the Natural Circulation based PWR being developed in Argentina (CAREM)(IAEA-TECDOC-1391, 2004; IAEA -TECDOC-1474, 2005; Mascari et al., 2010a). In particular the MASLWR (Modro et al., 2003), figure 1, is a small modular integral Pressurized Water Reactor (PWR) relying on natural circulation during both steady-state and transient operation. In the development process of these advanced nuclear reactors, the analysis of single and two-phase fluid natural circulation in complex systems (Zuber, 1991; Levy, 1999; Reyes & King, 2003; IAEA-TECDOC-1474, 2005; Mascari et al., 2011e), under steady state and transient conditions, is crucial for the understanding of the physical and operational phenomena typical of these advanced designs. The use of experimental facilities is fundamental in order to characterize the thermal hydraulics of these phenomena and to develop an experimental database useful for the validation of the computational tools necessary for the operation, design and safety analysis of nuclear reactors. In general it is expensive to design a test facility to develop experimental data useful for the analyses of complex system, therefore reduced scaled test facilities are, in general, used to characterize them. Since the experimental data produced have to be applicable to the full-scale prototype, the geometrical characteristics of the facility and the initial and boundary conditions of the selected tests have to be correctly scaled. Since possible scaling distortions are present in the experimental facility design, the similitude of the main thermal hydraulic phenomena of interest has to be assured permitting their accurate experimental simulation (Zuber, 1991; Reyes, 2005b; Reyes et al., 2007; Mascari et al., 2011e). Fig. 1. MASLWR conceptual design layout (Modro et al, 2003; Reyes et al., 2007; Mascari et al., 2011a). Different computer codes have been developed to characterize two-phase flow systems, from a system and a local point of view. Accurate simulation of transient system behavior of a nuclear power plant or of an experimental test facility is the goal of the best estimate thermal hydraulic system code. The evaluation of a thermal hydraulic system code’s calculation accuracy is accomplished by assessment and validation against appropriate system thermal hydraulic data, developed either from a running system prototype or from a scaled model test facility, and characterizing the thermal hydraulic phenomena during both steady state and transient conditions. The identification and characterization of the relevant thermal hydraulic phenomena, and the assessment and validation of thermal hydraulic systems codes, has been the objective of multiple international research programs (Mascari et al., 2011a; Mascari et al., 2011c). In this international framework, Oregon State University (OSU) has constructed, under a U.S. Department of Energy grant, a system level test facility to examine natural circulation phenomena of importance to the MASLWR design. The scaling analysis of the OSUMASLWR experimental facility was performed in order to have an adequately simulation of the single and two-phase natural circulation, reactor system depressurization during a blowdown and the containment pressure response typical of the MASLWR prototype (Zuber, 1991; Reyes & King, 2003; Reyes, 2005b). A previous testing program has been conducted in order to assess the operation of the prototypical MASLWR under normal full pressure and full temperature conditions and to assess the passive safety systems under transient conditions (Modro et al. 2003; Reyes & King, 2003; Reyes, 2005b; Reyes et al., 2007; Mascari et al., 2011e). The experimental data developed are useful also for the assessment and validation of the computational tools necessary for the operation, design and safety analysis of nuclear reactors. For many years, in order to analyze the LWR reactors, the USNRC has maintained four thermal-hydraulic codes of similar, but not identical, capabilities, the RAMONA, RELAP5, TRAC-B and TRAC-P. In the last years, the USNRC is developing an advanced best estimate thermal hydraulic system code called TRAC/RELAP Advanced Computational Engine or TRACE, by merging the capabilities of these previous codes, into a single code (Boyac & Ward, 2000; TRACE V5.0, 2010; Reyes, 2005a; Mascari et al., 2011a). The validation and assessment of the TRACE code against the MASLWR natural circulation database, developed in the OSU-MASLWR test facility, is a novel effort. This chapter illustrates an analysis of the primary/containment coupling phenomena characterizing the MASLWR design mitigation strategy during a SBLOCA scenario and, in the framework of the performance assessment and validation of thermal hydraulic system codes, a qualitative analysis of the TRACE V5 code capability in reproducing it.
Analysis of primary/containment coupling phenomena characterizing the MASLWR design during a SBLOCA scenario
D'AURIA, FRANCESCO SAVERIO
2012-01-01
Abstract
Today considering the world energy demand increase, the use of advanced nuclear power plants, have an important role in the environment and economic sustainability of country energy strategy mix considering the capacity of nuclear reactors of producing energy in safe and stable way contributing in cutting the CO2 emission (Bertel & Morrison, 2001; World Energy Outlook-Executive Summary, 2009; Wolde-Rufael & Menyah, 2010; Mascari et al., 2011d). According to the information’s provided by the “Power Reactor Information System” of the International Atomic Energy Agency (IAEA), today 433 nuclear power reactors are in operation in the world providing a total power installed capacity of 366.610 GWe, 5 nuclear reactors are in long term shutdown and 65 units are under construction (IAEA PRIS, 2011). In the last 20 years, the international community, taking into account the operational experience of the nuclear reactors, starts the development of new advanced reactor designs, to satisfy the demands of the people to improve the safety of nuclear power plants and the demands of the utilities to improve the economic efficiency and reduce the capital costs (D'Auria et al., 1993; Mascari et al., 2011c). Design simplifications and increased design margins are included in the advanced Light Water Reactors (LWR) (Aksan, 2005). In this framework, the project of some advanced reactors considers the use of emergency systems based entirely on natural circulation for the removal of the decay power in transient condition and in some reactors for the removal of core power during normal operating conditions (IAEA-TECDOC-1624, 2009; Mascari et al., 2010a; Mascari et al., 2011d). For example, if the normal heat sink is not available, the decay heat can be removed by using a passive connection between the primary system and heat exchangers (Aksan, 2005; Mascari et al., 2010a, Mascari, 2010b). The AP600/1000 (Advanced Plant 600/1000 MWe) design, for example, includes a Passive Residual Heat Removal (PRHR) system consisting of a C-Tube type heat exchanger immersed in the In-containment Refueling Water Storage Tank (IRWST) and connected to one of the Hot Legs (HL) (IAEA-TECDOC-1391, 2004; Reyes, 2005c; Gou et al., 2009; Mascari et al., 2010a). A PRHR from the core via Steam Generators (SG) to the atmosphere, considered in the WWER-1000/V-392 (Water Moderated, Water Cooled Energy Reactor) design, consists of heat exchangers cooled by atmospheric air, while the PRHR via SGs, considered in the WWER-640/V-407 design, consists of heat exchangers immersed in emergency heat removal tanks installed outside the containment (Kurakov et al., 2002; IAEA-TECDOC-1391, 2004; Gou et al., 2009; Mascari et al., 2010a). In the AC-600 (Advanced Chinese PWR) the PRHR heat exchangers are cooled by atmospheric air (IAEATECDOC 1281, 2002; Zejun et al., 2003; IAEA-TECDOC-1391, 2004; Gou et al., 2009; Mascari et al., 2010a) and in the System Integrated Modular Advanced Reactor (SMART) the PRHR heat exchangers are submerged in an in-containment refuelling water tank (IAEA-TECDOC- 1391, 2004; Lee & Kim, 2008; Gou et al., 2009; Mascari et al., 2010a). The International Reactor Innovative and Secure (IRIS) design includes a passive Emergency Heat Removal System (EHRS) consisting of an heat exchanger immersed in the Refueling Water Storage Tank (RWST). The EHRS is connected to a separate SG feed and steam line and the RWST is installed outside the containment structure (Carelli et al., 2004; Carelli et al., 2009; Mascari, 2010b; Chiovaro et al., 2011). In the advanced BWR designs the core water evaporates, removing the core decay heat, and condenses in a heat exchanger placed in a pool. Then the condensate comes back to the core (Hicken & Jaegers, 2002; Mascari et al., 2010a). For example, the SWR-1000 (Siede Wasser Reaktor, 1000 MWe) design has emergency condensers immersed in a core flooding pool and connected to the core, while the ESBWR (Economic Simplified Boiling Water Reactor) design uses isolation condensers connected to the Reactor Pressure Vessel (RPV) and immersed in external pools (IAEA-TECDOC-1391, 2004; Aksan, 2005; Mascari et al., 2010a). The designs of some advanced reactors rely on natural circulation for the removing of the core power during normal operation. Examples of these reactors are the MASLWR (Multi- Application Small Light Water Reactor), the ESBWR, the SMART and the Natural Circulation based PWR being developed in Argentina (CAREM)(IAEA-TECDOC-1391, 2004; IAEA -TECDOC-1474, 2005; Mascari et al., 2010a). In particular the MASLWR (Modro et al., 2003), figure 1, is a small modular integral Pressurized Water Reactor (PWR) relying on natural circulation during both steady-state and transient operation. In the development process of these advanced nuclear reactors, the analysis of single and two-phase fluid natural circulation in complex systems (Zuber, 1991; Levy, 1999; Reyes & King, 2003; IAEA-TECDOC-1474, 2005; Mascari et al., 2011e), under steady state and transient conditions, is crucial for the understanding of the physical and operational phenomena typical of these advanced designs. The use of experimental facilities is fundamental in order to characterize the thermal hydraulics of these phenomena and to develop an experimental database useful for the validation of the computational tools necessary for the operation, design and safety analysis of nuclear reactors. In general it is expensive to design a test facility to develop experimental data useful for the analyses of complex system, therefore reduced scaled test facilities are, in general, used to characterize them. Since the experimental data produced have to be applicable to the full-scale prototype, the geometrical characteristics of the facility and the initial and boundary conditions of the selected tests have to be correctly scaled. Since possible scaling distortions are present in the experimental facility design, the similitude of the main thermal hydraulic phenomena of interest has to be assured permitting their accurate experimental simulation (Zuber, 1991; Reyes, 2005b; Reyes et al., 2007; Mascari et al., 2011e). Fig. 1. MASLWR conceptual design layout (Modro et al, 2003; Reyes et al., 2007; Mascari et al., 2011a). Different computer codes have been developed to characterize two-phase flow systems, from a system and a local point of view. Accurate simulation of transient system behavior of a nuclear power plant or of an experimental test facility is the goal of the best estimate thermal hydraulic system code. The evaluation of a thermal hydraulic system code’s calculation accuracy is accomplished by assessment and validation against appropriate system thermal hydraulic data, developed either from a running system prototype or from a scaled model test facility, and characterizing the thermal hydraulic phenomena during both steady state and transient conditions. The identification and characterization of the relevant thermal hydraulic phenomena, and the assessment and validation of thermal hydraulic systems codes, has been the objective of multiple international research programs (Mascari et al., 2011a; Mascari et al., 2011c). In this international framework, Oregon State University (OSU) has constructed, under a U.S. Department of Energy grant, a system level test facility to examine natural circulation phenomena of importance to the MASLWR design. The scaling analysis of the OSUMASLWR experimental facility was performed in order to have an adequately simulation of the single and two-phase natural circulation, reactor system depressurization during a blowdown and the containment pressure response typical of the MASLWR prototype (Zuber, 1991; Reyes & King, 2003; Reyes, 2005b). A previous testing program has been conducted in order to assess the operation of the prototypical MASLWR under normal full pressure and full temperature conditions and to assess the passive safety systems under transient conditions (Modro et al. 2003; Reyes & King, 2003; Reyes, 2005b; Reyes et al., 2007; Mascari et al., 2011e). The experimental data developed are useful also for the assessment and validation of the computational tools necessary for the operation, design and safety analysis of nuclear reactors. For many years, in order to analyze the LWR reactors, the USNRC has maintained four thermal-hydraulic codes of similar, but not identical, capabilities, the RAMONA, RELAP5, TRAC-B and TRAC-P. In the last years, the USNRC is developing an advanced best estimate thermal hydraulic system code called TRAC/RELAP Advanced Computational Engine or TRACE, by merging the capabilities of these previous codes, into a single code (Boyac & Ward, 2000; TRACE V5.0, 2010; Reyes, 2005a; Mascari et al., 2011a). The validation and assessment of the TRACE code against the MASLWR natural circulation database, developed in the OSU-MASLWR test facility, is a novel effort. This chapter illustrates an analysis of the primary/containment coupling phenomena characterizing the MASLWR design mitigation strategy during a SBLOCA scenario and, in the framework of the performance assessment and validation of thermal hydraulic system codes, a qualitative analysis of the TRACE V5 code capability in reproducing it.I documenti in IRIS sono protetti da copyright e tutti i diritti sono riservati, salvo diversa indicazione.
