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FIG. II-11. Containment vessel of NuScale [41].

Figure 59 II-11. Containment vessel of NuScale [41].

Related Figures (101)

The Agency’s Statute was approved on 23 October 1956 by the Conference on the Statute of the IAEA held at United Nations Headquarters, New York; it entered into force on 29 July 1957. The Headquarters of the Agency are situated in Vienna. Its principal objective is “to accelerate and enlarge the contribution of atomic energy to peace, health and prosperity throughout the world”’. The following States are Members of the International Atomic Energy Agency:
The Agency’s Statute was approved on 23 October 1956 by the Conference on the Statute of the IAEA held at United Nations Headquarters, New York; it entered into force on 29 July 1957. The Headquarters of the Agency are situated in Vienna. Its principal objective is “to accelerate and enlarge the contribution of atomic energy to peace, health and prosperity throughout the world”’. The following States are Members of the International Atomic Energy Agency:
FIG. 3. The elevations and locations of structures and components at the Fukushima Daiichi NPP with reference to the Onahama Port [1].
FIG. 3. The elevations and locations of structures and components at the Fukushima Daiichi NPP with reference to the Onahama Port [1].
FIG. 6. The accident progressions at the Fukushima Daiichi nuclear plant complex. There are indications that, after about 68 hours, the RCIC system failed. It was therefore no longer possible to inject water into the RPV because it was at high pressure. It is estimated that the Unit 2 reactor core began to melt about 76 hours after the tsunami. The accident progression at the Fukushima Daiichi nuclear power plant is briefly depicted in FIG. 6. Situation in Unit 3 was similar to that in Unit 2 except that some DC power was still functioning after the tsunami. At the which was manually actuated by oper. RCIC provided injection to the core water level was reached. The tsunami beginning the core cooling was provided until it automatically stopped when a hig fortunately the indications for pressure and reactor water level were not affected ind off operation for RCIC was used by operator to maintain the water level a by RCIC ator about 20 minutes after the reactor scram. This n reactor hen caused the loss of some DC power systems but . An ‘on about 4 meters above the top of active fuel (TAF). Pressure inside the reactor vessel was controlled oy safety relief valves (SRVs). However the condition did not stay long. After 14 hours o1 continued operation of the emergency HPCI system, the Unit 3 operators became concerned about the reliability and possible failure of the system’s turbine powering the njection pump, which was by then operating at low reactor steam pressure. The concern was related to the possibility of turbine damage and the creation of a release path from the ‘eactor vessel. This would result in an uncontrollable release of radioactive steam directly yutside the primary containment. This concern was heightened when the turbine did not
FIG. 6. The accident progressions at the Fukushima Daiichi nuclear plant complex. There are indications that, after about 68 hours, the RCIC system failed. It was therefore no longer possible to inject water into the RPV because it was at high pressure. It is estimated that the Unit 2 reactor core began to melt about 76 hours after the tsunami. The accident progression at the Fukushima Daiichi nuclear power plant is briefly depicted in FIG. 6. Situation in Unit 3 was similar to that in Unit 2 except that some DC power was still functioning after the tsunami. At the which was manually actuated by oper. RCIC provided injection to the core water level was reached. The tsunami beginning the core cooling was provided until it automatically stopped when a hig fortunately the indications for pressure and reactor water level were not affected ind off operation for RCIC was used by operator to maintain the water level a by RCIC ator about 20 minutes after the reactor scram. This n reactor hen caused the loss of some DC power systems but . An ‘on about 4 meters above the top of active fuel (TAF). Pressure inside the reactor vessel was controlled oy safety relief valves (SRVs). However the condition did not stay long. After 14 hours o1 continued operation of the emergency HPCI system, the Unit 3 operators became concerned about the reliability and possible failure of the system’s turbine powering the njection pump, which was by then operating at low reactor steam pressure. The concern was related to the possibility of turbine damage and the creation of a release path from the ‘eactor vessel. This would result in an uncontrollable release of radioactive steam directly yutside the primary containment. This concern was heightened when the turbine did not
FIG. 7. Emergency heat removal system through steam generator. produced by decay heat in the SG is routed to the heat exchangers where it is condensed. The condensate flows back to the SG through SG feed water inlet. SMR designs which implement this method, among others, are the passive emergency heat removal system (EHRS) of IRIS, the passive residual heat removal (PRHR) system of SMART and the decay heat removal system (DHRS) of NuScale. Residual heat removal using passively cooled condenser
FIG. 7. Emergency heat removal system through steam generator. produced by decay heat in the SG is routed to the heat exchangers where it is condensed. The condensate flows back to the SG through SG feed water inlet. SMR designs which implement this method, among others, are the passive emergency heat removal system (EHRS) of IRIS, the passive residual heat removal (PRHR) system of SMART and the decay heat removal system (DHRS) of NuScale. Residual heat removal using passively cooled condenser
Some advanced designs of pressurized tank systems provide longer injection capability, which eliminate the need of forced low pressure injection system, as in accumulator designs of APR1400, APWR and ATMEAI1. Some integral PWR type SMRs uses an elevated borated water tank which use gravitational force to inject emergency borated water into the primary system, as depicted in FIG. 10. The top of the tank is connected to the reactor vessel with normally open valve, and the discharge line at the bottom of the tank is isolated from the reactor vessel by a
Some advanced designs of pressurized tank systems provide longer injection capability, which eliminate the need of forced low pressure injection system, as in accumulator designs of APR1400, APWR and ATMEAI1. Some integral PWR type SMRs uses an elevated borated water tank which use gravitational force to inject emergency borated water into the primary system, as depicted in FIG. 10. The top of the tank is connected to the reactor vessel with normally open valve, and the discharge line at the bottom of the tank is isolated from the reactor vessel by a
FIG. 10. Gravity driven injection tank (core make-up tank). normally closed valve. During an emergency situation, the bottom valve opens and as a result the borated water flows down to the vessel, simultaneously cooling the core and terminating the fission. The SMR designs which implement this concept, among others, are the emergency boration tank of IRIS and the core make up tank of Westinghouse SMR.
FIG. 10. Gravity driven injection tank (core make-up tank). normally closed valve. During an emergency situation, the bottom valve opens and as a result the borated water flows down to the vessel, simultaneously cooling the core and terminating the fission. The SMR designs which implement this concept, among others, are the emergency boration tank of IRIS and the core make up tank of Westinghouse SMR.
In addition to gravity driven system or pressurized tank system, a core injection system is composed using a turbine driven pump. The steam from the reactor pressure vessel drives a pump which transfers water from a storage tank/pool into the reactor vessel. This approach of injection system is widely used in BWR reactors and known as the reactor core isolation cooling (RCIC) system.
In addition to gravity driven system or pressurized tank system, a core injection system is composed using a turbine driven pump. The steam from the reactor pressure vessel drives a pump which transfers water from a storage tank/pool into the reactor vessel. This approach of injection system is widely used in BWR reactors and known as the reactor core isolation cooling (RCIC) system.
FIG. 12. Low-head pump injection system. An active low pressure injection system is implemented in APR1400, APWR and ATMEAI. In this system, low-head pump is used to inject water from in-containment water pool to the vessel, as shown in FIG. 12. Moreover, the high head injection system is useful if the LOCA is small. It develops high pressure and low flow which makes up for the coolant loss rate. If the LOCA is large the pressure in reactor is low and the coolant loss rate is high. Low-head injection system is designed for this. The cooling flow from ultimate heat-sink is usually provided to a heat-exchanger in the low-head injection system. This is because after the initial high head injection, low-head injection system is needed to remove the decay heat.
FIG. 12. Low-head pump injection system. An active low pressure injection system is implemented in APR1400, APWR and ATMEAI. In this system, low-head pump is used to inject water from in-containment water pool to the vessel, as shown in FIG. 12. Moreover, the high head injection system is useful if the LOCA is small. It develops high pressure and low flow which makes up for the coolant loss rate. If the LOCA is large the pressure in reactor is low and the coolant loss rate is high. Low-head injection system is designed for this. The cooling flow from ultimate heat-sink is usually provided to a heat-exchanger in the low-head injection system. This is because after the initial high head injection, low-head injection system is needed to remove the decay heat.
FIG. 13. Passive low pressure injection system.
FIG. 13. Passive low pressure injection system.
FIG. 14. Safety injection using pressurized tank. Pressurized tank is also used for low pressure injection as appears in the Emergency Injection System of CAREM25. The tank containing borated water is connected to RPV with primary function of preventing core uncovery during LOCA. During LOCA, when the pressure in the reactor vessel becomes relatively low (1.5 MPa), the rupture disks separating the accumulator tanks and the RPV would open to refill reactor vessel, as shown in FIG. 14.
FIG. 14. Safety injection using pressurized tank. Pressurized tank is also used for low pressure injection as appears in the Emergency Injection System of CAREM25. The tank containing borated water is connected to RPV with primary function of preventing core uncovery during LOCA. During LOCA, when the pressure in the reactor vessel becomes relatively low (1.5 MPa), the rupture disks separating the accumulator tanks and the RPV would open to refill reactor vessel, as shown in FIG. 14.
Some designs use water pool/tank for carrying out pressure suppression. High temperature steam released from the reactor vessel (through breaks, safety relief valve and ADS) is directed to a suppression pool/tank, as shown in FIG. 16. The steam condenses in the pool, thus mitigating the pressure increases in the containment. This type of containment has been used in BWR designs for many years. SMR designs which implement this method for their containment are CAREM25 and IRIS.
Some designs use water pool/tank for carrying out pressure suppression. High temperature steam released from the reactor vessel (through breaks, safety relief valve and ADS) is directed to a suppression pool/tank, as shown in FIG. 16. The steam condenses in the pool, thus mitigating the pressure increases in the containment. This type of containment has been used in BWR designs for many years. SMR designs which implement this method for their containment are CAREM25 and IRIS.
external surface of the reactor vessel is cooled by water induced from in-containment water pool. Reactor vessel cavity is flooded so that lower part of the vessel is always submerged in the water during severe accident, as depicted in F/G. 20. In this way the integrity of the vessel is maintained and corium can be retained inside the vessel. This kind of strategy to deal with severe accident is used by many advanced reactor designs such as AP1000, SMART, IRIS, mPower, etc. In order to prevent the molten core material from escaping the containment building, a specific structure/device is placed below the reactor vessel. It is a space made of heat- resistant concrete ceramic to prevent molten core from penetrating through. Core catcher spreads the molten core and vessel material to decrease thermal density and it is usually designed with a cooling mechanism to keep the molten core stable for extended period, as shown in FIG. 21. Advanced reactor designs which implement core catcher include ESBWR and ATMEAI. In the ESBWR, the BiMac core catcher is used.
external surface of the reactor vessel is cooled by water induced from in-containment water pool. Reactor vessel cavity is flooded so that lower part of the vessel is always submerged in the water during severe accident, as depicted in F/G. 20. In this way the integrity of the vessel is maintained and corium can be retained inside the vessel. This kind of strategy to deal with severe accident is used by many advanced reactor designs such as AP1000, SMART, IRIS, mPower, etc. In order to prevent the molten core material from escaping the containment building, a specific structure/device is placed below the reactor vessel. It is a space made of heat- resistant concrete ceramic to prevent molten core from penetrating through. Core catcher spreads the molten core and vessel material to decrease thermal density and it is usually designed with a cooling mechanism to keep the molten core stable for extended period, as shown in FIG. 21. Advanced reactor designs which implement core catcher include ESBWR and ATMEAI. In the ESBWR, the BiMac core catcher is used.
3. Hydrogen control devices
3. Hydrogen control devices
TABLE 1. DESIGN FEATURES OF WATER COOLED SMR DESIGNS CONTRIBUTING TO LEVEL 1 OF DEFENCE IN DEPTH
TABLE 1. DESIGN FEATURES OF WATER COOLED SMR DESIGNS CONTRIBUTING TO LEVEL 1 OF DEFENCE IN DEPTH
TABLE 5. DESIGN FEATURES OF WATER COOLED SMR DESIGNS CONTRIBUTING TO LEVEL 5 OF DEFENCE IN DEPTH
TABLE 5. DESIGN FEATURES OF WATER COOLED SMR DESIGNS CONTRIBUTING TO LEVEL 5 OF DEFENCE IN DEPTH
TABLE 6. STRENGTHENING MEASURES AGAINST EXTREME NATURAL HAZARDS AND CONSEQUENTIAL EFFECTS
TABLE 6. STRENGTHENING MEASURES AGAINST EXTREME NATURAL HAZARDS AND CONSEQUENTIAL EFFECTS
TABLE 10. ENHANCING DESIGN OF SAFETY-RELATED STRUCTURES, SYSTEMS AND COMPONENTS
TABLE 10. ENHANCING DESIGN OF SAFETY-RELATED STRUCTURES, SYSTEMS AND COMPONENTS
TABLE 12. CONSIDERATIONS TO ENHANCE THE CONTAINMENT VENTING AND FILTERING SYSTEM
TABLE 12. CONSIDERATIONS TO ENHANCE THE CONTAINMENT VENTING AND FILTERING SYSTEM
TABLE 13. ENSURING HARDENED INSTRUMENTATION AND CABLES FOR SAFETY—RELATED PARAMETERS AND MONITORING EQUIPMENT
TABLE 13. ENSURING HARDENED INSTRUMENTATION AND CABLES FOR SAFETY—RELATED PARAMETERS AND MONITORING EQUIPMENT
TABLE 17. ENHANCING HUMAN CAPABILITIES AND CAPACITIES
TABLE 17. ENHANCING HUMAN CAPABILITIES AND CAPACITIES
TABLE 18. STRENGTHEN OFF-SITE INFRASTRUCTURE AND CAPABILITY
TABLE 18. STRENGTHEN OFF-SITE INFRASTRUCTURE AND CAPABILITY
TABLE 19. STRENGTH AND RESPONSE ENING NATIONAL ARRANGEMENTS FOR EMI] ERGENCY PREPAREDNESS
TABLE 19. STRENGTH AND RESPONSE ENING NATIONAL ARRANGEMENTS FOR EMI] ERGENCY PREPAREDNESS
TABLE 20. ENHANCING INTERACTION AND COMMUNICATION WITH THE INTERNATIONAL COMMUNITY
TABLE 20. ENHANCING INTERACTION AND COMMUNICATION WITH THE INTERNATIONAL COMMUNITY
TABLE 21. CONSIDERATIONS TO REVIEW AND CLARIFY REGULATORY AND EMERGENCY RESPONSE FRAMEWORK
TABLE 21. CONSIDERATIONS TO REVIEW AND CLARIFY REGULATORY AND EMERGENCY RESPONSE FRAMEWORK
TABLE 22. REINFORCING SAFETY REGULATORY BODIES AND LEGAL STRUCTURES
TABLE 22. REINFORCING SAFETY REGULATORY BODIES AND LEGAL STRUCTURES
The BWR/4 steam supply system mainly consists of a reactor pressure vessel — which contains the core, steam dryer and separator— that is connected to pipes to allow re- circulation pumps to circulate the coolant water inside the vessel. In addition, a set of control and safety systems to ensure safe operation and a containment system which is intended as barrier for radioactive release during accident and as protection against external hazards are provided. The safety system is made up of redundant and multiple equipment to guarantee the intended functions. Some of their functionalities are to provide emergency core cooling, to control and maintain containment’s pressure and temperature below its design limit, and to provide additional control for core reactivity. In BWR/4 the emergency core cooling function is facilitated by the low pressure coolant injection (LPCI) mode of residual heat removal (RHR) system, RCIC system, HPCI system, automatic depressurization system (ADS), and core spray system (CS) as can be seen in FIG. I-/. Meanwhile the other two functions are provided by the containment spray system (CSS) and standby liquid control (SLC) system.
The BWR/4 steam supply system mainly consists of a reactor pressure vessel — which contains the core, steam dryer and separator— that is connected to pipes to allow re- circulation pumps to circulate the coolant water inside the vessel. In addition, a set of control and safety systems to ensure safe operation and a containment system which is intended as barrier for radioactive release during accident and as protection against external hazards are provided. The safety system is made up of redundant and multiple equipment to guarantee the intended functions. Some of their functionalities are to provide emergency core cooling, to control and maintain containment’s pressure and temperature below its design limit, and to provide additional control for core reactivity. In BWR/4 the emergency core cooling function is facilitated by the low pressure coolant injection (LPCI) mode of residual heat removal (RHR) system, RCIC system, HPCI system, automatic depressurization system (ADS), and core spray system (CS) as can be seen in FIG. I-/. Meanwhile the other two functions are provided by the containment spray system (CSS) and standby liquid control (SLC) system.
When the condensate and feed water system is not available or the main steam line is isolated, core cooling is provided by the RCIC system as shown in FIG. J-2. The system consists of a steam turbine driven pump and associated valves and piping capable of supplying water to the reactor vessel at operating conditions. The turbine is powered by steam produced from decay heat in the core, which flows through a steam line to the RCIC turbine and then discharges to the suppression pool. 1.1.3. High pressure coolant injection system
When the condensate and feed water system is not available or the main steam line is isolated, core cooling is provided by the RCIC system as shown in FIG. J-2. The system consists of a steam turbine driven pump and associated valves and piping capable of supplying water to the reactor vessel at operating conditions. The turbine is powered by steam produced from decay heat in the core, which flows through a steam line to the RCIC turbine and then discharges to the suppression pool. 1.1.3. High pressure coolant injection system
FIG. I-4. BWR/4 core spray system [20]. The main purpose of the core spray system (CS) is to provide a low pressure spray cooling to the reactor core under LOCA conditions. There are two independent pumping loops of CS, each consists of a set of spray nozzles located on independent ring spargers put within the core shroud above the fuel assembly (FA). The nozzles are arranged to provide uniform coolant flow to the fuel assemblies. The system provides enough cooling water and can be powered by emergency power system. The schematic diagram of CS is shown in FIG. I-4.
FIG. I-4. BWR/4 core spray system [20]. The main purpose of the core spray system (CS) is to provide a low pressure spray cooling to the reactor core under LOCA conditions. There are two independent pumping loops of CS, each consists of a set of spray nozzles located on independent ring spargers put within the core shroud above the fuel assembly (FA). The nozzles are arranged to provide uniform coolant flow to the fuel assemblies. The system provides enough cooling water and can be powered by emergency power system. The schematic diagram of CS is shown in FIG. I-4.
FIG. I-5. ECCS of BWR/5 with Mark IT containment [19]. For dealing with the design basis accident (DBA), the safety feature of this reactor is provided by emergency core cooling system as shown in F/G. J-5, which consists of high pressure core spray system (HPCF), ADS, low pressure core spray system (LPCS), and LPCI mode of residual heat removal [19]. These systems are elaborated in more detail in the following.
FIG. I-5. ECCS of BWR/5 with Mark IT containment [19]. For dealing with the design basis accident (DBA), the safety feature of this reactor is provided by emergency core cooling system as shown in F/G. J-5, which consists of high pressure core spray system (HPCF), ADS, low pressure core spray system (LPCS), and LPCI mode of residual heat removal [19]. These systems are elaborated in more detail in the following.
FIG. I-6. ABWR plant system (Reproduced courtesy of GE Hitachi Nuclear Energy) [17]. The standard ABWR plant design was certified and licensed in the United States of America, followed by Japan and Taiwan, China. The developer also produces an adapted design to meet European requirement which has net power output of about 1,600 MW(e) (4300 MW(th)). This plant is designed to have 60 years life time, refueling interval of 18 — 24 months and availability greater than 90%. The overall plant system of the ABWR is shown in FIG. I-6.
FIG. I-6. ABWR plant system (Reproduced courtesy of GE Hitachi Nuclear Energy) [17]. The standard ABWR plant design was certified and licensed in the United States of America, followed by Japan and Taiwan, China. The developer also produces an adapted design to meet European requirement which has net power output of about 1,600 MW(e) (4300 MW(th)). This plant is designed to have 60 years life time, refueling interval of 18 — 24 months and availability greater than 90%. The overall plant system of the ABWR is shown in FIG. I-6.
The emergency core cooling system (ECCS) of ABWR has been improved in many areas compared to the previous generation of BWRs, providing a higher level of defence-in- depth against accidents, contingencies, and incidents. The overall system is divided up into 3 divisions; where each division has capability to terminate the limiting fault/design basis accident (DBA) prior to core uncovery, even in the event of loss of offsite power and loss of proper feed water. The ECCS consists of high pressure core flooder (HPCF), reactor core isolation cooling, residual heat removal system and automatic depressurization system. The configuration of ABWR’s ECCS is shown in FIG. J-7. [he HPCF provides several functionalities: first, it is to maintain the RPV inventory during 1 small break LOCA which does not depressurize the reactor vessel. Secondly, it serves a: 1 backup for the RCIC system in response to transients. The system is provided in twe livisions and can be powered by | EDG in case auxiliary power is unavailable. Both HPCE ystems take their suction from the condensate storage pool (primary source) or from th« uppression pool (secondary source). The discharge capacity is 182 m’/hour at ratec yressure. This system is automatically activated by either high pressure in the dry well o: ow water level in the reactor vessel. The HPCF of ABWR is shown in FIG. J-8.
The emergency core cooling system (ECCS) of ABWR has been improved in many areas compared to the previous generation of BWRs, providing a higher level of defence-in- depth against accidents, contingencies, and incidents. The overall system is divided up into 3 divisions; where each division has capability to terminate the limiting fault/design basis accident (DBA) prior to core uncovery, even in the event of loss of offsite power and loss of proper feed water. The ECCS consists of high pressure core flooder (HPCF), reactor core isolation cooling, residual heat removal system and automatic depressurization system. The configuration of ABWR’s ECCS is shown in FIG. J-7. [he HPCF provides several functionalities: first, it is to maintain the RPV inventory during 1 small break LOCA which does not depressurize the reactor vessel. Secondly, it serves a: 1 backup for the RCIC system in response to transients. The system is provided in twe livisions and can be powered by | EDG in case auxiliary power is unavailable. Both HPCE ystems take their suction from the condensate storage pool (primary source) or from th« uppression pool (secondary source). The discharge capacity is 182 m’/hour at ratec yressure. This system is automatically activated by either high pressure in the dry well o: ow water level in the reactor vessel. The HPCF of ABWR is shown in FIG. J-8.
The system automatically initiates at RPV water level signal and its capacity is sufficient for make up on loss of feed water without support from any other make up system. The ECCS. In the event of LOCA, the RCIC is designed to be RCIC is considered as part of | able to pump water to the vessel from full pressure down to about 150 psig. The steam generated by reactor decay he at is then directed to the wet well suppression pool. To maintain the pool temperature within acceptable limit, the RHR heat exchanger is used.
The system automatically initiates at RPV water level signal and its capacity is sufficient for make up on loss of feed water without support from any other make up system. The ECCS. In the event of LOCA, the RCIC is designed to be RCIC is considered as part of | able to pump water to the vessel from full pressure down to about 150 psig. The steam generated by reactor decay he at is then directed to the wet well suppression pool. To maintain the pool temperature within acceptable limit, the RHR heat exchanger is used.
This system is intended to maintain an inert atmosphere within the primary containment and always active during all plant operating mode except during plant shutdown for refueling and maintenance. When activated, the system capable to reduce the concentration of oxygen to lower than 3.5% by volume in less than 4 hrs. The ACS is also equipped with a containment overpressure protection system (COPS) which is designed to reduce the containment pressure whenever the containment integrity is challenged by overpressure following an accident. 9. Flammability control system (FCS)
This system is intended to maintain an inert atmosphere within the primary containment and always active during all plant operating mode except during plant shutdown for refueling and maintenance. When activated, the system capable to reduce the concentration of oxygen to lower than 3.5% by volume in less than 4 hrs. The ACS is also equipped with a containment overpressure protection system (COPS) which is designed to reduce the containment pressure whenever the containment integrity is challenged by overpressure following an accident. 9. Flammability control system (FCS)
FIG. I-13. ESBWR safety system configuration (Reproduced courtesy of GE Hitachi Nuclear Energy) [27]. ESBWR is a BWR designed by GE Hitachi Nuclear Energy. The reactor has a rated power of about 4,500 MW(th) (1600 MW(e)) and its construction time can be completed within 36 months. The distinctive feature of this reactor is the use of natural forces for its normal and accident conditions which eliminates the need for safety grade diesel generators. Natural circulation is used for in-vessel recirculation during normal operation. Hence, no recirculation pump is needed to bring the water from the steam separator to down comer. In addition, passive safety feature is implemented to response the design basis accidents and severe accidents involving core melt. The passive safety system consists of isolation condenser system (ICS), Emergency core cooling - gravity driven cooling system (GDCS) and passive containment cooling system (PCCS). With all these functions, no operator action is required for 72 hours during the design basis accident [11]. The configuration of these systems is shown in FIG. J-13.
FIG. I-13. ESBWR safety system configuration (Reproduced courtesy of GE Hitachi Nuclear Energy) [27]. ESBWR is a BWR designed by GE Hitachi Nuclear Energy. The reactor has a rated power of about 4,500 MW(th) (1600 MW(e)) and its construction time can be completed within 36 months. The distinctive feature of this reactor is the use of natural forces for its normal and accident conditions which eliminates the need for safety grade diesel generators. Natural circulation is used for in-vessel recirculation during normal operation. Hence, no recirculation pump is needed to bring the water from the steam separator to down comer. In addition, passive safety feature is implemented to response the design basis accidents and severe accidents involving core melt. The passive safety system consists of isolation condenser system (ICS), Emergency core cooling - gravity driven cooling system (GDCS) and passive containment cooling system (PCCS). With all these functions, no operator action is required for 72 hours during the design basis accident [11]. The configuration of these systems is shown in FIG. J-13.
ATMEAI is a 3-loops PWR type plant with electric generation capability about 1100 MW(e) that jointly developed by two nuclear plant suppliers: the Mitsubishi Heavy Industry and ARI thermal efficienc EVA. The reactor is designed using proven technologies to achieve higher y, better availability, and high level of safety. In addition, its deployment is adaptive to various site conditions and be able to provide flexible operability in response to customer need. s. The general plant layout inside containment is shown in FIG. I-16. Typical safety system design of ATMEAI consists of 3-trains active systems which are individually installed to the three reactor loops. Each train is adequate to mitigate accidents, to ensure safe shut-down and to perform residual heat removal of the overall reactor. The train is also complemented with passive features that are provided by advanced accumulators and in in-containment water storage. Besides, the system is arranged so that it is fully independent from other safety system and from the operational system. The trains are installed into dedicated areas, called ‘divisions’ where each division is physically separated from the others (by walls, floors, etc.) so spreading of internal hazards from one division to another can be avoided. Emergency power sources are also segregated, making them less susceptible to a common cause failure. In order to deal with
ATMEAI is a 3-loops PWR type plant with electric generation capability about 1100 MW(e) that jointly developed by two nuclear plant suppliers: the Mitsubishi Heavy Industry and ARI thermal efficienc EVA. The reactor is designed using proven technologies to achieve higher y, better availability, and high level of safety. In addition, its deployment is adaptive to various site conditions and be able to provide flexible operability in response to customer need. s. The general plant layout inside containment is shown in FIG. I-16. Typical safety system design of ATMEAI consists of 3-trains active systems which are individually installed to the three reactor loops. Each train is adequate to mitigate accidents, to ensure safe shut-down and to perform residual heat removal of the overall reactor. The train is also complemented with passive features that are provided by advanced accumulators and in in-containment water storage. Besides, the system is arranged so that it is fully independent from other safety system and from the operational system. The trains are installed into dedicated areas, called ‘divisions’ where each division is physically separated from the others (by walls, floors, etc.) so spreading of internal hazards from one division to another can be avoided. Emergency power sources are also segregated, making them less susceptible to a common cause failure. In order to deal with
station blackout (SBO) additional alternative AC power system (AAC) or diversity in EPSs is provided. The safety system configuration to cope with the Design Basis Event is shown in FIG. I-17. The system includes safety injection system, advanced accumulator, and RHRS/CS [30}] — [321].
station blackout (SBO) additional alternative AC power system (AAC) or diversity in EPSs is provided. The safety system configuration to cope with the Design Basis Event is shown in FIG. I-17. The system includes safety injection system, advanced accumulator, and RHRS/CS [30}] — [321].
The use of advanced accumulator in ECCS provides longer injection time to the reactor vessel post LOCA initiation. The design allows two rates of flow going out from the accumulator tank. At the beginning of blow down phase, the flow injection rate is high and then at subsequent core re-flooding phase (when water level inside the accumulator tank is lower), the flow damper switches the injection flow rate passively into low head injection. This prolongs the injection time and at the same time integrates function of low head injection system. Consequently the low head injection system (LHIS) is no longer needed
The use of advanced accumulator in ECCS provides longer injection time to the reactor vessel post LOCA initiation. The design allows two rates of flow going out from the accumulator tank. At the beginning of blow down phase, the flow injection rate is high and then at subsequent core re-flooding phase (when water level inside the accumulator tank is lower), the flow damper switches the injection flow rate passively into low head injection. This prolongs the injection time and at the same time integrates function of low head injection system. Consequently the low head injection system (LHIS) is no longer needed
FIG. II-1. Reactor vessel of CAREM25 (Reproduced courtesy of CNEA, Argentina) [34]. The reactor vessel internals and major design characteristics of CAREM25 are shown in FIG, II-1 and TABLE II-1.
FIG. II-1. Reactor vessel of CAREM25 (Reproduced courtesy of CNEA, Argentina) [34]. The reactor vessel internals and major design characteristics of CAREM25 are shown in FIG, II-1 and TABLE II-1.
TABLE II-1. DESIGN CHARACTERISTIC OF CAREM25. [34], [35] Design internalization of Defence in Depth concept Defence in Depth (DiD) concept was internalized in the design of CAREM25 since the conceptual engineering. It is the base for structures, systems and components safety classification, which allows a clear assignation of design rules and requirements to systems important to safety.
TABLE II-1. DESIGN CHARACTERISTIC OF CAREM25. [34], [35] Design internalization of Defence in Depth concept Defence in Depth (DiD) concept was internalized in the design of CAREM25 since the conceptual engineering. It is the base for structures, systems and components safety classification, which allows a clear assignation of design rules and requirements to systems important to safety.
FIG. II-4. Reactor system configuration of SMART (Reproduced courtesy of KAERI) [36].
FIG. II-4. Reactor system configuration of SMART (Reproduced courtesy of KAERI) [36].
TABLE II-2. Basic design information of SMART
TABLE II-2. Basic design information of SMART
TABLE II-3. SUMMARY OF ENGINEERED SAFETY FEATURE FOR SMART.
TABLE II-3. SUMMARY OF ENGINEERED SAFETY FEATURE FOR SMART.
FIG. II-5. Schematic diagram of passive safety systems of SMART (Reproducec courtesy of KAERI). The reactor can be shut down under any circumstances by inserting control rods or boron injection. The passive sa fety system of SMART maintains the plant in a safe shutdown condition following design basis accidents such as LOCA or non-LOCA transient events without safety grade AC power or operator actions. Passive residual heat removal system prevents over-heating and over-pressurizing of the primary system in case of emergency situations where normal steam extraction or feed water supply is unavailable. It removes the decay and sensible heat by natural circulation of a two-phase fluid. The core is maintained undamaged without any corrective action taken by the operator for at least 72 hours. The reactor overpressure is prevented through the opening of the pressurizer safety valve. The containment building is resistant to any kind of seismic activity and can withstand possible air-crash incident. A schematic diagram of the safety systems of the SMART is shown in FIG. I/-5. TABLE II-3 provides summary of the ESFs adopted in the SMART.
FIG. II-5. Schematic diagram of passive safety systems of SMART (Reproducec courtesy of KAERI). The reactor can be shut down under any circumstances by inserting control rods or boron injection. The passive sa fety system of SMART maintains the plant in a safe shutdown condition following design basis accidents such as LOCA or non-LOCA transient events without safety grade AC power or operator actions. Passive residual heat removal system prevents over-heating and over-pressurizing of the primary system in case of emergency situations where normal steam extraction or feed water supply is unavailable. It removes the decay and sensible heat by natural circulation of a two-phase fluid. The core is maintained undamaged without any corrective action taken by the operator for at least 72 hours. The reactor overpressure is prevented through the opening of the pressurizer safety valve. The containment building is resistant to any kind of seismic activity and can withstand possible air-crash incident. A schematic diagram of the safety systems of the SMART is shown in FIG. I/-5. TABLE II-3 provides summary of the ESFs adopted in the SMART.
TABLE II-4. POST-FUKUSHIMA ACTION ITEMS APPLIED TO SMART. It is clear that maintaining continuous and proper core cooling capability after shutdown of a reactor is very important. A series of realistic simulation showed that the SMART PRHRS working on the secondary side of steam generators effectively removes decay heat, and maintains the reactor in a stable condition for 20 days without external power and operator action. This grace time can be extended continuously when the PRHRS emergency cool down tanks (ECTs) are periodically replenished. After the Fukushima Daiichi accident, the Korean nuclear regulatory authority formed a task team to perform a through safety audit on all the nuclear plants and facilities under operation, construction, or development. As a result, fifty (50) improvement orders (action items) were derived to enhance the safety level against Fukushima Daiichi type extreme hazards including earthquake, tsunami, and flooding. Among them, ten (10) items are applied to SMART while other items are not directly relevant to the SMART design. TABLE II-4 shows the post-Fukushima action items applied to the standard design of SMART.
TABLE II-4. POST-FUKUSHIMA ACTION ITEMS APPLIED TO SMART. It is clear that maintaining continuous and proper core cooling capability after shutdown of a reactor is very important. A series of realistic simulation showed that the SMART PRHRS working on the secondary side of steam generators effectively removes decay heat, and maintains the reactor in a stable condition for 20 days without external power and operator action. This grace time can be extended continuously when the PRHRS emergency cool down tanks (ECTs) are periodically replenished. After the Fukushima Daiichi accident, the Korean nuclear regulatory authority formed a task team to perform a through safety audit on all the nuclear plants and facilities under operation, construction, or development. As a result, fifty (50) improvement orders (action items) were derived to enhance the safety level against Fukushima Daiichi type extreme hazards including earthquake, tsunami, and flooding. Among them, ten (10) items are applied to SMART while other items are not directly relevant to the SMART design. TABLE II-4 shows the post-Fukushima action items applied to the standard design of SMART.
TABLE II-5. IRIS MAJOR DESIGN PARAMETERS [39].
TABLE II-5. IRIS MAJOR DESIGN PARAMETERS [39].
FIG. II-8. The mPower reactor diagram (Reproduced courtesy of B&W Generation mPower) [40].
FIG. II-8. The mPower reactor diagram (Reproduced courtesy of B&W Generation mPower) [40].
FIG. II-9. Decay heat removal strategy of the mPower [40].
FIG. II-9. Decay heat removal strategy of the mPower [40].
TABLE II-6. THE NUSCALE BASIC PLANT PARAMETERS
TABLE II-6. THE NUSCALE BASIC PLANT PARAMETERS
FIG, II-12. Schematic of the decay heat removal system of NuScale [41].
FIG, II-12. Schematic of the decay heat removal system of NuScale [41].
The ECCS operates in the event of a LOCA or loss of the main feed water flow in combination with the loss of the DHR system. It consists of two independent reac valves and two independent reactor recirculation valves. ECCS removes the core opening the vent valve located in lines exiting the top of the reactor pressure vessel. The opening lets the steam from the reactor to the containment which is then coo condensed on the inside surface of the containment vessel by the pool water. The condensate water then accumulates in the lower containment region. When the ed and evel of water in the region rises above the top of recirculation valves, the recirculation valves open. This establishes a natural circulation path from the core to the containment.
The ECCS operates in the event of a LOCA or loss of the main feed water flow in combination with the loss of the DHR system. It consists of two independent reac valves and two independent reactor recirculation valves. ECCS removes the core opening the vent valve located in lines exiting the top of the reactor pressure vessel. The opening lets the steam from the reactor to the containment which is then coo condensed on the inside surface of the containment vessel by the pool water. The condensate water then accumulates in the lower containment region. When the ed and evel of water in the region rises above the top of recirculation valves, the recirculation valves open. This establishes a natural circulation path from the core to the containment.
a.-Pressure-Vessel-and Internals b.-Containment-‘System The plant is equipped with passive safety features derived from the AP1000 plant design. Configuration of the passive cooling system is shown in FIG. I-14. The main components of the passive core cooling systems are a high pressure steel containment vessel, four core makeup tanks (CMTs), an in-containment pool (ICP) and associated ICP tanks, an automatic depressurization system (ADS), an outside containment pool (OCP) and two ultimate heat sink (UHS) tanks. Connected to the CMTs are passive residual heat removal heat exchangers. Altogether, these components serve as the protection required to mitigate the various initiating faults. The safety systems is designed to safely shut down the nuclear reaction, remove decay heat following shutdown, guarantee that the reactor core remains covered with water to maintain effective cooling, and provide long term cooling and shutdown.
a.-Pressure-Vessel-and Internals b.-Containment-‘System The plant is equipped with passive safety features derived from the AP1000 plant design. Configuration of the passive cooling system is shown in FIG. I-14. The main components of the passive core cooling systems are a high pressure steel containment vessel, four core makeup tanks (CMTs), an in-containment pool (ICP) and associated ICP tanks, an automatic depressurization system (ADS), an outside containment pool (OCP) and two ultimate heat sink (UHS) tanks. Connected to the CMTs are passive residual heat removal heat exchangers. Altogether, these components serve as the protection required to mitigate the various initiating faults. The safety systems is designed to safely shut down the nuclear reaction, remove decay heat following shutdown, guarantee that the reactor core remains covered with water to maintain effective cooling, and provide long term cooling and shutdown.
The secondary side of each CMT is connected through a closed loop of piping to a heat exchanger that sits in one of two UHS tanks. Each UHS tank is designed to accommodate decay heat removal from the core and spent fuel pool for at least 72 hours. When combined with the water in the OCP, seven days of decay heat removal capability is available. The two UHS tanks are physically separated to prevent an external event from compromising both tanks. Connections to each UHS tank allow for the addition of water to extend the decay heat removal indefinitely.
The secondary side of each CMT is connected through a closed loop of piping to a heat exchanger that sits in one of two UHS tanks. Each UHS tank is designed to accommodate decay heat removal from the core and spent fuel pool for at least 72 hours. When combined with the water in the OCP, seven days of decay heat removal capability is available. The two UHS tanks are physically separated to prevent an external event from compromising both tanks. Connections to each UHS tank allow for the addition of water to extend the decay heat removal indefinitely.
b. CMT Natural Circulation / Draining
b. CMT Natural Circulation / Draining
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