The Progress of I&C System Development in Small Modular Reactor ACP100. Chen Zhi Nuclear Power Institute of China 22 nd, May, 2013,Vienna

Transcription

1 The Progress of I&C System Development in Small Modular Reactor ACP100 Chen Zhi Nuclear Power Institute of China 22 nd, May, 2013,Vienna.1

2 CONTENTS 1. Background 2. The Main Technical Characteristics of ACP100 and Its R&D Progress 3. The Basic Design Principles and the Overall Architecture of ACP100 I&C System 4. The Preliminary Schemes of the Main I&C Systems of the ACP100 Nuclear Steam Supply System 5. Summary

3 1.1 The SMRs in the world Background The sole development of large NPPs is not meet with the extensive application demands of the huge electricity generation and non-electricity applications. More and more countries are developing the advanced small and medium sized reactors (SMRs) to meet the more extensive requirements. 13 SMRs are under construction in six countries and the approximately 45 innovative SMR concepts research for electricity generation and other applications is being carried out.

4 Background 1.2 The requirements for the SMRs in China The requirement in electricity generation area. SMR will be the best choice for vast inland areas and outlying areas of China.

5 Background 1.2 The requirements for the SMRs in China (Continued) The requirement in industrial and process heat supply area. The approximately 900 millions tons of industrial steams are consumed in China every year. The emission of greenhouse gases in the course of producing these steams occupies 10% of the total emission in China.

6 Background 1.2 The requirements for the SMRs in China (Continued) The requirement in desalination area. The fresh water sources are very lack in China. Most of industries locate in coastal areas, the serious lack of fresh water resources having become the bottleneck.

7 Background 1.2 The requirements for the SMRs in China (Continued) The requirement in city heat supply area. The energy demands of city heat become larger and lager in north cities of China, especially in Three North regions. The pollution condition is worse than before with the large consumption of fossil fuel due to the increasing demands of heat supply.

8 1.3 What is the ACP100? Background Considering the above background, China National Nuclear Corporation (CNNC) is carrying out the development of the small modular reactor, which is coded ACP100. It is an innovative PWR based on existing PWR technology, adapting passive safety system and integrated reactor design technology. ACP100

9 CONTENTS 1. Background 2. Main Technical Characteristics and R&D Progress of ACP Basic Design Principles and the Overall Architecture of ACP100 I&C System 4. The Preliminary Schemes of the Main I&C Systems of the ACP100 NSSS 5. Summary

10 Main Technical Characteristics 2.1 Main technical parameters of ACP100 Type Integrated PWR Thermal power 310 MWt Electrical power ~100 MWe Design life 60 years Refueling period 2 years Coolant average temperature 303 Operation pressure 15MPa(a) Fuel active section height 2150 mm

11 Main Technical Characteristics 2.1 Main design parameters of ACP100 (Continued) Fuel assembly number 57 Drive mechanism type magnetism lifting Control rod number 25 Reactivity control method Control rod solid burnable poison and boron Steam generator type OTSG Main steam pressure 4MPa(a) Main pump type canned pump SSE level ground seismic peak acceleration 0.3g

12 Main Technical Characteristics 2.2 Main technical characteristics of ACP100 Integrated layout of primary system and equipment. So the large LOCA accident is eliminated. And the dimension and the amount of the penetration in RPV can be also reduced. Large primary coolant inventory. The thermal inertia is increased. Small radioactivity storage quantity. ACP100

13 Main Technical Characteristics 2.2 Main technical characteristics of ACP100(Continued) The layout of RPV and equipment is benefit for natural circulation. Smaller decay thermal power. And it is easier to achieve safety by the way of passive. Reactor and spent fuel pool are laid under the ground level. So it is better to withstand exterior accident and good for the reduction of radioactive material release.

14 Main Technical Characteristics 2.3 Research and development progress Design work. Standard design, is completed by the end of The preliminary safety analysis report (PSAR) is also finished.

15 Main Technical Characteristics 2.3 Research and development progress Testing and verification aspects. The six test research subjects, including control rod drive line anti-earthquake test, passive emergency core cooling system integration test, etc., is planed to completed in Thermal hydraulic testing hall Passive emergency core cooling system

16 Main Technical Characteristics 2.3 Research and development progress (Continued) Licensing. The contract of ACP100 combined research with National Nuclear & Radiation Safety Center (NNRSC) was signed in And several specific research programmers and standard design safety analysis combined research with NNRSC will be carried out in year 2013.

17 Main Technical Characteristics 2.3 Research and development progress (Continued) Site selection. The demonstration ACP100 nuclear power plant, with two 310Mwth reactors, will be located in Putian City, Fujian Province in the east coast area of China. ACP100 Demonstration Site Expecting construction in June, 2014

18 CONTENTS 1. Background 2. Main Technical Characteristics and R&D Progress 3. Basic Design Principles and the Overall Architecture of ACP100 I&C System 4. Preliminary Schemes of the Main I&C systems of ACP100 NSSS 5. Summary

19 Basic Design Principles Meet the defense-in-depth concept Compliance with the single failure criterion Diversity design of I&C system

20 Basic Design Principles Meet the defense-in-depth concept During the normal operation and operational transients, the plant control systems will act to maintain and restore the plant normal operation; The reactor protection system will act to limit the consequences of any anticipated transient of malfunction; The reactor protection system will initiate selected protective functions to mitigate the consequences of design basis events; During the serious accidents, providing the serious accident monitoring and control function to limit the consequences of core melting and radioactive substances releasing.

21 Basic Design Principles Compliance with the single failure criterion RPS individual inputs and the logic outputs of the reactor trip: 4 divisions. Engineered Safety Feature Actuation System (ESFAS): also consists of individual divisions; The supporting systems (Power supply) follow the same redundancy design criteria the 4 protection division sets are power supplied by 4 independent the diverse actuation system is supplied by 2 independent UPS.

22 Basic Design Principles Diversity design of I&C system The different hardware and software platform will be applied to the safety and non-safety I&C systems separately. The digital RPS is designed according to functional diversity. For the CMF of the RPS, the diverse actuation system provides corresponding protection.

23 Overall Architecture I&C system overall architecture Site management level Operation and information level Control and protection level process interface level

24 CONTENTS 1. Background 2. Main Technical Characteristics and R&D Progress 3. Basic Design Principles and the Overall Architecture of ACP100 I&C System 4. Preliminary Schemes of the Main I&C systems of ACP100 NSSS 5. Summary

25 4.1 System configuraion System Configuration I&C systems of ACP100 Nuclear Steam Supply System (NSSS) include: Reactor nuclear instrumentation system Reactor protection system Diverse actuation system Reactor control system Rod control and rod position monitoring system Reactor in-core instrumentation system Loose parts and vibration monitoring system Other process control systems

26 General Control Scheme 4.2 The general control scheme Automatic control combined with the manual control Reactor power control system and the steam generator (SG) feedwater control system use the manual scheme instead of automatic scheme below 20% full power. The other control systems have an automatic control range from 0 to 100% full power. The systems are designed to return automatically to equilibrium conditions following the load variations in steps of ±10% FP transient and load variations in continuous ramps with a gradient of ±5%FP per minute transient.

27 System Preliminary scheme reactor nuclear instrumentation system system function continuously monitor reactor power signals are used for control and protection block automatic and manual rod withdrawal when neutron flux exceeds the interlock setpoint initiate reactor trip on high nuclear flux and high neutron flux variation

28 Systems Preliminary scheme reactor nuclear instrumentation system system scheme (A) source range consists of 4 identical and independent channels which furnish redundant neutron flux signals during shutdown and initial plant startup. Detector covers a flux range from 11-9 %FP to 10-3 %FP. (B) intermediate range consists of 4 identical and independent channels which furnish redundant neutron flux signals. Detector covers a range from about 10-6 %FP to 100%FP.

29 Systems Preliminary scheme reactor nuclear instrumentation system system scheme (C) power range consists of 4 identical and independent channels which furnish redundant neutron flux signals ã 180 ã 225 ã 270 ã ã 315 ã 0 ã 22.5 ã Detector covers a flux range from 10-6 %FP to 200%FP. The all detectors are arranged in the 12 independent holes outside the RPV. The cabinets include four safety protection cabinets and a non-safety control cabinet. 135 ã 45 ã ã 90 ã :sourge range(proportional counter) :intermediate range(compensated ion chamber) :power range(uncompensated ion chamber)

30 Systems Preliminary scheme SG feedwater control system system function maintain the OTSG secondary side pressure at a constant make the feedwater flow accommodate the load requirements

31 Systems Preliminary scheme SG feedwater control system system scheme (A) During 0 to 20%Pn, the feedwater flow is controlled by manual The bypass control valves are used to adjust the feedwater flow to the steam generator manually through the startup feedwater line and the startup feedwater pump. (B) During the 20%Pn to 100%Pn The feedwater flow is regulated automatically through the main feedwater pump, main feedwater control valve and the main feedwater line. The bypass feedwater channel is not operation at this condition.

32 Systems Preliminary scheme SG feedwater control system system scheme (Continued) The automatic control of the feedwater flow is accomplished by the main feedwater flow control valve in conjunction with the feedwater pump speed control. P across the main valve P setpoint pump speed pump speed controller SG steam pressure feedwater flow steam flow feedwater valve position feedwater valve controller

33 Systems Preliminary scheme reactor power control system system function maintain the reactor coolant average temperature to a constant under steady state conditions. enable the nuclear plant to accept a step load increase or decrease of 10% FP and a ramp increase or decrease of 5%FP per minute within the given load range without reactor trip or the steam dump system actuation.

34 Systems Preliminary scheme reactor power control system system scheme (A) the temperature channel compares the coolant average temperature measurement with the reference temperature (a fixed value). The error between them is the primary control signal of the RPC system. (B) the power mismatch channel receives the nuclear power signal and the total feed water flow of the secondary side. The error of the power mismatch channel is added to the temperature error signal. The final error signal is processed by the rod speed program which produces the control rod travel speed signal and two direction logic signals.

35 Systems Preliminary scheme reactor in-core instrumentation system system function The reactor in-core instrumentation (RII) system is made up of the neutron flux measurement subsystem the in-core temperature measurement subsystem the vessel level measurement subsystem.

36 Systems Preliminary scheme reactor in-core instrumentation system system function 1) The neutron flux measurement subsystem: measuring the real time core neutron flux and making flux map; providing the necessary inputs to the core online monitoring system and other data processing system; combining the other condition data of the reactor to check the calibration of the ex-core nuclear instrumentation.

37 Systems Preliminary scheme reactor in-core instrumentation system system function 2) The in-core temperature measurement subsystem : provides the temperature values of the reactor cooling water at the fuel assembly outlet. to calculate the reactor coolant maximum temperature, average temperature and the minimum core saturation margin. 3) The vessel level measurement subsystem provides the information whether the key points in the vessel is submerged or not.

38 Systems Preliminary scheme reactor in-core instrumentation system system scheme The all detectors are inserted into the core through the upper section of the RPV. The in-core neutron detector is self-power detector. The SPND and the thermocouple are integrated to the neutron and temperature detectors assemble. The in-core temperature measurement subsystem complies with the redundant principle. All equipments, including the Inadequate Core Cooling Monitoring (ICCM) cabinets, are divided into train A and train B, and are physically and electrically separated. The vessel level measurement subsystem also complies with the redundant principle and is divided into train A and train B.

39 Systems Preliminary scheme reactor in-core instrumentation system system scheme The layout of the detectors of the RII system. Thermocouples(TRAIN A) Thermocouples(TRAIN B) Level detectors(train A) Level detectors(train B) SPND layout Thermocouple and level detectors layout

40 Systems Preliminary scheme reactor protection system system function monitor the physical parameters which are essential to the reactor safety detect any transient changes in these parameters and triggered as required the operation of safety systems. limit the consequences of any accident conditions and then to ensure safety of the reactor.

41 Systems Preliminary scheme reactor protection system system scheme The ACP100 RPS has a configuration of four redundant divisions. Reactor trip Four redundant measurements, using four separate sensors. A partial trip signal for a parameter is generated if one channel s measurement exceeds its predetermined or calculated limit. Each division sends its partial trip status to each of the other three divisions over isolated data links. Each division is capable of generating a reactor trip signal if two or more of the redundant channels of a single variable are in the partial trip state. There are eight reactor trip switchgear breakers. Each of the four reactor trip divisions consists of two reactor trip circuit breakers. The reactor is tripped when two or more actuation divisions output a reactor trip signal. The reactor trip operates upon loss of voltage.

42 Systems Preliminary scheme reactor protection system system scheme Engineered Safety Feature Actuation System Four sensors normally monitor each variable used for an ESF actuation. When the measurement exceeds the setpoint, the output of the comparison results in a channel partial trip condition. The signals are combined within each division of ESF coincidence logic to generate a system-level signal. The ESF actuation operates on presence of the voltage.

43 Systems Preliminary scheme reactor protection system ACP100 RPS architecture

44 Systems Preliminary scheme Diverse actuation system system function The diverse actuation system (DAS) is a non-safety system that provides a diverse backup to the reactor protection. DAS monitors some plant variables, accomplishing reactor trip and ESF actuation in the case of the CMF occurrence in the RPS system.

45 Systems Preliminary scheme Diverse actuation system system scheme The diverse actuation system uses the different I&C platform from that being used by the reactor protection system. The diverse actuation system uses sensors that are separate from those being used by the protection system. Actuation interfaces are different between the diverse actuation system and the protection system.

46 Systems Preliminary scheme Diverse actuation system system scheme (continued) The signal from the sensor is hardwired to the DAS system. The measurements are compared against the setpoints for the diverse actuation to be generated. When the measurement exceeds the setpoint, the output of the comparison results in a channel partial trip condition. As a diverse design, the reactor trip signal generated by the DAS will be transmitted to the CRDM power supply cabinet instead of the reactor trip breaker. DAS architecture

47 CONTENTS 1. Background 2. Main Technical Characteristics and R&D Progress 3. Basic Design Principles and the Overall Architecture 4. Preliminary Schemes of the Main I&C systems 5. Summary

48 Summary ACP100 uses the passive safety technology and integrated reactor technology and has the high intrinsic safety. According to the primary system design, ACP100 I&C system standard design is completed by the institutes of CNNC. The general I&C architecture, main control room design, electricity supply plan, I&C equipment design, etc., will be optimized in the next design works aiming at the demonstration projects.

49 Thanks for your attention!.49