This standard specifies the basic safety requirements that the nuclear reactor protection system shall meet.
It is applicable to all types of nuclear reactor protection systems.
2 Normative References
The following normative documents contain provisions which, through reference in this text, constitute provisions of this standard. For dated references, subsequent amendments (excluding corrections), or revisions, of any of these publications do not apply to this standard. However parties to agreements based on this standard are encouraged to investigate the possibility of applying the most recent editions of the normative documents indicated below. For undated references, the latest edition applies.
GB/T 5204 Periodic Tests and Monitoring of the Safety System of Nuclear Power Plant (GB/T 5204-1994, neq ANSI/IEEE 338:1987)
GB/T 5963 Separation Criteria for Reactor Protection System (GB/T 5963-1995, eqv IEC 60709:1981)
GB/T 7163 Requirements of Reliability Analysis for Nuclear Power Plant Safety System (GB/T 7163-1999, eqv IEEE Std 577:1976)
GB/T 8993 Environmental Conditions and Testing Procedures for Nuclear Instrumentation
GB/T 9225 General Principles of Reliability Analysis for Nuclear Power Plant Safety Systems (GB/T 9225-1999, eqv ANSI/IEEE Std 352:1987)
GB/T 11684 Electromagnetic Environment Conditions and Testing Procedures for Nuclear Instrumentation
GB/T 12505 Specification for Computer Software Configuration Management Plan
GB/T 12727 Nuclear Power Plants - Electrical Equipment of the Safety System - Qualification
GB 13284-1998 Criteria for Safety Systems for Nuclear Power Plants (eqv IEEE Std 603:1991)
GB/T 13625 Seismic Qualification of Safety Class Electrical Equipment for Nuclear Power Plants (GB/T 13625-1992, eqv IEC 60980:1988)
GB/T 13629-1998 Applicable Criteria for Digital Computers in Safety Systems of Nuclear Power Plants (eqv IEEE Std 7-4.3.2:1993)
EJ/T 529-1990 Programmed Digital Computers Important to Safety for Nuclear Power Stations (eqv IEC 60987:1989)
EJ/T 797 Practice for Application of Human Factors Engineering to Systems, Equipment and Facilities of Nuclear Power Plant
EJ/T 1058-1998 Software for Computers in the Safety Systems of Nuclear Power Plants (eqv IEC 60880:1986)
HAD 102/10(1988) Nuclear Power Plants and Related Facilities of Nuclear Power Plant
3 Terms and Definitions
For the purposes of this standard, the following terms and definitions apply.
3.1
Reactor protection system
The system which generates the output signal required for triggering the action of safety actuator and safety system support (auxiliary) facilities for preventing the reactor state from exceeding the specified safety limits or mitigating the consequence from exceeding the safety limit. It includes all equipment (including hardware and software) from sensitive elements to the input terminal of safety actuator or the input terminal of safety system support (auxiliary) facilities.
Note: Reactor protection system includes reactor trip system and engineering safety feature actuation system.
3.2
Reactor trip system
A part of the reactor protection system. It triggers the action of a safety actuator to make the reactor to shut down quickly.
3.3
Engineering safety feature actuation system
A part of the reactor protection system. It triggers the action of a special safety facility to mitigate the consequences of accidents and prevent the leakage of radioactive materials.
3.4
Safety interlock
It allows certain operations that affect the safety of the reactor only when there are specified conditions.
3.5
Safety monitoring assembly
The device used for monitoring the safety of reactor. It generally includes sensitive elements, signal conditioning and/or processing components.
3.6
Safety logic assembly
It is connected to a safety monitoring assembly to achieve predetermined logic functions and send the output signals to one or more safety actuator(s).
3.7
Safety actuator
A device that directly controls the action of the actuator in accordance with instructions from one or more safety logic assembly(ies). For example, emergency trip circuit breaker, valve and pump controller, etc.
3.8
Safe failure
A failure increasing the probability of a safe action in the protection system.
3.9
Unsafe failure
A failure decreasing the probability of a safe action in the protection system.
3.10
Spurious shutdown
Automatic shutdown due to one or more safe failure(s) in the protection system during normal operation of the reactor.
3.11
Protective setpoint
A value predetermined according to safety analysis. When the monitored variable reaches the value, the protection system triggers the action of a safety actuator.
3.12
Operational by-pass
The action and measure for inhibiting some specific functions of the protection system according to the needs of operation.
Foreword i 1 Scope 2 Normative References 3 Terms and Definitions 4 Design Basis 5 Safety Criteria 6 Additional Requirements based on Computer System Bibliography
1 Scope
This standard specifies the basic safety requirements that the nuclear reactor protection system shall meet.
It is applicable to all types of nuclear reactor protection systems.
2 Normative References
The following normative documents contain provisions which, through reference in this text, constitute provisions of this standard. For dated references, subsequent amendments (excluding corrections), or revisions, of any of these publications do not apply to this standard. However parties to agreements based on this standard are encouraged to investigate the possibility of applying the most recent editions of the normative documents indicated below. For undated references, the latest edition applies.
GB/T 5204 Periodic Tests and Monitoring of the Safety System of Nuclear Power Plant (GB/T 5204-1994, neq ANSI/IEEE 338:1987)
GB/T 5963 Separation Criteria for Reactor Protection System (GB/T 5963-1995, eqv IEC 60709:1981)
GB/T 7163 Requirements of Reliability Analysis for Nuclear Power Plant Safety System (GB/T 7163-1999, eqv IEEE Std 577:1976)
GB/T 8993 Environmental Conditions and Testing Procedures for Nuclear Instrumentation
GB/T 9225 General Principles of Reliability Analysis for Nuclear Power Plant Safety Systems (GB/T 9225-1999, eqv ANSI/IEEE Std 352:1987)
GB/T 11684 Electromagnetic Environment Conditions and Testing Procedures for Nuclear Instrumentation
GB/T 12505 Specification for Computer Software Configuration Management Plan
GB/T 12727 Nuclear Power Plants - Electrical Equipment of the Safety System - Qualification
GB 13284-1998 Criteria for Safety Systems for Nuclear Power Plants (eqv IEEE Std 603:1991)
GB/T 13625 Seismic Qualification of Safety Class Electrical Equipment for Nuclear Power Plants (GB/T 13625-1992, eqv IEC 60980:1988)
GB/T 13629-1998 Applicable Criteria for Digital Computers in Safety Systems of Nuclear Power Plants (eqv IEEE Std 7-4.3.2:1993)
EJ/T 529-1990 Programmed Digital Computers Important to Safety for Nuclear Power Stations (eqv IEC 60987:1989)
EJ/T 797 Practice for Application of Human Factors Engineering to Systems, Equipment and Facilities of Nuclear Power Plant
EJ/T 1058-1998 Software for Computers in the Safety Systems of Nuclear Power Plants (eqv IEC 60880:1986)
HAD 102/10(1988) Nuclear Power Plants and Related Facilities of Nuclear Power Plant
3 Terms and Definitions
For the purposes of this standard, the following terms and definitions apply.
3.1
Reactor protection system
The system which generates the output signal required for triggering the action of safety actuator and safety system support (auxiliary) facilities for preventing the reactor state from exceeding the specified safety limits or mitigating the consequence from exceeding the safety limit. It includes all equipment (including hardware and software) from sensitive elements to the input terminal of safety actuator or the input terminal of safety system support (auxiliary) facilities.
Note: Reactor protection system includes reactor trip system and engineering safety feature actuation system.
3.2
Reactor trip system
A part of the reactor protection system. It triggers the action of a safety actuator to make the reactor to shut down quickly.
3.3
Engineering safety feature actuation system
A part of the reactor protection system. It triggers the action of a special safety facility to mitigate the consequences of accidents and prevent the leakage of radioactive materials.
3.4
Safety interlock
It allows certain operations that affect the safety of the reactor only when there are specified conditions.
3.5
Safety monitoring assembly
The device used for monitoring the safety of reactor. It generally includes sensitive elements, signal conditioning and/or processing components.
3.6
Safety logic assembly
It is connected to a safety monitoring assembly to achieve predetermined logic functions and send the output signals to one or more safety actuator(s).
3.7
Safety actuator
A device that directly controls the action of the actuator in accordance with instructions from one or more safety logic assembly(ies). For example, emergency trip circuit breaker, valve and pump controller, etc.
3.8
Safe failure
A failure increasing the probability of a safe action in the protection system.
3.9
Unsafe failure
A failure decreasing the probability of a safe action in the protection system.
3.10
Spurious shutdown
Automatic shutdown due to one or more safe failure(s) in the protection system during normal operation of the reactor.
3.11
Protective setpoint
A value predetermined according to safety analysis. When the monitored variable reaches the value, the protection system triggers the action of a safety actuator.
3.12
Operational by-pass
The action and measure for inhibiting some specific functions of the protection system according to the needs of operation.
Contents of GB/T 4083-2005
Foreword i
1 Scope
2 Normative References
3 Terms and Definitions
4 Design Basis
5 Safety Criteria
6 Additional Requirements based on Computer System
Bibliography
