- •Iaea safety standards
- •Objective
- •Structure
- •Relationship to other standards
- •2. Management systems for I&c design
- •2.8. The management systems for development of I&c systems should comply with the recommendations of Safety Guides gs-g-3.1, Ref. [5] and gs-g-3.5, Ref. [6].
- •Generic management system processes
- •Configuration management
- •2.15. All I&c configuration items and their associated configuration documents should be designated, given a unique identification, and placed under configuration control.
- •2.31. Insights gained from probabilistic safety assessments (psAs) should be considered in the design of I&c systems.
- •Documentation
- •2.33. Before I&c systems are declared operable their documentation should be complete and should reflect the as-built configuration.
- •2.34. I&c documentation should:
- •2.36. I&c documents should be grouped according to their primary or secondary role in the design process.
- •2.38. Documentation for I&c systems and components should, as a minimum, cover the following topics:
- •3. Design bases
- •Inputs to I&c design bases
- •Identification of I&c functions
- •3.4. The required functions of the I&c systems should be determined as part of the nuclear power plant design process.
- •Content of I&c design bases
- •3.7. The overall I&c system architecture and each I&c system should have a design basis.
- •3.9. The I&c systems required for the safety of the plant should be identified systematically.
- •3.10. I&c system design bases should specify the following:
- •3.12. In addition to the above the design basis for the reactor protection system should specify the following:
- •Variables and states that must be displayed so that the operators can confirm the operation of protective system functions;
- •4. Guidance for overall I&c system architecture architectural design
- •4.3. The overall I&c architecture should:
- •4.4. The inputs to the overall I&c architecture design process should refer to the plant safety design basis documents, which should provide the following information:
- •Defence in depth
- •4.28. When diverse I&c systems are provided to meet requirements for defence-in-depth, the diverse systems should not both be subject to the same errors in design or fabrication.
- •5. Safety classification of I&c functions, systems, and equipment
- •6. Life cycle activities
- •Process implementation
- •Verification that the effects of automatic control system failures will not exceed the acceptance criteria established for anticipated operational occurrences.
- •6.58. The I&c architecture should be designed to fully satisfy the system requirements, including system interfaces and non-functional requirements (e.G., performance and reliability).
- •6.109. The benefits of changes should be weighed against potential negative safety consequences and this assessment documented as part of the justification for the changes.
- •Design for reliability
- •Single failure criterion
- •7.10. Each safety group should perform all actions required to respond to a pie in the presence of:
- •7.15. Non-compliance with the single failure criterion should be exceptional and clearly justified in the safety analysis.
- •7.19. I&c systems should be redundant to the degree needed to meet design basis reliability requirements.
- •Independence
- •7.27. When isolation devices are used between systems of different safety importance, they should be a part of the system of higher importance.
- •7.29. The adequacy of design features provided to meet the independence recommendations should be justified. Physical separation
- •7.31. Items that are part of safety systems should be physically separated from items of lower safety classification.
- •7.32. Redundant items of safety systems should be physically separated from each other.
- •Electrical isolation
- •Diversity
- •7.49. The decision to use diversity or not use diversity should be justified.
- •7.50. Where diversity is provided to cope with ccf several types of diversity should be used.
- •7.51. Where diversity is provided the choice of the types of diversity used should be justified.
- •Failure modes
- •7.57. The failure modes of I&c components should be known and documented.
- •7.60. Failures of I&c components should be detectable by periodic testing or self-revealed by alarm or anomalous indication.
- •7.73. Analysis that is part of the evidence of equipment qualification should include a justification of the methods, theories and assumptions used.
- •7.75. Traceability should be established between each installed system and component important to safety and the applicable evidence of qualification.
- •Suitability and correctness
- •7.81. The equipment qualification program should demonstrate that the as-built I&c systems and installed components correctly implement the qualified design.
- •7.90. Environmental qualification of safety components that must operate in harsh environments should include type testing.
- •7.102. Detailed emc requirements should be determined for safety systems and components and their compliance with the requirements demonstrated.
- •7.105. Equipment and systems, including associated cables, should be designed and installed to withstand the electromagnetic environment in which they are located.
- •7.109. Limits on radiated and conducted electromagnetic emissions should be established for all plant equipment.
- •7.112. The equipment qualification program should show that electromagnetic emissions of plant equipment are within the defined limits.
- •7.114. Instrumentation cables should have twisting and shielding sufficient to minimize interference from electromagnetic and electrostatic interference.
- •Design to cope with ageing
- •7.119. Ageing mechanisms that could significantly affect I&c components and means for following the effects of these mechanisms should be identified during design.
- •7.122. Maintenance programs should include activities to identify any trend towards degradation (ageing) that could result in the loss of operability of equipment.
- •Control of access to systems important to safety
- •7.130. Access to equipment in I&c systems should be limited to prevent unauthorized access and to reduce the possibility of error.
- •Testing and testability during operation
- •Test provisions
- •7.150. Arrangements for testing should neither compromise the independence of safety systems nor introduce the potential for common cause failures.
- •Test interfaces
- •7.153. Provisions for testing I&c systems and components should:
- •7.154. Where equipment to be tested is located in hazardous areas, facilities should be provided to allow testing from outside the hazardous area.
- •7.164. The test program should define processes for periodic tests and calibration of systems that:
- •Individually test each sensor, to the extent practicable.
- •7.165. In addition to the recommendations of paragraph 7.164, the processes defined for periodic tests and calibration of safety systems should:
- •Independently confirm the functional and performance requirements of each channel of sense, command, execute, and support functions;
- •Include as much of the function under test as practical (including sensors and actuators) without jeopardizing continued normal plant operation;
- •Maintainability
- •7.169. The design of I&c systems should include maintenance plans for all systems and components.
- •Setpoints
- •7.185. Trip setpoints used to initiate safety actions should be selected to ensure that required mitigating actions occur before the monitored variable reaches the analytical limit.
- •Operational identification of items important to safety
- •7.186. A consistent and coherent method of naming and identifying all I&c components should be determined and followed throughout the design, installation and, operation phases of the plant.
- •7.190. I&c components in the plant should be marked with their identifying information.
- •8.4. To the extent practicable, the plant conditions of concern should be monitored by direct measurement rather than being inferred from indirect measurements.
- •8.17. Means should also be provided to manually initiate the mechanical safety systems and the individual components necessary to initiate and control performance of their safety functions.
- •Digital computer systems and digital equipment
- •8.68. Specific skilled staff should be available during operation to allow controlled software and configuration data changes to be made when necessary to computer based systems.
- •8.91. Data received and data transmitted should be stored in separate, pre-determined memory locations.
- •8.154. Tools should be used to support all aspects of the I&c life cycle where benefits result through their use and where tools are available.
- •8.173. Confirmation of the suitability and correctness of industrial digital devices for their intended functions should produce evidence:
- •V&V at each stage of development for the final product;
- •9.4. The I&c system should allow the operator in the control room to initiate or take manual control of each function necessary to control the plant and maintain safety.
- •9.21. Instrumentation performing the functions given in 9.20 items a, b, and c should be classified as safety systems.
- •9.32. The main control room, the supplementary control room, and the Emergency Control Centre should have at least two diverse communications links with:
- •9.42. The Human System Interface (hmi) design should retain positive features and avoid hfe issues and problems of previous designs.
- •9.57. Where hmi stations are distributed, plant staff should have means to access these different locations in a safe and timely manner.
- •10.4. Development of software for systems should follow a previously defined life cycle, be duly documented and include thorough verification and validation. (See Chapter 6.)
- •10.49. Coding rules should be prescribed and adherence verified.
- •10.72. Verification should include the following techniques:
- •Software tools
- •Glossary
- •Annex I defense in depth in I&c systems
- •Annex II traceability to previouse I&c safety guides
- •Annex III bibliography of supporting international standards
9.4. The I&c system should allow the operator in the control room to initiate or take manual control of each function necessary to control the plant and maintain safety.
9.5. There should be sufficient displays in the control room to monitor all functions important to safety.
9.6. Safety classified controls should be provided for implementing emergency operating procedures and to execute manually controlled safety actions.
9.7. Non-safety controls may also be provided for safety items and non-safety displays may be provided safety instrumentation channels if the independence recommendations of paragraphs 4.11-4.15, 7.21-7.47, 8.50-8.52, and 8.110 are met.
9.8. If part of a system has failed or been intentionally made inoperative, this condition should be displayed in the control room.
9.9. Changes in the status of safety systems should be annunciated, and the status should be indicated in the control room.
9.10. Change in status needing alarms might include deviations from normal operational limits, or loss of availability of safety systems, or unavailability of standby equipment due to failure or maintenance and testing.
Supplementary control room
9.11. SSR 2/1 requirement 66 states:
Instrumentation and control equipment shall be kept available, preferably at a single location (supplementary control room) that is physically independent, electrically and functionally separate from the control room of the nuclear power plant, The supplementary control room shall be such that the reactor can be placed and maintained in a shut down state, residual heat can be removed, and the essential plant variables can be monitored if there is a loss of ability to perform these fundamental safety functions in the control room.
9.12. Sufficient controls, indications, alarms and displays should be provided in the supplementary control rooms to bring the plant to a safe state, to provide assurance that a safe state has been reached and maintained, and to provide operators with information on the status of the plant and the trends in key plant parameters.
9.13. Where it is impractical to provide controls for special plant conditions in the supplementary control room, the controls at local control points may be used.
9.14. Suitable provision outside the main control room should be made for transferring priority control to a new location and isolating the equipment in the main control room whenever the main control room is abandoned.
9.15. The design of the main control room and supplementary control room should be such that no fire or PIE can simultaneously affect both rooms to the extent that maintenance of the fundamental safety functions cannot be ensured.
ACCIDENT MONITORING
9.16. SSR 2/1 Paragraph 6.31 states
Instrumentation and recording equipment shall be provided to ensure that essential information is available for monitoring the status of essential equipment and the course of accidents; for predicting the locations of release and amount of radioactive material that could be released from the locations that are so intended in the design, and for post-accident analysis.
9.17. Information displays for monitoring accident conditions in the plant should be provided in the main control room.
9.18. The supplementary control room should contain information displays for monitoring plan conditions as needed to support response to events which may result from situations that necessitate evacuation of the main control room.
9.19. The collection of displays for monitoring accident conditions are usually called an “Accident Monitoring System” or a “Post Accident Monitoring System.” These displays may be provided as part of another system or may be a collection of individual instrument channels.
9.20. The accident monitoring system should indicate the values of primary plant variables needed by plant operators to, under accident conditions:
Take pre-planned manual actions to bring the plant to a safe state;
Determine if the fundamental safety functions are being achieved;
Determine the potential for or actual breach of the barriers to fission product release (e.g., fuel clad, reactor coolant pressure boundary, and containment);
Determine the status and performance of plant systems supporting systems necessary to mitigate a design basis accident and bring the plant to a safe state; and
Determine the need to initiate action to protect the public from radiation release.