CNSC welcomes feedback on any regulatory document at any time REGDOC- supersedes RD, Design of New Nuclear Power. CNSC has issued its Fukushima report – posted on the CNSC website on that the design intent complies with CNSC design requirements (RD, RD-. Re: The Approvals Process for New Reactors in Canada – RD & RD ( CNSC) request for feedback on the comments received on the.
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In addition, the means for adding or modifying the chemical constituents of fluid streams are specified.
The choice of construction material is commensurate with the designed service life and potential life extension of the plant. The design considers all natural and cnsx external events that may be linked with significant radiological risk.
Design for reliability includes consideration of mission times for SSCs important to safety.
SSCs important to safety are designed and located in a manner that minimizes the probability and effects of fires and explosions caused by external or internal events. The design of the SCR ensures that appropriate lighting levels and thermal environment are maintained, and noise levels align with applicable standards and codes.
RD-337: Design of New Nuclear Power Plants
The plant design considers the role of structures, pathways, equipment, and instrumentation in providing detection, delay, and response to threats. Where possible, this is achieved without operator intervention.
The basis for the safety assessment is the data derived from the safety analysis, previous operational experience, results of supporting research, and proven engineering practices. Examples are wires, transistors, integrated circuits, motors, relays, solenoids, pipes, fittings, pumps, tanks and valves, etc.
Probabilistic safety assessment A comprehensive and integrated assessment of the safety of the nuclear power plant that, by considering the initial plant state and the probability, progression, and consequences of equipment failures and operator response, derives numerical estimates of a consistent measure of the safety of the plant.
RD Design of New Nuclear Power Plants – Canadian Nuclear Safety Commission
The preliminary safety analysis assists in the establishment of the design-basis requirements for the items important to safety, and demonstrates whether the plant design meets applicable expectations. All pressure-retaining SSCs of the reactor coolant system and auxiliaries are designed with an appropriate safety margin to ensure that the pressure boundary will not be breached, and that fuel design limits will not be exceeded in normal operation, AOOs, or DBA conditions.
Stationary equipment is provided for monitoring the effluents prior to or during discharge to the environment. The reactor coolant system is fitted with isolation devices to limit any loss of radioactive coolant outside containment. Leak detection is an acceptable method when the SSC is leak-before-break qualified. Fnsc information provided herein is intended to facilitate high quality design, and consistency with modern international codes and standards, for new water-cooled NPPs.
The design provides for a main control room MCR from which the plant can be safely operated, and from which measures can be taken to maintain the plant rd3-37 a safe state or to bring it back into such a state after the onset of AOOs, DBAs, and, to the extent practicable, following BDBAs. This approach includes verification at each step of the development process to demonstrate that the respective product is correct, and validation to demonstrate that the resulting computer-based system or equipment meets its td-337 and performance requirements.
A systematic approach is used to assess the potential bio-physical environmental effects of the NPP on the environment, and the effects of the environment on the NPP. The containment structure and the rv-337 and components affecting the leak tightness of the containment system are designed to allow leak rate testing:. Correct operation of the EHRS equipment following an accident is not dependent on power supplies from the electrical grid or from the turbine generators associated with any reactor unit that is located on the same site as the reactor involved in the accident.
All closed piping service systems have at least one single isolation valve on each line penetrating the containment, with the valve being located outside of, but as close as practicable to, the containment structure. The design also includes the capability to trend and automatically record measurement of any derived parameters that are important to safety. With the defence-in-depth approach, if a failure were to occur it will be detected and compensation made, or it would be corrected.
The human factors engineering program also facilitates the interface between the operating personnel and the plant by promoting attention to plant layout and procedures, maintenance, inspection, training, and the application of ergonomic principles to the design of working areas and working environments.
The design authority identifies credible BDBAs, based on operational experience, engineering judgment, and the results of analysis and research. The design considers the time allowed for each equipment outage and the respective response actions.
All water-cooled nuclear power reactors are to be equipped with an emergency core cooling system ECCS. There should be sufficient structural integrity to protect important systems. The expectations for reactor and fuel assembly design apply in the event of changes in fuel management strategy or in operating conditions over the lifetime of the plant. Containment also prevents uncontrolled releases of radioactivity after this period.
Severe accident A beyond design basis accident that involves significant core degradation. Unintended actions and failure of passive components are considered as two of the modes of failure of a safety group.
The design also identifies SSCs to which design limits are applicable. Systems designed for this purpose are considered part of the containment system, and are capable of:. The design also considers the needs for related testing when specifying the commissioning requirements for the plant.
The design applies the principle that plant states that could result in high radiation doses or radioactive releases have a very low frequency of occurrence, and plant states with significant frequency of occurrence have only minimal, if any, potential radiological consequences.
The NPP design draws on operational experience that has been gained in the nuclear industry, and on the results of relevant research programs. Consideration is given to the characteristics and importance of the isolation and its reliability targets. The aim of the fifth level of defence is to mitigate the radiological consequences of potential releases of radioactive materials that may result from accident conditions. The design includes provision of instrumentation to monitor plant variables and systems over the respective ranges for normal operation, AOOs, DBAs, and BDBAs, in order to ensure that adequate information can be obtained on plant status.
Achievement of defence-in-depth level three calls for provision of inherent safety features, fail safe design, engineered design features, and procedures that minimize the consequences of DBAs.
The design includes instrumentation for monitoring seismic activity at the site for the life of the plant. The components of the reactor coolant pressure boundary are designed, manufactured, and arranged in a manner that permits adequate inspections and tests of the boundary throughout the lifetime of the plant. The acceptance criteria for the fuel for DBAs are consistent with these expectations.
The design is expected to provide for ready and reliable detection of any significant breach of the containment envelope.
Document History of REGDOC-2.5.2, Design of Reactor Facilities: Nuclear Power Plants
Relevant and proven systematic analysis techniques are used to address human factors issues within the design process. In addition, the design limits the amount of activated material and its build-up.
The set of design basis accidents sets the boundary conditions according to which SSCs important to safety are designed. The design provides physical features such as protection against design basis threats DBTsin accordance with the requirements of the Nuclear Security Regulations.
The extent of trip coverage provided by all available parameters is documented for the entire spectrum of failures for each set of PIEs. New designs are tested before being brought into service, and rd-37 then monitored in service to cjsc that the expected behaviour is achieved.