CNSC RD-337 PDF

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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It establishes a set of comprehensive design expectations that are risk-informed and align with accepted international codes and practices. This document provides criteria pertaining to the safe design of new water-cooled NPPs, and offers examples of optimal design characteristics where applicable. All aspects of the design are taken into account, and multiple levels of defence are promoted in design considerations.

To the extent practicable, the guidance provided herein is technology-neutral with respect to water-cooled reactors.

RD-337: Design of New Nuclear Power Plants

Designand the adaptation of those principles to align with Canadian expectations. Similar to NS-R-1, RD considers all licensing phases, because information from the design process feeds into the processes for reviewing an application for a Licence to Construct an NPP, and other licence applications.

Nothing contained in this document is to be construed as relieving any applicant or licensee from requirements associated with conventional codes and standards. The purpose of this regulatory document is to set out the expectations of the Canadian Nuclear Safety Commission CNSC with respect to the design of new water-cooled nuclear power plants NPPs or plants. This document sets out CNSC expectations with respect to the design of new water-cooled NPPs, and provides examples of optimal design characteristics.

The information provided herein is intended to facilitate high quality design, and consistency with modern international codes and standards, rrd-337 new water-cooled NPPs. It is recognized that specific technologies may use alternative approaches. If a design other than a water-cooled reactor is to be considered for licensing in Canada, the design is subject to the safety objectives, high level safety concepts and safety management expectations associated with this regulatory document.

However, CNSC review of such a design will be undertaken on a case by case basis. Conventional industrial safety is addressed only from a high-level perspective, with a cncs on design considerations that are related to nuclear safety. To the extent practicable, this document is technology-neutral with respect to water-cooled reactors, and includes direction concerning:.

Designand cnsv adaptation of those principles to align with Canadian practices. The scope of NS-R-1 has been expanded to address the interfaces between NPP design and other topics, such as environmental protection, radiation protection, ageing, human factors, security, safeguards, transportation, and accident and emergency response planning.

This objective relies on the establishment and maintenance of effective defences against radiological hazards in NPPs. The general cbsc safety objective is supported by two complementary cnwc objectives dealing with radiation protection and with the technical aspects of the design. The technical safety objective is interdependent with administrative and procedural rd-373 that are taken to ensure defence against hazards due dd-337 ionizing radiation.

The radiation protection objective is to provide that during normal operation, or during anticipated operational occurrences, radiation exposures within the NPP or due to any planned release of radioactive material from the NPP are kept below prescribed limits and as low as reasonably achievable ALARA.

The technical safety objectives are to provide all reasonably practicable measures to prevent accidents in the NPP, and to mitigate the consequences of accidents if cbsc do occur.

R-337 takes into account all possible accidents considered in the design, including those of very low probability. With achievement of these objectives, any radiological consequences should be minor and below prescribed limits, and the likelihood dnsc accidents with serious radiological consequences is expected to be extremely low.

The NSCA and the technical safety objectives provide the basis for the following criteria and goals:. Safety analyses chsc performed to confirm that these criteria and goals are met, to demonstrate effectiveness of measures for preventing accidents, and mitigating radiological consequences of accidents if they do occur. The committed whole-body dose for average members of the critical groups who are most at risk, at or beyond the site boundary is calculated in r-337 deterministic safety analysis for a period of 30 days after the analyzed event.

A limit is placed on the societal risks posed by nuclear power plant operation. For this purpose, the following two qualitative safety goals have been established:. Quantitative Application of the Safety Goals. For practical application, quantitative safety goals are established to achieve the rd-37 of the qualitative safety goals. The three quantitative safety goals are:.

A core damage accident results from a postulated initiating event PIE followed by failure of one or more safety system s or safety support system s.

They also represent measures of risk to society and to the environment due to the operation of a nuclear power plant. The sum of frequencies of all event sequences that can lead to significant core degradation is less than 10 -5 per reactor year. The sum of frequencies of all event sequences that can lead to a release to the environment of more than 10 15 becquerel of iodine is less than 10 -5 per reactor year.

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RD Design of New Nuclear Power Plants – Canadian Nuclear Safety Commission

A greater release may require temporary evacuation of the local population. The sum of frequencies of cmsc event sequences that can lead to a release to the environment of more than 10 14 becquerel of cesium is less cmsc 10 -6 per reactor year.

A greater release may require long term relocation of the local population. To demonstrate achievement of the safety objectives, a comprehensive hazard analysis, a deterministic safety analysis, and a probabilistic safety assessment are carried out. These analyses identify all sources of exposure, in order r-337 evaluate potential radiation doses to workers at the plant and to the public, and to evaluate potential effects on the environment.

Based on these analyses, the capability of the design to withstand postulated initiating events PIEs and accidents can be confirmed, the effectiveness of the items important to safety can be demonstrated, and requirements for emergency fd-337 can be established. The results of the safety analyses are fed back into the design.

The design includes provisions to limit radiation exposure in normal operation and AOOs to ALARA levels, and to minimize the likelihood of an accident that could lead cns the loss of normal control of the source of radiation.

However, given that there is a remaining probability that an accident rd-37 occur, measures are taken to mitigate the radiological consequences of accidents. 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 concept of defence-in-depth is applied to all organizational, behavioural, and design-related safety and security activities to ensure that they are subject to overlapping provisions.

With the defence-in-depth approach, if a failure were to occur it will be detected and compensation re-337, or it would be corrected. This concept is applied throughout the design process and operation xnsc the plant to provide a series of levels of defence aimed at preventing accidents, and ensuring appropriate protection in the event that prevention fails. The design provides all five levels of defence during normal operation; however, some relaxations may be specified for certain shutdown states.

These levels are introduced in general terms below, and are discussed in greater detail in subsection 6. The fd-337 of the first level of defence is to prevent deviations from normal operation, and to prevent failures of systems, structures, and components SSCs. The aim of the second level of defence is to detect and intercept deviations from normal operation in order to vnsc AOOs from escalating to accident conditions, and to rx-337 the plant to a state of normal operation.

The aim of the third level of defence is to minimize the consequences of accidents by providing inherent safety features, fail-safe design, additional equipment, and mitigating procedures. The aim of the fourth level of defence is to ensure that radioactive releases caused by severe accidents are kept as low as practicable.

The aim of the fifth cnec of defence is to mitigate the radiological consequences of potential releases of radioactive materials that may result from accident conditions. An important aspect of implementing defence-in-depth in the NPP design is the provision of a series of physical barriers to confine radioactive material at specified locations.

Operational limits rd-3337 conditions OLCs are the set of limits and conditions that can be monitored by or on behalf of the operator, and that can be controlled by the operator.

The OLCs are established to ensure that plants operate in accordance with design assumptions and intent parameters and componentsand include the limits within which the facility has been shown to be safe. The Rv-337 are documented in a manner that is readily accessible for control room personnel, with the roles csc responsibilities clearly identified. Some OLCs may include combinations ncsc automatic functions and actions by personnel.

The basis on which the OLCs are derived will be readily available in order to facilitate the ability of plant personnel to interpret, observe, and apply the OLCs. The design process is carried out by technically qualified and appropriately trained staff at all levels, and includes such management arrangements as:. During the design phase, formal design authority typically rests with the organization that has overall responsibility for the design.

Prior to plant start-up, this authority may be transferred to the operating organization. The design authority may assign responsibility for the design of specific parts of the plant to other organizations, known as responsible designers.

The tasks and functions of the design authority and any responsible designer need to be established in formal documentation; however, the overall responsibility remains with the design authority. The applicant confirms that the design authority has achieved the following objectives during the design phase:. A quality assurance program is established as part of the overall management arrangements by which the plant will function to achieve objectives.

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With respect to the plant design, this includes identifying all performance and assessment parameters for the design, as well as detailed plans for csc SSC to ensure consistent quality of the design and the selected components.

The quality assurance program is such that the initial design, and any subsequent change or safety improvement, is carried out in accordance with established procedures that call on appropriate standards and codes, and that incorporate applicable requirements and design bases. Appropriate quality assurance also facilitates identification and control of design interfaces. The adequacy of the design, including design tools and design inputs and outputs, are verified or validated by individuals or groups that are independent from those who originally performed the work.

Verifications, validations, and approvals are completed before the detailed design is implemented. The design authority identifies the modern standards and codes that will be used for the plant design, and evaluates those standards and codes for applicability, adequacy, and sufficiency to the design of SSCs important to safety.

Where needed, codes and standards may be supplemented or modified to ensure that the final quality of the design is commensurate with the necessary safety functions.

SSCs important to safety are of proven designs, and are designed according to the standards and codes identified for the NPP. Where a new SSC design, feature, or engineering practice is introduced, r-337 safety is proven by a combination of supporting research and development programs, and by examination of relevant experience from similar applications.

An adequate qualification program is established to verify that the new design meets all applicable safety expectations. New designs are tested before being brought into service, and are then monitored in service to verify that the expected behaviour is achieved.

The design authority establishes an adequate qualification program to verify that the new design meets all applicable safety design requirements. In the selection of equipment, due attention is given to spurious operation and to unsafe failure modes e. Where the design has to accommodate an SSC failure, preference is given to equipment that exhibits known and predictable modes of failure, and that facilitates repair or replacement.

The NPP design draws on operational experience that has been gained in the nuclear industry, and on the results of relevant research programs.

Safety assessment is a systematic process applied throughout the design phase to ensure that the design meets all relevant safety requirements. This includes the requirements set by the operating organization and by regulatory authorities. 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.

The safety assessment is part of the design process, with iteration between the design and rr-337, and increases in scope and level of detail as the design process progresses. Before the design is submitted, an independent peer review of the safety assessment is conducted by individuals or groups separate from those carrying out the design. Safety assessment documentation identifies those aspects of operation, maintenance, and management that are important to safety.

This documentation is maintained dr-337 a dynamic suite of documents to reflect changes in design ccnsc the plant evolves. Safety assessment documentation is presented clearly and concisely, in a logical and understandable format, and will be made readily accessible to designers, operators, and the CNSC. Defence-in-depth is achieved at the design phase through application of design provisions specific to the five levels of defence.

Achievement of defence-in-depth level one calls for conservative design and high-quality construction to provide confidence that plant failures and deviations from normal operations are minimized and accidents are prevented.

This entails careful attention to selection of appropriate design codes and materials, design procedures, equipment qualification, control of component fabrication and plant construction, and use of operational experience. Defence-in-depth level two is achieved by controlling plant behaviour during and following a PIE using both inherent and engineered cnsx features to minimize or exclude uncontrolled transients to the extent possible. Achievement of defence-in-depth level three calls for rd3-37 of inherent safety features, fail safe design, engineered design features, and procedures that minimize the consequences of DBAs.

These provisions are capable of leading the plant first to a controlled state, and then to a safe shutdown state, and maintaining at least one barrier for the confinement of radioactive material. Automatic activation of the engineered design features minimizes the need for operator actions in rd-3377 early phase of a DBA.