Asce 7.22

[Saurabh Cloudas]

This regulatory document is part of the CNSC's Physical Design series of regulatory documents, which also covers: design of uranium mines and mills; design of fixed radiography installations; design of nuclear substance laboratories and nuclear medicine rooms; and exposure devices. The full list of regulatory document series is included in the back of this document and can be found on the CNSC's website.

REGDOC-2.5.2, Design of Reactor Facilities: Nuclear Power Plants, sets out requirements and guidance for new licence applications for water-cooled nuclear power plants (NPPs or plants). It establishes a set of comprehensive design requirements and guidance 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. All aspects of the design are taken into account, and multiple levels of defence are promoted in design considerations. To the extent practicable, the requirements and guidance provided herein are technology-neutral with respect to water-cooled reactors. An applicant or licensee may put forward a case to demonstrate that the intent of a requirement is addressed by other means and demonstrated with supportable evidence.

To a large degree, this regulatory document represents the CNSC's adoption of the principles set forth by the International Atomic Energy Agency in SSR-2/1, Safety of Nuclear Power Plants: Design as adapted to align with Canadian requirements.

This regulatory document considers all licensing phases, as information from the design process feeds into the processes for reviewing an application for a licence to construct an NPP, and other licence applications.

This document is intended to form part of the licensing basis for a regulated facility or activity within the stated scope of the document. It is intended for inclusion in licences as either part of the conditions and safety and control measures in a licence, or as part of the safety and control measures to be described in a licence application and the documents needed to support that application.

Guidance contained in this document exists to inform the applicant, to elaborate further on requirements, or to provide direction to licensees and applicants on how to meet requirements. It also provides more information about how CNSC staff evaluate specific problems or data during their review of licence applications. Licensees are expected to review and consider this guidance; if they choose not to follow it, they should explain how their selected approach still meets regulatory requirements.

Important note: Where referenced in a licence either directly or indirectly (such as through licensee-referenced documents), this document is part of the licensing basis for a regulated facility or activity.

The licensing basis sets the boundary conditions for acceptable performance at a regulated facility or activity, and establishes the basis for the CNSC's compliance program for that regulated facility or activity.

Where this document is part of the licensing basis, the word "shall" is used to express a requirement to be satisfied by the licensee or licence applicant. "Should" is used to express guidance or that which is advised. "May" is used to express an option or that which is advised or permissible within the limits of this regulatory document. "Can" is used to express possibility or capability.

Nothing contained in this document is to be construed as relieving any licensee from any other pertinent requirements. It is the licensee's responsibility to identify and comply with all applicable regulations and licence conditions.

This regulatory document sets out the requirements of the Canadian Nuclear Safety Commission (CNSC) for the design of new water-cooled nuclear power plants (NPPs, or plants). It establishes a set of comprehensive design requirements and guidance that are risk-informed and align with accepted national and international codes and practices.

This regulatory document deals with a wide variety of topics related to the design of new NPPs. To the extent practicable, this document is technology-neutral with respect to water-cooled reactors, and includes requirements and guidance for:

To a large degree, this document represents the CNSC's adoption of the principles set forth in the International Atomic Energy Agency (IAEA) document SSR-2/1, Safety of Nuclear Power Plants: Design, and the adaptation of those principles to align with Canadian practices.

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 requirements associated with this regulatory document. However, the CNSC's review of such a design will be undertaken on a case-by-case basis.

Four common plant states are defined: normal operation; anticipated operational occurrence (AOO); design-basis accident (DBA); and beyond-design-basis accident (BDBA). This document also introduces the plant state "design extension conditions" (DECs), as a subset of BDBAs that are considered in the plant design.

In support of the NSCA and its associated regulations, the CNSC endorses the objective established by the IAEA that NPPs be designed and operated in a manner that will protect individuals, society and the environment from harm. This objective relies on the establishment and maintenance of effective defences against radiological hazards in NPPs.

The general nuclear safety objective is supported by three complementary safety objectives, which deal with radiation protection, the technical aspects of the design, and environmental protection. The technical safety objective is interdependent with administrative and procedural measures that are taken to ensure defence against hazards due to 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 they do occur. This takes into account all possible accidents considered in the design, including those of very low probability.

The environmental protection objective is to provide all reasonably practical mitigation measures to protect the environment during the operation of an NPP and to mitigate the consequences of an accident.

Safety analyses shall be 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 acceptance criteria for normal operations are provided in section 6.4.

The committed whole-body dose for average members of the critical groups who are most at risk, at or beyond the site boundary, shall be calculated in the deterministic safety analysis for a period of 30 days after the analyzed event.

This dose shall be less than or equal to the dose acceptance criteria of:

The values adopted for the dose acceptance criteria for AOOs and DBAs are consistent with accepted international practices, and take into account the recommendations of the IAEA and the International Commission on Radiological Protection.

Individual members of the public shall be provided a level of protection from the consequences of NPP operation, such that there is no significant additional risk to the life and health of individuals.

Societal risks to life and health from NPP operation shall be comparable to or less than the risks of generating electricity by viable competing technologies, and shall not significantly add to other societal risks.

A core damage accident results from a postulated initiating event (PIE) followed by the failure of one or more safety system(s) or safety support system(s). Core damage frequency is a measure of the plant's accident prevention capabilities.

Small release frequency and large release frequency are measures of the plant's accident mitigation capabilities. They also represent measures of risk to society and to the environment due to the operation of an NPP.

The sum of frequencies of all event sequences that can lead to a release to the environment of more than 1015 becquerels of iodine-131 shall be less than 10-5 per reactor year. A greater release may require temporary evacuation of the local population.

The sum of frequencies of all event sequences that can lead to a release to the environment of more than 1014 becquerels of cesium-137 shall be less than 10-6 per reactor year. A greater release may require long term relocation of the local population

A comprehensive probabilistic safety assessment (PSA) considers the probability, progression and consequences of equipment failures or transient conditions, to derive numerical estimates for the safety of the plant. Core damage frequency is determined by a Level 1 PSA, which identifies and quantifies the sequence of events that may lead to significant core degradation. The small release frequency and large release frequency are determined by a Level 2 PSA, which starts from the results of a Level 1 PSA, analyzes the containment behaviour, evaluates the radionuclides released from the failed fuel, and quantifies the releases to the environment. An exemption for performing a Level 2 PSA is granted if it is shown that core damage frequency in the Level 1 PSA is sufficiency low (i.e., less than the large release frequency limit).

Calculations of the safety goals include all internal and external events as per REGDOC-2.4.2, Probabilistic Safety Assessment (PSA) for Nuclear Power Plants. However, aggregation of internal event and other hazard risk metrics performed through simple addition to demonstrate that the risk metrics (core damage frequency, small release frequency and large release frequency) are not exceeded might not be appropriate. It is recognized that when the risk metrics for external events are conservatively estimated, their summation with the risk metrics for internal events can lead to misinterpretation. Should the aggregated total exceed the safety goals, conclusions should not be derived from the aggregated total until the scope of the conservative bias in the other hazards is investigated.

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