Language selection

Government of Canada / Gouvernement du Canada

Search

Research Support Summaries 2024-2025

Table of contents

RSP-543.3 - SITEX Network Corporate Membership

The CNSC joined the Sustainable network for Independent Technical EXpertise on radioactive waste management (SITEX_Network) under a 5 year corporate membership in 2020.

The objective of the network is to enhance cooperation at the international level to achieve independent, high quality expertise in the field of radioactive waste management. The network operates independently from waste producers and organizations implementing waste management programs with the aim of supporting nuclear regulatory authorities and civil society. The network includes civil society organizations and involves discussions on initiating early engagement during the planning process for deep geological repositories.

Participating in the network allowed the CNSC to access knowledge and lessons learned from European engagement activities with the public and civil society organizations. The network provided linkages to technical support organizations who participated in the European Joint Programme on Radioactive Waste Management.

RSP-656.2 - Licensing of safety critical software for nuclear reactors - Common position of international nuclear regulators and authorized technical support organizations

In Canada, regulatory document REGDOC-2.5.2, Design of Reactor Facilities: Nuclear Power Plants, is the Canadian Nuclear Safety Commission document focused on the regulation of the design of nuclear power plants. REGDOC-2.5.2 provides both requirements and guidance regarding design.  However, there is limited guidance regarding how to meet the requirements when computer-based systems are used to implement safety functions. In order to develop more comprehensive guidance for reviewing the safety critical software, CNSC joined the Regulator Task Force on Safety Critical Software with other regulators, including the major European regulators and US NRC.

This project is to develop up-to-date practical guidance in the regulatory review of safety critical software. The project report is intended to be useful:

  • to coordinate regulators’ and safety experts’ technical viewpoints in licensing practices, or design and revision of guidelines;
  • as a reference in safety cases and demonstrations of safety of software-based systems;
  • as guidance for manufacturers, I&C suppliers, designers and potential licensees on the international market, for creating new products, issuing bids and developing new applications.

The project document is the result of the work of a group of regulator and safety authorities’ experts through periodically organized technical meetings. The major result of the work is the identification of consensus and common technical positions on a set of important licensing issues raised by the design and operation of computer-based systems used in nuclear power plants for the implementation of safety functions. The purpose is to introduce greater consistency and more mutual acceptance into current practices. To achieve these common positions, detailed consideration was paid to the licensing approaches followed in the different countries represented by the experts of the task force.

A copy of the report can be downloaded.

RSP-660.2 - A Study on Large Truck Accident Statistics in Canada and Ontario

Canadian regulations on the transport of radioactive material are aligned with the requirements set by the International Atomic Energy Agency (IAEA) and ensure protection of the environment and the health and safety of people. As a complement to current knowledge, through this research project the CNSC is aiming to broaden its understanding of transport safety from the perspective of probabilities of accidents occurring. This research project builds on an earlier feasibility study undertaken in 2019 to gauge whether sufficient data exists to underpin quantitative assessments using probabilities in an event tree for large truck accidents in Canada. Accidents involving radioactive material transports are very rare worldwide, so it was not possible to base the statistical data on radioactive material transport accidents alone. Instead, general traffic accident data was used since it was assumed that the distribution of accident probabilities would be similar to regular types of large truck transports.

This research project builds on the 2019 feasibility study on large truck accident probabilities. The objectives of this research project are to:

  1. Analyze available data on causes of large truck accidents in Canada and Ontario, especially those that relate to: 1) drivers being under the influence of alcohol and/or drugs; 2) drivers disobeying traffic controls, and 3) maintenance issues.
  2. Refine and update the event tree in the feasibility study to reflect more recent data on large truck accidents in Canada.
  3. Create a corresponding event tree for large truck accidents in Ontario and determine whether significant differences exist between the accident probabilities in Canada and Ontario.
  4. Gather information on regulatory requirements and industry practices for large truck transport in Canada and other countries where radioactive material transport is prevalent.

In this study, general data on large truck accidents that occurred in Canada and Ontario in 2011-2020 was gathered, screened, analyzed and compared to broaden the understanding of aspects relevant for road transport safety. Event trees were also created to illustrate different types of accidents and their relative distribution.

Request the RSP-660.2 final report

RSP-691.7 - Performance of buffer and seals under elevated thermal and chemical gradient

Clay bentonite is a component of the engineered barrier system in a deep geological repository being proposed for used nuclear fuel disposal in Canada. Bentonite is rich in swelling clay minerals, such that when confined and if allowed to hydrate with clean water, the bentonite would swell to form a tight seal between the used nuclear fuel canisters and the surrounding bedrock. If allowed to hydrate with pore fluids from the surrounding rock, swell pressures would be expected to be reduced because of dissolved chemicals (like sodium and calcium) within the groundwater. Consideration of site-specific pore fluid interactions on clay properties like swell pressure and permeability is essential. In addition to chemical exposure, the bentonite will also experience elevated temperature at repository depths from natural geothermal gradients and from heat generated by the waste.

The goal of this project was to quantify the engineering (swell pressure, permeability and diffusion) properties of bentonite when hydrated with pore fluids expected in Canadian crystalline and sedimentary rocks and at temperatures up to 95°C.

Laboratory experiments were conducted to determine hydro-mechanical-chemical characteristics of MX-80 bentonite under elevated temperature up to 95°C and porewater chemical concentrations representative of porewater solutions in crystalline rock of the Canadian Shield and in the Cobourg limestone of the Michigan Basin. These experiments were conducted to: measure the swell pressure of the bentonite at elevated temperature during hydration with the porewater solutions, measure the hydraulic conductivity of the bentonite following hydration with the porewater solutions at elevated temperature, examine the performance of bentonite under a chemical gradient following hydration at elevated temperature, and study possible changes to the bentonite when exposed to the porewater solutions at elevated temperature.

Request the RSP-691.7 final report

RSP-715.2 - Technical Seminar - ASME Sec III Div 5 “High Temperature Reactors”

The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code Section III Division 5 rules govern the construction of vessels, storage tanks, piping, pumps, valves, supports, core support structures and nonmetallic core components for use in high temperature reactor systems and their supporting systems. Construction, as used here, is an all-inclusive term that includes material, design, fabrication, installation, examination, testing, overpressure protection, inspection, stamping, and certification. Furthermore, high temperature reactors include gas-cooled reactors (HTGR, VHTR, GFR), liquid metal reactors (SFR, LFR), and molten salt reactors using liquid fuel (MSR) or solid fuel (FHR).

This technical seminar provided an overview the background, code structure, technical basis and rational for ASME Section III Division 5 code requirements as they relate to the structural integrity of High Temperature Reactors to CNSC staff and Authorized Inspection agencies.  The focus of the seminar was on the important aspects or parameters in materials, design, fabrication and installation, examination, testing, overpressure protection for ensuring safety operation of high temperature reactors, specifically high temperature SMRs

It was a two-day in-person and in-virtual seminar. Dr. Mark C.  Messner from Argonne National Laboratory (ANL) led the technical seminar. He is the Group Leader of the Thermal and Structural Materials Modeling and Simulations Group in the Applied Materials Division at Argonne where he works on modeling, simulating, and designing high temperature materials and structures. The seminar covered the basics of running through the Class A design rules for simple sample problems.  The seminar covered the 2023 edition of the Code, with an overview of how the Code rules were expected to change in the upcoming 2025 addition.  Additional topics included a description of resources for further study and a high-level description of the Code sections not covered by the detailed modules.

Request the RSP-715.2 final report

RSP-741.1 - The Effect of changing extreme climate events to structural loads of containment structures due to wind, snow and temperature

Extreme weather events are becoming more frequent, intense, and severe worldwide due to climate change. As climate patterns change, relying solely on historical data and trends are not sufficient to assess the safety and resilience of critical nuclear infrastructure. While existing research has mainly focused on isolated factors, real-world conditions involve complex interactions between multiple stressors, such as high temperatures, salt spray, and storm surges. This underscores the urgent need for comprehensive multi-hazard assessments to evaluate the compounded risks posed by climate change.

The purpose of this report is to examine the multifaceted impacts of climate change on these facilities and explore potential future risks under various climate scenarios. The findings provide essential insights into the vulnerabilities of nuclear power plants (NPPs) and the adaptive measures required to address evolving challenges.

This study evaluates the physical risks that climate change poses to nuclear power plants (NPPs) in Canada, focusing on how expected changes in environmental loads due to climate change may impact concrete containment structures over the next 50 years. Key climate related hazards were studies such as extreme temperature, precipitation, wind, and sea-level rise.

Request the RSP-741.1 final report

RSP-749.1 - Review of the most current risk information related to exposure of embryos and fetuses to ionizing radiation

In Canada, the Radiation Protection Regulations (CNSC, 2021) include an effective dose limit for nuclear energy workers (NEWs) of 4 mSv, for the balance of the pregnancy starting from the date on which the licensee has been informed by the NEW, in writing, of the pregnancy.

The CNSC dose limit for pregnant NEWs was informed by a comparative risk analysis, public meetings and by a stakeholder consultation, conducted in the 1990s, which sought views from female workers.

In order to ensure that the CNSC’s framework for protection is based on the latest scientific information, the CNSC sought a review of the current understanding of risks to the embryo/fetus from in utero exposure to ionizing radiation.

The purpose of the study is to provide a comprehensive literature review of the risks to embryos and fetuses from ionizing radiation exposure. In particular, this study reviews updated risk information and reviews comparable dose limits and regulatory approaches in various countries globally.

This report provides a comprehensive review of the risks to embryos and fetuses associated with ionizing radiation exposure to occupationally exposed workers. The report also summarizes information regarding radiation dose limits across various countries.

Request the RSP-749.1 final report

RSP-761.1 - Phenomena Identification and Ranking Table for BWRX-300 Design Extension Conditions

The CNSC requested expert support and consultation from Canadian Nuclear Laboratories (CNL) on design extension conditions (DEC) in the BWRX-300, a boiling water reactor (BWR) design from GE Hitachi Nuclear Energy. DEC are a subset of beyond-design-basis accidents that are considered in the design of a facility that may include some severe accidents.

Specifically, CNSC requested that CNL prepare a phenomena identification and ranking table (PIRT) for DEC in the BWRX-300 while exclusively referencing information that was available in the open literature.

Three postulated DEC were considered in this PIRT: (1) the inadvertent closure of an isolation valve on the main steam line coincident with the failure of the primary shutdown system, labelled DEC-A, (2) the break of a small line off the reactor pressure vessel at an elevation below the top of the core, labelled DEC-B1, and (3) a station blackout coincident with common failure of all isolation condenser system (ICS) trains, labelled DEC-B2. The DEC-B scenarios were only considered to the point just before vessel lower head failure, excluding consideration of ex-vessel severe accident phenomena that could obfuscate unique features of the BWRX-300.

Request the RSP-761.1 final report

RSP-771.1 - Geoscientific considerations for a DGR in the Canadian Shield

Canada’s plan for safe, long-term management of its spent nuclear fuel involves containment and isolation from the surface environment in a deep geological repository DGR). The Canadian spent fuel DGR concept was first developed by Atomic Energy of Canada Limited (AECL) (late 1970s – late 1990s);  AECL investigated the feasibility of developing a DGR potentially hosted at a site in the Canadian Shield; the project did not advance to the site selection stage owing to lack of broad public support. Since 2010, the Nuclear Waste Management Organization (NWMO) has been engaged in a site selection process, which considered identifying a Canadian DGR site in a willing and informed host community. The NWMO’s site selection process began with identification of 22 volunteer host communities, 20 potential sites were eliminated [at the time of report writing] as a result of scientific or societal considerations. Selection of one of the two remaining candidates- the Revell batholith in the Canadian Shield of Ontario- and Cobourg limestone in the South Huron area of Ontario, will be announced late in 2024.

The purpose of this report is to review geoscientific considerations for a DGR in the Canadian Shield, based on existing Canadian reports (from the AECL and NWMO projects) and international experience in similar shield environments. In particular, the report reviews the three-dimensional characteristics of crystalline rock settings in the Canadian and Fennoscandian shields as potential repository hosts, in light of the requirements of the safety case to isolate and contain nuclear waste for at least one million years.

Considerations range in scale from continental-scale plate motions that govern in-situ stresses, to microbiota that buffer the oxidation state of groundwater, to nano-scale pores that provide clues to the age and origin of deep-crustal fluids. Temporally, considerations range from Archean (ca. 2.7 Ga) tectonic processes that formed and deformed the crust, through glacial events within the last million years, to predictions of how a DGR will behave through the next million years.

The data used in this report are sourced from publicly-accessible documents: peer-reviewed research articles, conference reports, site reports hosted by implementer websites, and the International Nuclear Information System (INIS) repository of the International Atomic Energy Agency (IAEA).

Request the RSP-771.1 final report

RSP-772.1 - Probabilistic Modelling Approaches for Flaws Growth Rate Estimation in a Multi-Component System

Nuclear power plant components are subject to in-service inspections, which serve as means to detect new degradation mechanisms and monitor component conditions for already known degradation processes. After the detection of flaws, the estimation of growth rate is required to assess the remaining component life and plan for future inspections and maintenance activities accordingly.

Since degradation mechanisms and their growth rates exhibit considerable variation, probabilistic methods have been used to account for uncertainties in fitness for service and operational assessments of reactor components, such as steam generator (SG) tubing. Another complicating factor in the growth rate estimation is uncertainties associated with inspection methods that are used to examine the condition of components and record the flaw dimensions. Inspection uncertainties are generally described by the probability of detection (POD) and sizing error associated with the inspection tooling.

The goal of this project is to investigate existing probabilistic approaches and gain insights into the estimation of flaw growth rates following the discovery of a new degradation mechanism in a multi-component population e.g. CANDU steam generators.

This project reviews probabilistic models of degradation processes, probability of detection (POD) models, and probabilistic analysis of the sizing error associated with inspection methods to consider a probabilistic approach to estimate the growth rate distribution of flaws using the in-service inspection data.

Request the RSP-772.1 final report

RSP-778.1 - Feasibility study on applications of Drones to support CNSC compliance activity

Drones and robotics are rapidly improving technologies that have been begun being utilized in day-to-day operation across the nuclear industry. This project investigated the feasibility of utilizing these technologies to assist CNSC staff in compliance activities.

The CNSC is interested in determining the feasibility of utilizing drones and robotics for compliance activities to find safer, more efficient, and cost-effective techniques for such activities. The five (5) main test cases of interest include:

  • Aquatic Sampling
  • Remote Sampling
  • Monitoring of mine and mill sites
  • Observing security exercises
  • Monitoring construction activities.

This report is limited to input and recommendations from the Canadian Nuclear Laboratories’ (CNL) Uncrewed Aerial Vehicle (UAV) team on the five (5) areas of interest. This report outlines the specific challenges faced by the CNSC in the areas of interest and provides examples of drones/robots and accompanying instruments that could be used to alleviate some or all challenges. This report also describes how each drone/robot could solve the outlined challenges and compares numerous applicable drones/robots based on price and performance.

Request the RSP-778.1 final report

Page details

Date modified: