Research report summaries 2020–2021

Contractors’ reports are only available in the language in which they are submitted to the CNSC.

RSP-690.1, Best Practices for Probabilistic Fracture Mechanics Evaluations

This study reviewed current practices for Probabilistic Fracture Mechanics (PFM) assessments, identified strengths and gaps, and ultimately recognized a set of best practices for PFM assessments. Based on this information, a guideline document for best practices for PFM analyses and evaluations was developed to assist in future PFM applications and regulatory considerations. This project was a two-year study that began in July 2019 and was funded by the CNSC.

Fracture mechanics assessments for pressure boundary components are typically performed using deterministic evaluation methodologies where uncertainties are recognized and implicitly accounted for through the selection of bounding values for input parameters and the use of safety factors.  Deterministic methodologies have been widely adopted by standards bodies and incorporated into nuclear power plant regulations in many countries. Development of PFM approaches began in the late 1990s in the nuclear industry, which explicitly consider the uncertainty in input parameters and models and the effects of those uncertainties on the results of a fracture mechanics calculation. With advancements in computational technologies, significant developments in PFM analysis since the late 1990s include specification of useful PFM models and uncertainties, development of PFM software, understating the limitations of PFM and regulatory decision-making using PFM results. In recent years, the Canadian nuclear industry has expressed interest in wider spread adoption of PFM to evaluate the condition of pressure boundary components.

The scope of the project involved the follow key elements:

  • Performing a literature review of best practices for PFM assessments in nuclear and non-nuclear industries
  • Identification of critical elements of PFM based on the literature review
  • Development of draft guidelines for PFM best practices that summarizes the team’s experience, results, and decisions, including uncertainty analysis

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RSP-762.1, Review of the Canadian Nuclear Safety Commission’s Regulatory Framework for Readiness to Regulate Fusion Technologies

Reviewing the implementation of the CNSC’s current Regulatory Framework with respect to novel technologies is an area of broad interest and concern. Fusion is widely viewed as a cleaner, safer, and cheaper means of producing power. To date no entity has succeeded in commercializing fusion as all research fusion reactors require more energy in to maintain criticality than energy outputted. There are, however, fusion organizations on a path with established utilities to build demonstration facilities to prove the feasibility of commercial fusion technology. The CNSC is working under the assumption that there may be fusion technology companies that will seek to enter into a formalized Vendor Design Review (VDR) agreement in the near future.

The purpose of this project was to establish an external research contract to review the CNSC regulatory framework’s readiness for regulating fusion, and to suggest areas where modification may be required as appropriate. This research helps ensure that the CNSC’s regulatory framework reflects the necessary readiness and agility to review license applications for fusion technology.

The scope of this project involved the following elements:

  • Developing hypothetical preliminary descriptions of fusion facilities covering a range of approaches to fusion to use to test the Regulatory Framework
  • Researching and interviewing regulators and stakeholders on their approach to regulation of fusion technologies in the USA, UK, France, and Japan, as well as relevant IAEA reports
  • Assessing the readiness of the CNSC’s Regulatory Framework to license a fusion facility using the hypothetical models

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RSP-720.1, Class II Equipment Servicing Accreditation Program

This report, prepared for the Canadian Nuclear Safety Commission, describes the potential safety hazards associated with maintenance and repair of linear accelerators and cyclotrons. The Canadian Nuclear Safety Commission (CNSC) seeks to identify the hazards associated with accelerator maintenance beyond the radiation hazards, in order to ensure that Canadian technicians who service this equipment have the appropriate knowledge of these hazards. The results of this report may lead to some form of accreditation process, knowledge requirement or teaching aid for these technicians.

Safety in accelerators is a vast subject. Despite this, very few resources are available to train maintenance personnel. Training courses are provided by the equipment manufacturer, however these are mostly technical in nature and do not teach the Canadian regulations or safety standards. This work supports the development of a training program for service technicians so they understand and are prepared to address safety hazards within the accelerator environment. Unique hazards encountered during accelerator maintenance are discussed. A review of Canadian safety regulations and standards that apply to accelerator maintenance is provided.

The CNSC conducted a survey of Canadian accelerator maintenance workers to obtain information on their training and the nature of their work. Results from this survey are discussed here. The survey found that the level of training which maintenance workers receive is not consistent. While the majority of workers attended manufacturer’s training courses, most reported not having received formal in-house training. Some reported having received no training other than informal on-the-job training and self-learning.

Training in safety and in the regulations are regulatory requirements in Canada. While the CNSC regulates the use of particle accelerators from a radiation safety perspective, their regulations do not cover other aspects of the work such as electrical or chemical hazards, or rules for working on pressure vessels. This report recommends that a safety training program be developed for Class II equipment maintenance personnel which covers three principle training paradigms- technical training, safety, and regulatory training, in order to meet the definition given of competency as given in the regulations.

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RSP-734.1, Development of an Innovative Probabilistic Compliance Reliability Index for Nuclear Power Plant Oversight

The Canadian Nuclear Safety Commission (CNSC) conducts regulatory activities under the powers of the Nuclear Safety and Control Act to assess licensee compliance with regulatory requirements. The licensing basis sets the boundary conditions for acceptable performance at a nuclear power plant and thus establishes the basis for the CNSC’s Compliance Verification Program. The CNSC conducts compliance verification activities to confirm that licensees carry out their licensed activities in compliance with the regulations and conditions specified in their licenses. Through this project, findings from the CNSC’s compliance verification activities are tracked and trended to develop a performance index of licensee adherence to policies and procedures. This performance index is referred to as a “Compliance Reliability Index” (CRI).

The main objective of this project was to develop a Compliance Reliability Index (CRI) for the CNSC to assess regulatory compliance. The project aimed to develop an innovative probabilistic approach to assess the level of programmatic compliance by licensees to their policies, procedures, and regulations. The development of the CRI is beneficial to the CNSC as it feeds into the overall compliance assessment process of CNSC staff activities at the Safety and Control Area (SCA) as well as the Specific Area (SpA) level.

The approach of this project was based on the principle of reliability theory that is used in the nuclear industry as well many other fields of engineering and science. This project drew from other fields of engineering and science to apply as an improvement to how CNSC staff assesses compliance of Nuclear Power Plant (NPP) licensees within Canada. The Bayesian method was used for the CRI as it proves to be a systematic and transparent assessment of partial or uncertain information. This report provides all necessary details of the Bayesian CRI evaluation method with many illustrative examples accompanied with insightful discussions. It is expected that the information presented in this Report would serve as a guidebook to CNSC staff for an effective use of the proposed method for compliance verification. Computation formulas developed in the project are easily implemented in MS Excel.

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RSP-722.1, Independent Expert Review of Bruce Power Mitigation Measures Study

Following the renewal of Bruce Power’s operating licence in 2018, a condition was established in the Licence Conditions Handbook that requires Bruce Power to conduct an assessment of feasible mitigation measures for thermal effluent and impingement/entrainment improvements. To satisfy this obligation, Bruce Power submitted a final mitigation measures study to Canadian Nuclear Safety Commission (CNSC) on March 31,2020. Kinectrics Inc. was contracted by CNSC to provide an independent review of the Bruce Power draft report in 2019 and final mitigation measures study in 2020. The overall goal of this review was to provide an assessment as to whether the objective and intent of the mitigation measures study was fulfilled.

Kinectrics has completed a review of the Bruce Power final mitigation measures study. The report provides a comprehensive description of a wide range of technologies for potential mitigation of impingement/entrainment and thermal discharge issues at Bruce Power. Bruce Power already has an underwater water intake which minimizes fish impacts compared to the more common surface water intakes. This makes it difficult to justify major capital-intensive solutions to impingement/entrainment and thermal discharge issues without combining the technology with additional operating savings or generation efficiencies.

The report sufficiently assessed most practical available mitigation measures. The report meets the objectives of the study and provides the necessary information that would be required to move to the next level of evaluation. The methodology followed in the evaluation was scientifically sound and provided the required information to generate the data necessary to complete the mitigation matrix. The mitigation matrix is an accepted tool for ranking a range of technologies for potential solutions to a problem. It is recommended any future evaluations are conducted on solutions that can provide mitigation to fish and thermal impacts by the station in addition to reducing debris and/or operating costs to the station or improving operating efficiencies to assist with offsetting the costs of the mitigation measures only. If zebra mussel or algae content in the cooling water could be reduced or increased condenser performance can be linked to these mitigation measures it would be a win/win technology and more likely to be supported at all levels.

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RSP-718.2, Environmental Chemistry of Wild Harvested Berries: Depositional and Uptake Chemistry and Human Health Assessment - Laboratory Analysis

The Wild Berry Project is a research collaboration between the CNSC, the University of Ottawa, and the University of Alberta. In northern Saskatchewan, there are CNSC-licensed facilities, specifically active and decommissioned uranium mines and mills. In these areas, nearby Indigenous communities harvest foods and wild berries as an important part of their traditional diets. Wild blueberries and the soil they were growing in were sampled near CNSC-licensed facilities in northern Saskatchewan. The samples were tested using extremely accurate methods to find out what kinds of radioactive elements and non-radioactive metals were in the samples and how much. To compare with the Saskatchewan blueberry and soil results, commercially-available blueberries were collected from Ontario farms and grocery stores. The goal of this research project is to help inform CNSC’s regulatory decision-making process and this research will support future human health risk communication with Indigenous communities.

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RSP-715.1, Identifying Limitations of ASME Section III Division 5 For Advanced SMR Designs

Many Small Modular Reactor (SMR) designs reviewed by the Canadian Nuclear Safety Commission (CNSC) operate at temperatures up to 750°C, while accident conditions may reach over 1000°C. The vendors have proposed to use ASME Section III Division 5 “High Temperature Reactors” for the design and construction of these SMRs. Due to the complexity of code rules and design analyses, the CNSC organized a technical seminar to provide staff with in-depth information and insight to better understand the code and its correct use in SMR design review.

The purpose of this report is to provide an overview of the ASME Boiler & Pressure Vessel Section III, Division 5 rules for the design and construction of high temperature nuclear reactor components. The overview focuses on the application of the rules to the design of Small Modular Reactors (SMRs).

Objectives of this report are to:

  1. Enhance CNSC’s staff knowledge in understanding the important aspects and challenges for design and construction of metallic components operating at high temperatures;
  2. Enhance CNSC staff’s knowledge in understanding the ASME Section III Division 5 requirements that relate to structural integrity of metallic components, using technical basis and rational for these code requirements;
  3. Identify any conditions/restrictions of using Division 5 code, and train CNSC staff in using the code correctly for reviewing specific SMR designs.

The discussion of this report covers the general ASME Code rules for base metal design and construction, the rules for designing weldments, and provides an overview of environmental degradation mechanisms affecting reactor structural materials. The analysis includes historical context on the development of the ASME design approach and a description of what actions could be taken to mitigate the gaps identified in the report. The report concludes with a summary of the key gaps identified in the rules, as they apply to SMR, and a list of recommendations on how those gaps might be addressed.

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RSP-709.1, Human Workload and Fatigue Literature Review

One of CNSC’s safety and control areas (SCAs) is on human performance management, which covers activities that enable effective human performance through the development and implementation of processes that ensure that licensees have sufficient staff in all relevant job areas with the necessary knowledge, skills, procedures and tools in place to safely carry out their duties. An important component of ensuring sufficient qualified staff at nuclear facilities is to ensure that the physical and mental workload of minimum staff complement is achievable during normal operations, anticipated operational occurrences, and design basis accidents. Minimum staff complement refers to defining the minimum number of workers with specific qualifications who will be available to the nuclear facility at all times. It is important for CNSC staff and Nuclear Power Plant (NPP) operators to understand how workload and fatigue impact performance, as well as the tools used in measuring workload.

To improve understanding on this topic and ensure CNSC is aware of useful advancements based on the latest research and development, The Human and Organizational Performance Division of the CNSC has commissioned a literature review of how workload levels and stress impact human performance.

The scope of the literature review is mainly focused on objective and subjective methods for measuring workload, fatigue, situation awareness, and human performance. This concerns how humans perform under acute stress, which is sudden, novel and of relatively short duration, and not of cumulative or life stress conditions.

The literature review also looks into the use of tools, especially those that are wearable, such as eye trackers, sweat gland activity sensors, heartrate and blood pressure sensors, and electrical brain activity sensors. This aims to see the viability of these tools for NPP operators to better understand how workload impacts operators’ ability to complete tasks during normal and emergency scenarios.

Lastly, the literature review assesses a variety of workload modeling tools that can be used to understand the workload of operators in NPPs.

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RSP-706.1, Probabilistic Assessments: Principles and Computational Methods

The risk-informed decision-making approach has been receiving increasing consideration by the nuclear industry and regulatory authorities worldwide. In Canada, using a risk-informed decision-making approach has been a successful way to address safety of nuclear reactor components. This success motivated the introduction of probabilistic methodologies into evaluations of CANDU reactor components including pressure tubes, steam generator tubing and feeder piping.

The objective of this project was to investigate and evaluate generic principles and computational methodologies used in probabilistic frameworks to assess reactor components and systems. In the project, concepts of a time-dependent reliability theory were applied to develop essential attributes of the probabilistic assessment of nuclear systems, structures and components (SSCs). An overall goal of discussions provided in the report is to pave the way for adequate applications of the reliability theory to the risk-informed decision making in the nuclear industry.

The scope of this project included the investigation on probabilistic principles for conceptually rigorous probabilistic frameworks and computational probabilistic methodologies for nuclear components, and the evaluation of generic effects of the assumptions made in probabilistic evaluations on the results and interpretation of the results. The outcomes of this project are conceptual features that could serve as the foundation for probabilistic assessments of passive components or systems.

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RSP-703.1, Analysis of histological types of incident lung cancer among workers exposed to radon and gamma radiation during uranium mining and milling at Eldorado Nuclear Ltd., Canada

The CNSC conducts health studies to guide decisions affecting its regulatory framework. Lung cancer is the most commonly diagnosed and leading cause of cancer death in Canada. Radon and radon decay products are a known cause of lung cancer among uranium workers.

The primary objective of this study was to assess the radon risk of lung cancer by histological subtype (squamous cell, small cell and adenocarcinoma) among 16,752 Eldorado uranium workers from the Port Radium and Beaverlodge uranium mines, and a radium and uranium refinery and processing facility in Port Hope, Canada. Workers were employed from 1932–1980 and followed up for cancer incidence from 1969–1999.

The study uses advanced statistical methods, Poisson regression, to estimate risks of lung cancer from radon (and radon decay product) exposures. The report concluded that incident cases of lung cancer were significantly higher among workers compared to the general Canadian male population. The radiation risk of lung cancer, based on 594 lung cancers, increased with increasing radon exposure. The study observed a non-statistically significant increased risk for small cell and squamous cell histological subtypes. Differences in radon risk by histological subtype of lung cancer helps to further our understanding of radon’s role in lung cancer development. This work advances the international understanding of radiation risk and supports the international radiation protection framework for radon.

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RSP-674.1, Investigation of two-phase flow phenomena in reactor headers

The primary heat transport system of CANDU reactors consists of a large number of horizontal fuel channels connected by small multi-section feeders to large horizontal inlet and outlet headers. For scenarios where two-phase flow in reactor headers can occur, the phase distribution through the header and consequently the flow through the feeders can be different from the average lumped model. . Such models and their uncertainties directly impact the assessment of fuel and fuel-channel integrity during various accident scenarios, such as loss of coolant accidents (LOCA). During LOCA, two-phase flow can occur in the reactor headers and this may affect flow distribution in the fuel channels, such that some fuel channels may receive larger amounts of steam than others, thus impairing adequate fuel cooling.

The objective of this report is to assess the adequacy of lump and 1-dimensinal modelling of reactor headers for scenarios of interest, such as SBLOCA and LBLOCA; to a) perform measurements of key parameters characterizing two-phase behavior in the header manifold including: pressure, gas and liquid flows at the header entrance, liquid levels in the header, gas and liquid flow rates in the attached pipes simulating feeders; b) observation and documentation of the flow patterns for various break orientations; d) perform steady state and transient numerical simulations for header conditions relevant to SBLOC; to advise CNSC staff on the capabilities of modern, three-dimensional thermalhydraulic simulation methodologies; and to assist the safety analyses of existing CANDU-type reactors by improved modelling of reactor headers.

This report summarizes the experimental and analytical research activity on air-water flows in a CANDU type header-feeder model during the period from February 15, 2019 (date of signing the contract) through October 31, 2020. Progress during the period preceding the start of the contract has been summarised as well. Plans for future experimental and analytical work are also outlined.

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RSP-602.3, Develop irradiated fuel bay Severe Accident Analysis computer code (PKPIRT) - Phase 3 (BAYLOCA)

Following the Fukushima Daiichi nuclear accident, CNSC staff identified the need for detailed computational models to analyse highly unlikely severe accident scenarios involving an irradiated fuel bay (IFB).

Previous work on an IFB phenomena and key parameter identification and ranking table determined that development of an analysis code was feasible and that an adequate knowledge base existed.

This work package’s objective was to develop an analytical code to model severe loss of cooling/coolant accidents in a CANDU irradiated fuel bay, specifically for use in the CNSC’s emergency operating center.

The code developed, named BAYLOCA, is a lumped parameter code that can generate results quickly to meet the requirements of the CNSC emergency operations center processes and is sophisticated enough to also be used to inform regulatory decision making. It models thermalhydraulic behaviour, heat transfer, fuel and fuel sheath oxidation and fission product release phenomena relating to fuel, fuel racking, the IFB coolant, IFB structures and surrounding building structure. The primary figures of merit that the code calculates are IFB water level, IFB water temperature, fuel temperature, fuel sheath integrity and source term release. CNSC staff will be able to use the code to more effectively track the progression of a IFB severe accident should it occur, assess the safety margins, and better understand the impact of mitigation measures.

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