Research report summaries 2019–2020
Contractors’ reports are only available in the language in which they are submitted to the CNSC.
- RSP-723.1, Assessment of the potential impacts of climate change on probable maximum precipitations applicable to nuclear facilities in Canada
- RSP-691.3, Natural analogues: Review of glacial erosion effects
- RSP-682.1, An Experimental Study of the Effects of Flat Bar Supports on Streamwise Fluidelastic Instability in Nuclear Steam Generators
- RSP-673.2, Oral Fluid Drug Testing Practices
- RSP-671.1, Studies of Molten Metal Solidification in Internal Pipe Flows
- RSP-669.1, Hydrogen/CO Combustion and Passive Autocatalytic Recombiner Behaviour
- RSP-658.2, Overview of State of Research to Characterize Properties of NaF-KF-UF4/ZRF4 Salts and Sodium Chloride and Actinide Trichloride Salts
- RSP-598.2, Integrated Framework for Propagation of Uncertainties in Nuclear Cross-Sections in CANDU Steady-State and Transient Reactor Physics Simulations
- RSP-590.1, Regulatory Assessment of Leakage Through Cracks in Piping Components
- RSP-523.1, Investigation of Consequences of Concrete Alkali–Aggregate Reaction on Existing Nuclear Structures
- RSP-730.1, High Temperature Gas Reactor Technical Seminar
RSP-723.1, Assessment of the potential impacts of climate change on probable maximum precipitations applicable to nuclear facilities in Canada
External hazards, such as flooding, present the potential to be common-cause initiators for nuclear facilities. They can cause damage to multiple systems simultaneously. The Fukushima-Daiichi accident is the most recent example of severe damages caused by floods. Safety assessments of initiating events (internal hazards, external hazards and combinations of the two) play an important role in the safety of nuclear facilities, as initiating events can impair safety systems that limit their potential consequences. While there have been many studies on probable maximum precipitation (PMP) in Canada, climate change and its potential impact on PMP has only been gradually recognized recently.
This project’s objective is to estimate the potential changes to PMP under climate change in Canada at or near Canadian nuclear facilities.
The study team estimates the possible range of PMP changes caused by climate change, through reviewing the following: the methodology of developing PMP estimates; the PMP studies that developed PMP estimates near Canadian nuclear facilities; and national and international studies on the potential impacts of climate change on hydrological parameters that are directly or indirectly used in PMP estimates.
RSP-691.3, Natural analogues: Review of glacial erosion effects
There is ample evidence of recent glacial erosion on the bedrock surface in Canada (Chamberlain, 1888; Prest, 1983; Glasser and Bennett, 2004). Unless the climate system switches drastically, future erosion by glaciers is likely to occur within the next ~100,000 years (Berger and Loutre, 2002; Mysak, 2010). This poses a potential risk to planned deep geological repositories (DGRs) of nuclear waste (Rutter, 1980; Kaszycki and Shilts, 1980; Hallet, 2011; Fischer et al., 2015), which should be factored into DGR site selection and design. Any prediction of glacial erosion in the near geological future should be rooted in an understanding of glacial erosion in the near geological past. What processes eroded bedrock, and when, where and at what rate did the erosion occur? The objective of this report is to review data and ideas pertaining to erosional processes, patterns and rates that have shaped the bedrock surface in Canada. Based on the review, future erosion is predicted (processes, patterns, rates), potential disruptive scenarios are discussed and recommendations on how to fill knowledge gaps are outlined. The findings constitute a test for numerical models that predict glacial erosion (e.g., Hildes et al., 2004; Melanson et al., 2013), such as those in current use at the CNSC.
RSP-682.1, An Experimental Study of the Effects of Flat Bar Supports on Streamwise Fluidelastic Instability in Nuclear Steam Generators
Several CANDU reactor steam generators are scheduled for replacement. Recent experience from the San Onofre Nuclear Generating Station shows that replacement steam generators (RSGs), although close to the original design, did not meet operational expectations. Two brand-new RSGs were shut down after 11 months in operation after experiencing failures due to streamwise fluidelastic instability (SFEI), which had never before been experienced in operational steam generators.
These problems were not anticipated by the manufacturer, the utility (operator), or the regulator. Understanding of the phenomena associated with SFEI is unsatisfactory and requires more research. In preparation for the regulatory challenges associated with CANDU RSGs, the CNSC initiated this pioneering research project to help fill knowledge gaps and to develop guidance for staff for improved fitness-for-service assessment of RSGs.
The scope of this research study focused on recognizing the uncertainties introduced by empiricism in current design guidelines. Wherever design details are dependent on empiricism, the justification for continued fitness for service is usually based on validation by operational experience. Special attention in this study was paid to the effects of void fraction, array geometry, pitch ratio, and tube-to-support clearance on array stability.
RSP-673.2, Oral Fluid Drug Testing Practices
The use of oral fluid as an alternative biological matrix to urine for workplace drug-testing programs is becoming a reality in many domestic and international workplaces. Technological advances and improved methodology for oral fluid drug tests have enabled adoption of this biological matrix as a viable forensic toxicology alternative to urine. Considerations for the use and forensic acceptability of oral fluid as an alternative biological matrix to urine are explored in the context of REGDOC-2.2.4, Fitness for Duty, Volume II: Managing Alcohol and Drug Use, which relies solely on the use of urine as the biological matrix for determining policy violations. In Canada, the United States and other countries, oral fluid drug-testing standards, guidelines and best practices are reviewed to inform the CNSC on the forensic acceptability of using oral fluid in a workplace drug-testing program.
The purpose of this study is to provide the CNSC with the latest research and best international practices on oral fluid drug testing in laboratories and points of collection.
Existing national and international workplace oral fluid drug-testing standards, guidelines and best practices were compared and contrasted with the latest research to provide recommendations to the CNSC on the various drug class cut-offs to be considered within an oral fluid drug-testing program. The advantages and limitations of oral fluid and urine as biological samples for drug testing are compared. Defined processes for oral fluid specimen collection include recommended oral fluid collection devices, collection procedures, training of oral fluid collectors, chain of custody, and packaging and transportation. Additionally, two point-of-collection testing devices for the initial screening of oral fluid for the presence of various drug classes are recommended.
RSP-671.1, Studies of Molten Metal Solidification in Internal Pipe Flows
The CANDU reactor calandria vessel (CV) contains heavy water moderator and is externally surrounded by the shield tank (calandria vault) light water at ambient temperature during normal operation. During a severe accident leading to a loss of cooling to the primary heat transport system with a loss of moderator cooling and loss of moderator makeup, the moderator water can boil off, exposing the fuel channels inside the CV. The loss of the moderator as a heat sink allows the fuel channels to heat up, deform, and disassemble. Eventually, the core debris from the disassembly collects at the bottom of the CV and forms a terminal debris bed. The core debris that is in contact with the inner surface of the CV is cooled by the shield tank water that surrounds the CV outer surface. The core debris at the central region of the debris bed can heat up to form a lava-like molten mixture called “corium”. The gravity-driven corium may run into the horizontal end fittings and melt their internal steel fuel-support plugs if the heat removal path from the end-fitting liner, through the end-fitting body, to the end shield cooling water does not provide sufficient cooling to solidify or freeze the corium. The same corium can also, via gravity, penetrate the vertical draining pipings in the CV.
Carleton University conducted experimental and numerical studies to investigate the phenomena involving molten metal flows in horizontal and vertical pipes and their potential for plugging due to complete melt solidification.
Both horizontal and vertical experimental set-ups were used to investigate melt solidification. Gallium was selected as the working fluid due to its low melting temperature of 29.76 °C as well as its non-toxicity. Gallium was passed through an empty acrylic test section surrounded by ice and water under which the solidification occurred.
A 1-D in-house code was created to simulate the penetration and solidification of gallium in the horizontal pipes. The code was also extrapolated to simulate corium at higher temperatures and different pipe diameters. Vertical studies were used to gauge preliminary feasibility.
RSP-669.1, Hydrogen/CO Combustion and Passive Autocatalytic Recombiner Behaviour
Concerns about hydrogen (H2) combustion due to hydrogen accumulation in containment have been raised within the international community following the Fukushima accident in 2011. In addition to H2, a significant amount of carbon monoxide (CO) can be produced during an ex-vessel molten core concrete interaction, which increases the flammability of the containment atmosphere. To reduce the risk associated with combustible gas mixtures (H2 and CO), mitigation measures exploit catalytic oxidation of H2 to H2O and CO to CO2 using passive autocatalytic recombiners (PARs). Experimental data on the behaviour of stratified H2-air-steam mixtures and PAR performance with CO recombination rates and poisoning effects are very limited. Research was therefore needed to investigate PAR mitigation effectiveness under severe accident conditions.
The objective of this project was to perform experimental studies on the combustion of stratified H2-air mixtures and H2-CO-air mixtures with local high H2 concentrations, and to characterize PAR effectiveness using H2-CO-air mixtures. The experimental results from this project provide a better understanding of the potential risk of non-uniform combustible mixtures and the effectiveness of H2 and CO mitigation measures under severe accidents conditions. The results also validated models in relevant safety analysis codes.
The CNSC commissioned the study of the combustion behaviour of various combustible gas mixtures and PAR performance with Canadian Nuclear Laboratories (CNL). CNL performed the following experiments in its Large-Scale Vented Combustion Test Facility:
- Combustion behaviour with H2-CO-air mixtures
- PAR performance with H2-CO-air mixtures
- Combustion dynamics with stratified H2-air mixtures
The results demonstrated a good alignment of experimental and analytical simulation of anticipated PAR behaviour. Analytical simulations confirmed the GOTHIC code’s capability of capturing hydrogen stratification. Predictions for combustion pressure dynamics were in agreement with measurements and demonstrate conservatism in the analysis results due to the over-predicted peak in pressurization from resulting combustion.
RSP-658.2, Overview of State of Research to Characterize Properties of NaF-KF-UF4/ZRF4 Salts and Sodium Chloride and Actinide Trichloride Salts
CNSC staff are in the process of assessing technical claims and supporting evidence being used by developers of molten salt reactor (MSR) technologies. One of the main novel aspects of these technologies is in the chemical properties of the molten salts that replace traditional water-cooled reactor systems (coolant salts) and, in some cases, the use of traditional solid nuclear fuels (fuel salts). Claims are being made on various long-term performance characteristics of fuel and coolant salts in areas including but not limited to:
- impacts of different salt compositions
- understanding of progression of phase changes during various temperature operating states (including precipitations)
- physical properties as a function of temperature and composition, such as viscosity, boiling/freezing point, and impurities solubility
- chemistry stability under radiation exposure over time
- potential for formation and transport of corrosion products
- potential for erosion effects on in-core reactor components (e.g. steels, graphite, other alloys)
- reactor performance and fission-product retention under all plant states, including design-basis accidents, design extension conditions and beyond-design-basis events
- penetration of salts into reactor component materials
- electrochemical contribution to in-core component aging
- thermodynamic and physics performance/impacts ranging from high-temperature liquid states to cold solid states
As part of its ongoing readiness to assess new designs, the CNSC requested a seminar from the Joint Research Centre (JRC) Karlsruhe, which has a long-standing program of research on MSR fuel and coolant salts. CNSC staff used the contract in place with the JRC to engage in information exchanges to further inform regulatory activities and planning. The CNSC also invited interested staff from the United States Nuclear Regulatory Commission in an effort to inform potential regulatory cooperation being planned for 2020.
JRC Karlsruhe provided a two-day technical seminar on the state of research to characterize properties of NaF-KF-UF4/ZRF4 salts and sodium chloride and actinide trichloride salts. Topics included the following:
- JRC introduction
- MSR history and R&D efforts
- Molten salt experimental facilities
- Key uncertainties existing in the data on fuel salt properties
- Boiling/freezing point and solubility of impurities
- Thermal properties
- Monitoring and assessment of corrosion
RSP-598.2, Integrated Framework for Propagation of Uncertainties in Nuclear Cross-Sections in CANDU Steady-State and Transient Reactor Physics Simulations
Current CNSC REGDOC-2.4.1, Deterministic Safety Analysis, allows for the use of more realistic methodologies, such as best-estimate (BE) reactor analysis simulations with consideration of uncertainties. Quantification and understanding of uncertainty sources is an essential requirement of BE analysis, as it provides a reliable metric by which the quality of the predictions can be assessed. In preparation for the independent verification of licensees’ safety cases using more realistic methodologies, the CNSC initiated a study to investigate the feasibility of the development of a first-of-a-kind integrated framework for uncertainty characterization (UC) with primary application to CANDU neutronics calculations. The framework is based on open-source libraries for standard UC process algorithms as well as novel algorithms.
Phase 2 of the project focused on developing an integrated automated capability for uncertainty analysis for the core simulator NESTLE-C in both steady-state and transient CANDU core calculations. The end result was a complete library of diffusion cross-sections, an associated uncertainty distribution function and covariance matrix, and an integrated platform that included an execution script, sampler software, NESTLE-C, and processing software that allows the user to estimate the joint probability distribution for all output results.
As a result, CNSC staff have enhanced their ability to independently assess licensees' submissions for Category 3 CANDU loss-of-coolant accident (LOCA) safety issues and more realistic LOCA power pulse simulations to enhance risk-informed decision-making processes.
RSP-590.1, Regulatory Assessment of Leakage Through Cracks in Piping Components
Several through-wall flaws in pressurized components (on both the nuclear and conventional side of a nuclear power plant) may remain in operation and potentially contribute to the total primary-to-secondary leakage and release of radioactive content. When using available simplified methods for the estimation of discharge rates, various suitable factors have to be chosen and applied. Currently, there is no broadly verified, validated and accepted method to evaluate leakage through cracked metallic walls.
The CNSC initiated this research project to develop a database of critical discharge through various realistic crack geometries with controlled subcooled liquid flow, stagnation and entrance conditions. This study is an important step toward understanding and quantifying the impact of the structural parameters of various cracks on the thermal hydraulics dynamic of escaping fluid.
There is very limited data on steam generator tube and thin-walled piping leak-rate measurements. The scope of the project was to provide experimental data and the predictive correlations and models needed to permit staff to independently evaluate the fitness for service of steam generator tubes as plants age and degradation occurs, as new forms of degradation appear, and as new degradation-specific fitness-for-service strategies are implemented.
RSP-523.1, Investigation of Consequences of Concrete Alkali–Aggregate Reaction on Existing Nuclear Structures
The University of Toronto conducted an investigation of the consequences of concrete alkali–aggregate reactions on existing nuclear structures. This research entailed three connected streams of study: materials level characterization, structural tests and modelling.
Alkali–aggregate reactions, particularly alkali–silica reaction (ASR), has been identified in the concrete of nuclear power plants (NPPs) in Canada and elsewhere. This has potential implications for the structural integrity and/or serviceability of aging NPPs, and no assessment criteria or guidelines are currently available to assess the consequences. This work focuses on three aspects of this issue: concrete materials, structural testing, and analysis and assessment. The materials part of the work focused on the characterization of unrestrained and restrained ASR-affected concrete. Some of the test measurements include longitudinal and transverse expansion; damage rating index along three planes; nano-mechanical properties by nano-indentation; microstructural analysis; and macro-mechanical properties by destructive and non-destructive tests.
The experimental structural part of the program involved testing of reinforced concrete walls to investigate the effects of ASR on shear strength. A total of six shear wall specimens were designed and cast using two different types of concrete. One type contained reactive aggregates (referred to as ASR concrete) and the other type contained non-reactive aggregates (regular concrete). A non-linear finite-element analysis procedure was developed for the modelling and analysis of ASR-affected reinforced concrete structures. Two mechanisms were considered to model the effects of ASR: the magnitude of the induced expansion and the changes in the mechanical properties of the concrete.
RSP-730.1, High Temperature Gas Reactor Technical Seminar
As part of the optional pre-licensing vendor design review process, the CNSC provides feedback to vendors on their reactor technology early in the design process. The objective of a review is to verify, at a high level, the acceptability of a nuclear power plant design with respect to Canadian nuclear regulatory requirements and expectations, as well as Canadian codes and standards.
Several vendors have proposed conceptual designs using new HTGR technologies. The CNSC supported the technical seminar to enhance its staff’s knowledge related to advanced reactor design concepts.
The CNSC and BriVaTech Consulting collaborated to provide a technical seminar on High Temperature Gas-Cooled Reactor (HTGR) Technologies. CNSC staff invited BriVaTech to share its research and operating experience on High Temperature Gas-Cooled Reactors. BriVaTech delivered a three-day technical seminar, covering different High Temperature Gas-Cooled Reactor concepts. The organization compiled research and operating experience and presented on topics related to nuclear physics, chemistry, materials, engineering design and safety analysis for High Temperature Gas-Cooled Reactors.
Dr. Gerd Brinkmann of BriVaTech Consulting previously prepared and delivered a technical seminar on HTGR technology for the UK Office of Nuclear Regulation (ONR) in early 2019. CNSC staff requested the same technical seminar that was developed for the UK ONR in order to procure this information session in the most effective manner. The scope of this contract included the presentation of a technical seminar on HTGR technology and follow-up discussions between CNSC staff and Dr. Brinkmann.
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