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Research report summaries 2018–2019

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

RSP-444.3, Effect of non-linear soil behaviour on the seismic response of soil-structure systems

Structures designed to contain nuclear material, such as nuclear power plants and tailings dams, and other structures used for containing radioactive waste, must be designed to retain their containment function in a design basis earthquake (DBE). When these containment structures are built on or within soil deposits, the soil responds to the DBE influences and interacts with the response of the structure. Consequently, soil-structure interaction phenomena need to be understood in order to verify that a structure can withstand a DBE without losing its containment function.

The objective of the research is to develop a tool to analyze the seismic response of containment structures built on or embedded in soil foundations with the effects of pore water pressure generation and redistribution.

The tool is a mathematical model expressed as governing partial differential equations to represent the motion of the soil, structure and groundwater in the soil mass. The governing equations are numerically solved with the finite element method using a commercial software, COMSOL Multiphysics. The model is validated by comparing its results with experimental results from cyclic triaxial tests and shaking table experiments. The model can successfully simulate phenomena like soil plasticity and liquefaction. A scoping analysis was also performed for a hypothetical small modular reactor embedded in a sand deposit.

Read the RSP-444.3 final report (PDF)

RSP-586.2, IAEA technical document on fuel safety criteria for pressurized heavy-water reactor fuel

Technical Review of Acceptance Criteria for Pressurized Heavy Water Reactor Fuel is a technical document (TECDOC) of the International Atomic Energy Agency (IAEA). The main objective of this TECDOC is to provide a collection of current practices in the fuel safety criteria used in CANDU-type pressurized heavy-water reactor (PHWR) countries, as well as technical justification available for such criteria. Argentina, Canada, China, India, Korea, Pakistan and Romania contributed to this TECDOC.

With Canadian expert support, the IAEA completed Version #1 of this TECDOC in November 2016. At the request of the IAEA, the CNSC agreed to lead the Canadian industry and CNSC staff review of Version #1 and produce Version #2. Version #2 was completed on February 22, 2019.

The IAEA sent Version #2 to the other countries participating in this project for review. This report is included as Annex I in Version #2 of the IAEA TECDOC and is part of the deliverables that M. Tayal had to submit to the CNSC under research contract RSP-586.2.

The purpose of Annex I is to describe an approach, currently being developed by SNC-Lavalin Nuclear, for the establishment of fuel safety criteria for a new fuel design for a new reactor and/or for new operating conditions that differ significantly from current operating CANDU reactors in Canada. This approach would also apply to currently operating CANDU-type PHWRs.

Annex I focuses on normal operating conditions. However, the approach described also applies to anticipated operational occurrences. Annex I contains no proprietary information; as the reference section indicates, it is entirely based on information found in open literature.

Read the RSP-586.2 final report (PDF)

RSP-613.4, Laboratory determination of sealing material performance

Two initiatives for the deep geological disposal of radioactive wastes are currently being proposed for Canada. The proposed repositories rely on both host rock and engineered barriers to contain and isolate waste from the biosphere for hundreds of thousands to millions of years. Bentonite is proposed worldwide for use in the engineered barrier system as the buffer and seal material. The characteristics of compacted bentonite are of key importance for the long-term performance of bentonite-based barrier systems. Therefore, there is a strong need to understand the long-term performance of bentonite in the Canadian environment.

The objective of this research is to obtain laboratory data on the performance of bentonite buffer and seal materials in a brine environment that exists at an anticipated repository depth of more than 500 metres in Canadian sedimentary rock. The CNSC will use the experimental data in the development of its numerical models, in order to assess the long-term performance of buffer and seal materials for repositories in sedimentary rock.

Queen’s University conducted laboratory experiments to measure the hydro-mechanical properties of compacted MX-80 bentonite and MX-80 bentonite-sand mixture. A synthetic multi-component hypersaline water was developed to mimic the constituents and concentrations of pore water found in the sedimentary rock of the Michigan Basin in Canada, and was used for the experiments. The behaviour of swelling, permeability, suction and compression of compacted bentonite and bentonite-sand mixture was investigated under conditions where the materials interact with both deionized water and the hypersaline water.

The experimental results indicate that pore water salinity plays an important role in the performance of bentonite as buffer and seal material for a deep geological repository and that the effects of hypersaline pore water should be considered in designing deep geological repositories in a highly saline environment.

Read the RSP-613.4 final report (PDF)

RSP-645.1, Statistical Modelling of Aging Effects in Failure Rates of Piping Components

The operating experience with safety- and non-safety-related piping in commercial nuclear power reactors is quite extensive. Equally extensive is the experience gained from implementing different degradation mitigation strategies. Regulatory organizations can use this body of field experience data, in combination with advanced piping reliability models, to assess the projected structural integrity of piping systems and vessels in operating nuclear power systems.

The objective of this project was to develop a technical manual on the underlying theory and associated methods and techniques for performing operability determinations through risk characterization of carbon steel passive components that exhibit structural degradation due to operating conditions and aging. The technical manual establishes a technical basis for a data-driven approach to piping reliability analysis. It acknowledges the potential negative impacts of material aging on pressure boundary integrity and compares them to the positive impacts of reliability program implementation and management on the integrity of pressure boundary components.

The technical manual summarizes the different approaches to fitness-for-service and operability determination and elaborates on the different ways that operating experience data can be used to validate fitness-for-service and operability determination results to support the associated decision-making processes.

The technical manual is intended to support four types of risk-informed structural integrity analyses:

  1. risk-informed operability determination of degraded conditions
  2. independent assessments of licensee applications that include structural reliability analysis results
  3. structural integrity projections of influence by changes to reliability and integrity management strategies
  4. independent assessments of risk-informed in-service inspection program applications.

Risk-informed operability determination is concerned with the risk characterization of a degraded piping pressure boundary and its impact on operating nuclear power plant risk. This manual includes examples of how to use probabilistic, statistical and data-driven models in a synergistic context.

Read the RSP-645.1 final report (PDF)

RSP-660.1, Radioactive Material Transport Probabilistic Risk Assessment – Large Truck Accidents on Canadian Roadways

Transport of radioactive material in Canada is jointly regulated by the Canadian Nuclear Safety Commission (CNSC) and Transport Canada (TC). Canadian regulations for this activity are in place to ensure the protection of the environment and the health, safety and security of people. The CNSC has initiated a project that uses probabilistic methods to quantitatively assess the safety risks associated with this activity. This study is intended to strengthen and facilitate communication of risks associated with radioactive material transport.

The overall purpose of this study is to explore the feasibility of using probabilistic methods to assess the safety risks associated with radioactive material transport by road in Canada. Specifically, the objective is to gather and analyze relevant road accident data and develop an accident event tree that is specific to Canadian conditions.

The project focused on large trucks that could transport Type B packages with a high radioactivity content, as these would pose the highest risk in the event of a traffic accident. The study covered accident data on public roadways only. Human factors and human risk assessment were assumed to be implicitly included in the accident data. The study also compared national collision data with the number of vehicle registrations for a given year to assess the likelihood of a large-truck accident occurring in the first place. Aspects of security, safeguards or the possibility of malevolent acts were considered out of scope. No specific transport routes were analyzed.

The generated event tree provides the relative distribution of different traffic accident scenarios and their respective conditional probability of occurrence. The study found that the probability of a large-truck accident ever occurring was very low, which provides additional confidence that the safety of transporting radioactive material is high. Further studies will be needed to integrate accident probabilities with accident consequences for Type B packages.

Read the RSP-660.1 final report (PDF)

RSP-567.1, Review and Standardization of Testing Procedures for Irradiated Zr-2.5Nb Pressure Tube Material

Leak-before-break and fracture protection must be assessed in order to demonstrate fitness-for-service of pressure tubes in the reactor core. Fracture toughness, or the material resistance to crack propagation, is an important parameter for these evaluations. Pressure tube fracture toughness is typically measured by bursting a section of the pressure tube with a machined through-wall crack. Another method to determine pressure tube fracture toughness uses small compact tension specimens to transfer fracture toughness to the reactor component.

The objective of this research project was to standardize small-scale fracture toughness testing procedures to analyze Zr-2.5Nb pressure tubes. The testing procedures were developed for compact tension specimens to address the pressure tube size and geometry limitations. Testing small specimens utilizes less material and allows for systematic parametric studies.

Fracture toughness standards of the American Society for Testing and Material (ASTM) were developed for flat specimens. However, compact tension specimens of pressure tube material are curved and relatively small due to the pressure tube geometry. The effects of specimen curvature, specimen design and crack tunnelling were assessed through a detailed numerical and experimental investigation. Finite element simulations were performed for stationary and growing through-wall cracks. Crack tunnelling was captured by adopting a damage mechanics approach based on critical strain. Laboratory experiments were carried out on 17 mm- and 34 mm-wide specimens. The experimental results showed a much longer crack extension for the 34 mm specimen, which was more similar to the results from the pressure tube section burst tests. The effect of specimen curvature was insignificant for the loads applied to testing pressure tube material with hydrides. The fracture resistance curve, or J-R curve, determined using the ASTM standard E-1820 was in excellent agreement with the numerical results. The study showed that the procedures in this standard are also applicable to the compact tension specimen geometries investigated in this work.

Read the RSP-567.1 final report (PDF)

RSP-686.2, Radiation basics and measurement software – Gamma Gear

CNSC staff wanted to develop a new, innovative way to engage youth and teach the basics of radiation. Spectral Labs, a company that specializes in the development of immersive computer-based training solutions, developed Gamma Gear for the CNSC. Gamma Gear is an educational tool on the CNSC’s learning portal. It is publicly accessible and provides a fun approach to learning concepts of radiation physics.

The objective of Gamma Gear is to engage youth and the young at heart in physics through play. Players learn the basics of radiation; they work their way through levels and discover how to detect radioactive sources, learning the fundamentals of radiation protection and how radiation is emitted from radioactive decay.

Gamma Gear is based on a training simulator developed by Spectral Labs for the Department of Energy in the United States. It was built using software known as a Realistic Adaptive Interactive Learning System (RAILS).
Originally geared to law enforcement officers and first responders, RAILS is designed to be a tool in training first responders how to use radiation detection equipment so they can adequately respond to a nuclear or radiological emergency.

For teachers looking for new ways of engaging students and using digital platforms to teach, or for CNSC representatives who want an innovative tool for outreach, Gamma Gear offers a simple and fun approach to bringing radiation physics to the classroom.

Play the Gamma Gear video game (153 MB)

RSP-699.1, Third-Party Evaluation of the Ontario Power Generation’s Development of Evacuation Time Estimates for Pickering NGS

Ontario Power Generation, which holds an operating licence from the CNSC, contracted with a U.S.-based company to conduct a traffic engineering evacuation time estimate (ETE) study for the planning zones (PZs) and various subsets of the PZs around the Pickering Nuclear Generation Station (NGS), located in the City of Pickering (Durham Region), Ontario. The commissioned ETE report, Development of Evacuation Time Estimates, describes the analysis undertaken and the results obtained by its study for the Pickering NGS. The simulation model used reflects the U.S. Nuclear Regulatory Commission (U.S. NRC) guidance provided in U.S. NUREG/CR-7002, Criteria for Development of Evacuation Time Estimate Studies.

Due to the complexity of a large-scale evacuation and its modelling, a third party was brought in to conduct a review of the Pickering report. The third party’s report summarizes its independent review and provides insight, comments and recommendations on the licensee’s study.

Among the many factors evaluated as part of the third-party review are population groups, trip generation times, and combinations of distributions for various evacuation time components, evacuation times for special facilities, use of traffic simulation models, and assessment of shadow evacuations. The review found that the initial conditions, assumptions, modelling methodology, scenarios and analysis conducted did not substantively deviate from similar studies, which utilized NUREG/CR-7002 and have been accepted by the U.S. NRC.

The ETE provided by Pickering NGS are validated against ETE times from 58 NPP sites in the U.S. The analysis found that while Pickering NGS is among one of the most highly populated sites reviewed, the ETEs are consistent with nuclear power plant sites with similar demographics. However, the review also shows deficiencies related to transit-dependent facilities and individuals who need assistance to evacuate.

Read the RSP-699.1 final report (PDF)

RSP-658.1, Technical Presentation by Oak Ridge National Laboratory – Design Overview and Operating Experience for the Molten Salt and Sodium-Cooled Fast Reactors

The CNSC and U.S. Nuclear Regulatory Commission (U.S. NRC) collaborated to support a technical seminar developed by Oak Ridge National Laboratory (ORNL). CNSC and U.S. NRC staff invited ORNL to share its research and operating experience on the Molten Salt Reactor Experiment (MSRE) and sodium-cooled fast reactors. ORNL presented a technical review of the advanced reactor design concepts and shared operating experience based on the MSRE and Argonne National Laboratory’s experimental breeder reactor (EBR-II), which operated for almost 30 years.

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 technologies that are based on research and operating experience from the EBR-II and MSRE. The CNSC supported the technical seminar to enhance its staff’s knowledge related to the advanced reactor design concepts.

ORNL delivered a two-day technical session on each advanced reactor concept. The organization compiled research and operating experience based on its research and development activities and presented on topics related to nuclear physics, engineering design and safety analysis for molten salt and sodium-cooled fast reactors.

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