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Research report summaries 2017–2018

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

RSP-692.1, Establishing Environmental Recovery Baselines at the Elliot Lake Historical Mine Sites

Assessment of environmental recovery requires knowledge of baseline contamination prior to initiation of activities (the status quo ante).

At Elliot Lake, Ontario, mining activities in the 1960s began without a comprehensive mine site pre-operational environmental impact study (EIS) and with limited knowledge of baseline conditions.

In this study, sedimentary core profiles are presented for use in estimating background concentrations in Lake Huron that have been altered by uranium operations upstream. Radionuclides and metals in water and sediment were measured in the impacted watershed (Serpent River) and an adjacent non-impacted watershed (Mississagi). These water and sediment concentrations will be used to derive background concentrations at the time operations were closing down.

This study demonstrates how analyses of sediment and water help in estimating historical background concentrations in an impacted watershed where no pre-operational mine site EIS was carried out. The environmental concentrations presented in this study will help in evaluating environmental recovery in the Serpent River watershed.

Read the RSP-692.1 final report (PDF)

RSP-688.1, Technical Seminar on ASME Construction Requirements for High-Temperature Reactors

Several small modular reactor (SMR) vendors have approached the CNSC with designs for reactors that would operate at temperatures higher than water-cooled reactors. Operating temperatures for these new designs can reach as high as 650°C, and under accident conditions could reach over 1,000°C. At these temperatures, structural integrity issues could lead to different failure modes and mechanisms. To address these differences, in 2011 the American Society of Mechanical Engineers (ASME) further developed Section III of its Boiler and Pressure Vessel Code to include Division 5, "High Temperature Reactors".

This technical seminar was presented to CNSC staff to enhance their expertise on important aspects of design and damage assessment of high-temperature reactors. Speakers presented an overview of Division 5 requirements, differences in design principle, and the technical basis for the code as it relates to the structural integrity of high-temperature reactors. The seminar covered safety areas of interest that are affected by material choice, fabrication and installation, testing, overpressure protection, and quality assurance for:

  • Class A metallic pressure boundary components
  • Class B metallic pressure boundary components
  • Class A and Class B metallic supports
  • Class A metallic core support structures
  • Class A non-metallic core support structures

Below are links to the seminar presentations for RSP-688.1:

RSP-613.6, Safety Assessment Code Development and Application

CNSC staff must acquire and maintain independent scientific knowledge to support informed licensing decisions and to disseminate objective information to stakeholders. Establishing the reliability of computer codes used in the safety assessment of nuclear waste repositories is an essential part of this knowledge base.

In 2015, CNSC funded a study called "Evaluation of Safety Assessment Codes for Used Fuel Disposal Facilities". The purpose of this study was to learn whether the United States Nuclear Regulatory Commission's (U.S. NRC) Scoping of Options and Analyzing Risk (SOAR1 ) computer code can be used to conduct independent scoping calculations to verify proponents' dose calculations at radioactive waste management/disposal facilities. The study concluded that, overall, SOAR is a robust assessment tool that is user friendly and flexible.

CNSC staff presented the study results at the Nuclear Waste Management Organization's Geoscience Seminar, held in Toronto from June 7 to 8, 2016, and identified a number of SOAR's limitations:

  • SOAR is not flexible enough to consider different waste package configurations and contaminant release mechanisms.
  • SOAR considers only 16 radionuclides.
  • SOAR considers only exposure through drinking water in the biosphere component. It does not include dose contributions from other pathways such as food ingestion, air inhalation, soil ingestion, external exposure to radionuclides in soil, air immersion or water immersion.
  • SOAR is not compatible with the latest version of GoldSim, the simulation software on which SOAR depends.
  • For the time being, the U.S. NRC has no plan to update SOAR.

To address these limitations, the new safety assessment code development project aims to develop a new dose calculation code based on SOAR and called DOC-WMF, and to evaluate DOC-WMF on a representative near-surface waste management facility.

The research will help CNSC staff develop the capacity to perform independent long-term safety assessments of geological repositories or other types of waste management/disposal facilities. Independent dose calculations constitute a powerful method for CNSC staff to verify key results provided by licensees and proponents in support of their applications.

The development and verification of the new DOC-WMF code can be tested on a representative near-surface waste management facility and the knowledge transferred to CNSC staff.

1 The SOAR model is a performance assessment tool intended to provide the U.S. NRC staff timely risk and performance insights for a variety of potential high-level radioactive waste disposal options.

Read the RSP-613.6 final report (PDF)

RSP-613.2, Gas Generation From Organic Waste Over a Seven-Year Period: Implication for the Management of Low- and Intermediate-Level Radioactive Waste

Deep geological repositories have been adopted in many countries for the permanent disposal of low- and intermediate-level radioactive waste (LILW). The primary focus when assessing long residence times of LILW in geological facilities is:

  • how gas pressure affects re-saturation with meteoric water of underground cavities and how these two forces affect the transport of aqueous radionuclides by groundwater and gaseous radionuclide migration through rock fractures or shaft seals
  • how microbial processes affect both the waste and the speciation and transport of radionuclides present in the waste

Over long time scales, in situ anaerobic biodegradation of the non-radiological components of the waste is expected to produce:

  • gas and volatile compounds, which could result in pressure buildup and consequently reduce the re-saturation rate of an underground cavity and delay migration of soluble radionuclides
  • acids, which can affect the initial integrity of the host bedrock and shaft seals that isolate and contain the radioactive waste

Hydrogen, carbon dioxide, methane and other volatile compounds are the gases expected to be generated by the biodegradable organic components of the waste. Over a seven-year period, project participants monitored the gas pressure evolution, headspace gas composition, and microbiology of candidate organic waste. The gas pressure evolution and changes in gas composition are interpreted according to the fungal, bacterial and archaeal composition of the candidate waste and in terms of the functional genes for methane and acetate formation. Both methane and acetate consume hydrogen and carbon dioxide, thereby reducing pressure buildup in an underground cavity.

In the experiment, methane formation appeared low while acetate formation was prevalent. In this scenario, acetogens could produce acidity and locally impact barriers, which may need to be considered in the design of shaft seals. Long-term safety cases of repositories typically have sufficient margins of safety to account for the presence of acetogens. However, including acetogenesis in safety assessments of repositories will reduce uncertainty.

Read the RSP-613.2 final report (PDF)

RSP-602.2, Phenomena Identification and Ranking Table (PKPIRT) for a Severe Accident in a CANDU Irradiated Fuel Bay

Following the Fukushima Daiichi nuclear accident, CNSC staff identified the need to improve detailed computational models in order to develop more-realistic severe accident scenarios involving an irradiated fuel bay.

The drafting of a report was the first step towards developing this model.

In 1988, the United States Nuclear Regulatory Commission developed the priority identification and ranking table (PIRT) process. It has since been routinely and widely used throughout the world as part of research and development prioritization planning and code development.

The application of the PIRT process in our study focuses on a scenario assuming a prolonged loss-of-cooling/coolant event in a CANDU irradiated fuel bay that leads to complete draining of the cooling water from the irradiated fuel bay. The period of time considered begins from normal operating conditions and ends seven days after full draining of the fuel bay.

The assessment of this application was performed by a panel of independent CANDU experts. The panel members used their extensive expertise and knowledge to develop the relevant list of phenomena for the accident, rank their relative importance and then assess the experts' current knowledge level.

Read the RSP-602.2 final report (PDF)

RSP-413.7, Modelling Thermal-Hydraulic-Mechanical-Chemical Processes in Rocks and Seals for Deep Geological Disposal

Nuclear power generation has resulted in a substantial amount of radioactive waste. Deep geological disposal being considered for long-term management of this waste relies on multiple barriers to prevent the migration of contaminants towards the biosphere.

This report summarizes a series of comprehensive investigations into the coupling of the mechanical, hydraulic, thermal and chemical factors that influence the short- and long-term behaviours of the clay seals and host rocks at a deep geological repository (DGR).

The investigations included the following:

Hydraulic-mechanical (HM) scoping assessment
Ontario Power Generation (OPG) is proposing to build a DGR for low- and intermediate-level waste (LILW). A preliminary scoping study of the short- and long-term behaviours of the proposed DGR was conducted. The numerical simulation was based on HM-coupled governing equations, with consideration of geoscientific site investigation results for various geological formations. The short-term assessment mainly addresses the excavation-induced poroelastic response of the host rock. The long-term assessment addresses the impact of glacial cycles on the stability of host rocks, vertical shafts, and horizontal waste caverns. The assessment shows that the rock formations around the DGR retain their effectiveness in containing radionuclides, both short and long term.

Thermal, hydraulic and mechanical (THM) assessment for seals and host rocks
A modelling investigation of the HE-E heater experiment, performed at the Mont Terri Underground Research Laboratory (URL) in Switzerland, was conducted. A THM-coupled model was developed to study the laboratory and field experimental observations of various physical processes. The model reproduces most of the measured temperature profile, moisture movement, and pore pressure fluctuations very well.

Visco-elastoplastic model for anisotropic shale
Using the data from creep tests, and monotonic and cyclic triaxial tests performed at the Canmet Laboratories in Ottawa, constitutive relationships were developed for the mechanical behaviour of the Tournemire shale. The model is based on the theory of plasticity, taking into consideration the inherent anisotropy due to the existence of bedding planes. The model was capable of reproducing the main physical processes that prevailed in tests.

Assessment of the HM behaviour of Cobourg limestone
Ontario Power Generation (OPG) is proposing to emplace LILW in Ordovician-aged argillaceous Cobourg limestone. The HM-coupled behaviour of Cobourg limestone was investigated, in order to better assess its long term evolution and performance as a radionuclide barrier. Using results of laboratory triaxial tests with permeability measurements, the short-term HM behaviour was interpreted with mathematical modelling. The mathematical model was also extended to take into account the long-term strength degradation of the limestone.

Hydraulic-mechanical-chemical (HMC) assessment of MX-80 bentonite seals infiltrated with brine
The pore fluids in Canadian sedimentary rocks being considered as candidate host formations for DGRs are brines with very high salinity concentrations that can substantially reduce the swelling potential of bentonite. The laboratory experiments conducted at Queen's University also indicated that the swelling pressure of MX-80 bentonite inundated with brine shows a large and early increase followed by a gradual decrease with time. A comprehensive HMC-coupled model has been developed to fully understand the underlying mechanism for these experimental observations.
A dual-porosity conceptual model was proposed in this study to reflect the transient variation of swelling pressure observed in MX-80 bentonite, when flooded with highly concentrated brine water.

HM scoping assessment – updated
Following the study on Cobourg limestone described above, an updated scoping assessment was performed for OPG's proposed DGR for LILW. Assumptions were reviewed, parameters were updated, and the scenarios were reassessed. Compared to the initial assessment, no major discrepancy in conclusions and findings is noticed in this study. The conservative scenario is confirmed to be justifiable. The excavation-induced damage zone (EDZ) extends across the Cobourg Formation into the underlying Sherman Fall Formation. The EDZ into the sidewall of the shaft appears to be limited to within one to two times the shaft diameter. The glacial cycle would not impact the mechanical stability of the host and cap formations.

Read the RSP-413.7 final report (PDF)

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