Research report summaries 2011–2012

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RSP-0272 – CSA/ASME code comparison for Class 1 vessels, piping and valves

The Multinational Design Evaluation Program (MDEP) is a multinational initiative undertaken by Canada, China, Finland, France, Japan, the Republic of Korea, the Russian Federation, South Africa, the United Kingdom, and the United States (the International Atomic Energy Agency also participates in generic MDEP activities). Canada participates at all levels of the program from policy, steering committee, and working group levels, including chairing the Codes and Standards Working Group (CSWG).

In Canada, the CSA President signed the international agreement to participate in the Codes Comparison Project under the auspices of MDEP Codes and Standards Working Group. The Code comparison work was given to the CSA Nuclear Standards Steering Committee, and assigned to the CSA N285A Technical Committee (N285A TC). Initially the lead was taken by AECL but due to their withdrawal the Canadian Nuclear Safety Commission (CNSC) decided to provide support and assigned ANRIC Enterprises Inc. to complete the work within the project time frame1.

The American Society of Mechanical Engineers (ASME) offered to manage this Code comparison activity in MDEP. It was decided to use the ASME Boiler and Pressure Vessel (BPV) Code Section III as the basis for the Code comparison, the methodology agreed to consist of performing a clause-by-clause comparison of the various national Codes and Standards against the requirements of ASME BPV Code Section III Class 1 for vessels, piping, pumps and valves. The countries involved in the comparison activity included Japan, Russia, France, Korea, United States and Canada (Standards Development Organizations Code Comparison Team -SDO CCT).

The initial draft of the Canadian SDO report, as found in, Attachment A-SDO Report, was presented to the CNSC and the SDO CCT at the February 2011 meeting in Seattle for review and comment. These comments were addressed in the revised editions of the Canadian SDO Report. The revised Canadian SDO Report was presented to the N285A TC for endorsement in May 2011 and accepted by the N285A TC at their meeting in Vancouver B.C. on June 16, 2011. The Technical Committee unanimously accepted the report with suggested changes. The attached report was updated to include the comments received from the N285A TC.

1 – This work had progressed for approximately 2 years before AECL formally declared to withdraw their support.

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RSP-0273 – Comparison of Canadian NPP design requirements with those of foreign regulators

The aim of this project was to provide support to the CNSC in the formulation of a regulatory position on the Canadian design requirements for new nuclear power plants as compared to those applied in other regulatory jurisdictions, specifically in the US, Finland, UK, France and in the European Union.

The regulatory requirements currently applied in various countries that have nuclear power programmes reflect the technology of choice for the respective reactors, the operating experience accumulated as well as the developments due to research and improvement in assessment tools and techniques. While the basic safety principles governing the design of nuclear power reactors are the same regardless of the reactor technology employed, differences arise when the regulatory requirements become prescriptive as regards the design of particular safety systems provided to prevent and / or mitigate accident scenarios which are specific for each reactor type.

During the last decade, a number of new reactor systems have been developed which include significant changes in technology when compared to the reactors currently in operation. These new reactors have been developed in observance of the regulatory requirements and industry standards of the country of origin, and more often than not difficulties arise when such a design is submitted for regulatory approval as part of the licensing process for construction in other countries, which usually have established their own national safety requirements and standards.

Although the harmonisation of nuclear safety standards at international level and the standardisation of nuclear power plants are advocated by several industry organisations and also the regulatory authorities are cooperating on this matter, this process is progressing slowly and differences exist also in the interpretation of the requirements as well as in the expectations regarding their implementation. The expectations with regard to the implementation of safety requirements on design of power reactors are usually expressed in quantitative safety criteria, the harmonisation of which has not been pursued to the same extent as that of the qualitative requirements. Moreover, the regulatory practices for independent review of the safety assessments that underpin the design of reactor safety systems are not harmonised.

The project included benchmarking of the Canadian Regulatory Documents RD-337 and RD-3102 against the requirements established by selected foreign regulators: the United States Nuclear Regulatory Commission (US NRC), the Finnish Radiation and Nuclear Safety Authority (STUK), the United Kingdom Nuclear Installations Inspectorate (NII), the French Nuclear Safety Authority (ASN) and also with the Reference Levels (RL) and Nuclear Safety Objectives (SO) for new reactors established by the Western European Nuclear Regulators' Association (WENRA) for the purpose of harmonising the regulatory requirements in the European Union. The lessons learned from the application of different requirements were also reviewed as part of the project. The regulatory documents addressed in the scope of the project are on different levels in the regulatory frameworks of the respective countries, varying from legally binding requirements to principles and guidelines used in the regulatory review.

The main differences identified in the benchmarking are related to the design measures for the protection against severe accidents, the design of the containment system, the treatment of aircraft crash, the dose acceptance criteria and safety goals, the application of the single failure criterion, the time available to the operators before their action would be required in response to accidents, the scope of the requirements on electrical systems and the inherent reactivity feedback characteristics of the reactor.

All the differences in requirements have the potential to lead to design changes for a reactor licensed in the above mentioned jurisdictions. Further design changes may arise due to the differences in regulatory review practices and criteria. Since the comparison of regulatory review practices and criteria was outside the scope of the present study, safety criteria were compared only for the cases where these were provided in the regulatory documents subjected to benchmarking. A qualitative discussion on the impact of differing requirements on the conservatism, safety benefits and costs was also provided as part of the project.

The detailed benchmarking and the findings of this study constitute a good basis for the CNSC to identify areas where further development or clarification of the Canadian regulatory requirements and review guidelines may be useful, as well as areas where particular attention should be paid if applications for licensing are submitted for reactor designs developed in accordance with the regulations and standards in force in other countries.

2 RD-310, even though not included in the project scope as defined by CNSC, was considered in the benchmark to the extent needed to minimize the number of differences that would have been identified due to foreign regulations benchmarked containing, to various extents and levels of detail, provisions for safety analyses.

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RSP-0274 – Uncertainties in calculation of kinetics parameters of CANDU cores

Modern analysis of nuclear reactor transients uses space-time reactor kinetics methods. In the Canadian Nuclear Industry, safety analysis calculations use almost exclusively the Improved Quasistatic (IQS) flux factorization method. The IQS method, like all methods based on flux factorization, relies on calculating effective point kinetics parameters, which dominate the time behavior of the flux, using adjoint-weighted integrals. The accuracy of the adjoint representation influences the accuracy of the effective kinetics parameters.

Routine full core calculations are never performed using detailed models and transport theory, but rather using a cell-homogenized model and two-group diffusion theory. This work evaluates the effect of homogenization and group condensation on the effective kinetics parameters of a CANDU core and suggests improvements to the current methodology for simulating neutronic transients of CANDU reactors.

To investigate the effect of lattice-level homogenization and group condensation, a simple model consisting of an infinite array of lattice cells is considered. The space- and energy-dependent direct flux and the corresponding adjoint are calculated in a 69-group formalism using the collision-probability code DRAGON. Firstly, the reference, many-group heterogeneous, effective delayed neutron fraction and generation time are calculated using a many-energy-group heterogeneous adjoint. Secondly, the approximate many-group, homogeneous effective delayed-neutron fraction and generation time are calculated using an approximate fine-energy-group cell-homogenized adjoint. Thirdly, the approximate two-group homogeneous effective delayed-neutron fraction and generation time are calculated using a two-group cell-homogenized adjoint. The results of the two approximations are then compared with the reference results at multiple burnup steps. Three fuel types are investigated: natural-uranium (NU) fuel, low-void reactivity (LVR) fuel and Advanced CANDU Reactor (ACR) fuel. The WIMS-D Library Update (WLUP) 69-group microscopic cross section library is used. The delayed-neutron data is obtained from ENDF/B-VI.8 using the Java-based processing code JANIS.

Results show that, depending on the fuel, the errors introduced by using a 2-group lattice-homogenized adjoint are of the order of 5% in the effective delayed neutron fractions and up to 1% in the effective neutron generation time. Errors tend to vary with burnup by approximately 1% (of the individual parameter value). If a 69-group lattice-homogenized adjoint is used instead, the errors drop to approximately 2% for the effective delayed neutron fractions and 0.5% for the effective neutron generation time. The variation of errors with burnup is also reduced. Nonetheless, the error is not reduced enough to warrant the use of such a method. Instead, a simple methodology for directly correcting the current (2-group homogeneous) values to emulate detailed adjoint weighting is proposed.

To get an estimate of the global effects of using the proposed correction factors, one can simply correct the full-core value of each parameter (obtained using a 2-group homogeneous model) by a "burnup-average" value of its correction factor. The following rough estimates of burnup-independent correction factors are suggested.

Table: Uncertainties in Calculation of Kinetics Parameters of CANDU Cores
Fuel Type Beta-K Gen-Time
NU 0.950 0.995
LVR 0.953 1.000
ACR 0.960 1.005

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RSP-0275 – OECD Piping failure data exchange project (OPDE)

The reliability of piping in commercial nuclear power plants is crucial to safe and efficient operation. In recognition of this, information on degradation and failure of piping components and systems is collected and evaluated by regulatory agencies (e.g., CNSC and NRC), international organizations (e.g., OECD/NEA and IAEA) and industry organizations worldwide. The OECD/NEA Piping failure Data Exchange (OPDE) project has successfully established an international data exchange program to share operational experience regarding piping degradation and failures with OECD member countries.

The project was formally launched in May 2002 under the auspices of the OECD/NEA. Organizations producing or regulating more than 80% of nuclear energy generation worldwide contribute data to the OECD-NEA OPDE data exchange project. The basis of the project is a relational database including information on pipe degradation and failure in light water reactors and heavy water reactors for the period 1970 to 2011.

The OPDE project aims to increase the safety and long-term reliability of commercial nuclear power plants through the sharing of operational experience in the form of piping degradation and failure events.

Member countries realize many benefits from the sharing of operational experience, both nuclear regulators and operators. The benefits include amongst others: acquiring additional information regarding operating experience with piping components and systems, life cycle management of aging components, and existing inspection technologies; increased knowledge of ongoing degradation mechanisms, mitigation strategies, regulatory concerns and approaches, and networking with industry experts.

The OPDE project is envisaged as including all possible events of interest with regard to piping failures in NPPs. The project addresses typical metallic piping components of the primary system, main process and standby safety systems, and support systems (i.e., ASME Code Class 1, 2 and 3, or equivalent, piping). It also covers non safety-related (non-Code) piping, which if leaking could lead to common-cause initiating events In summary, the OPDE database covers degradation and failure of high-energy and moderate-energy piping as well as safety-related and non-safety-related piping.

This report describes the extent of the OPDE project after 9 years of operation, and gives some insights based on close to 3800 piping failure events in the database. As of May 31st, 2011 the OPDE project was officially completed, but the success of OPDE has lead participants to create a new project entitled the “Component Operational-Experience Degradation and Ageing Programme” (CODAP), which will continue where the OPDE project finished.

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RSP-0276 – Coordinated assessment and research program on safety-related aspects of a radioactive waste repository in sedimentary rock formations: Stability of the geosphere under past and future climate change

This document represents the final report of a study initiated and financed by the Canadian Nuclear Safety Commission. It deals with the study of the impacts of past and future climate changes on Ontario's sedimentary rocks with respect to deep geologic disposal of nuclear waste (low- and intermediate-level radioactive waste).

The report includes a general review of the fundamental causes and results of long-term climate change, as well as a comprehensive, critical review and comparison of two long-term climate change models. The results show that the University of Toronto's climate change model can be used to predict the past and future climate changes in Ontario. Published geological, geochemical, mechanical, hydraulic and thermal data on sedimentary rock formations in Ontario, which are relevant to the study of the impacts of past and future glaciations, are compiled and critically analyzed. These data are used in the development of conceptual models of the study area.

A coupled thermo-hydro-mechanical-chemical model to analyze the impact of glaciations on sedimentary rocks is developed and validated. The verification and validation results show that there is a relatively good agreement between the results predicted by the model and those obtained analytically or experimentally (field data). The developed model is used to numerically analyze the effects of past and future cycles of glaciation and deglaciation on Ontario's host sedimentary rocks. Valuable results are obtained. It is found that past and future glaciations had or will have a significant impact on the pore water pressure gradient, hydraulic gradient and effective stress distribution within the sedimentary rocks in Ontario. Furthermore, the permafrost depth is found to be limited to shallow depth (£50m). It is also found that the past and future climate changes did not have and will not have significant effects on solute transport in Ontario's host rock, and that solute transport in the rock barrier (limestone, shale) is diffusion dominant. However, the simulation results show, for the case of a deteriorated shaft, that the future climate changes can lead to the development of high vertical effective stresses in the shaft. These stresses could potentially lead to further physical degradation of the shaft seal material. The simulation results reveal that a failed shaft can become a preferential path for contaminant (solute) transport.

Read the RSP-0276 Final Report (PDF)

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RSP-0277 – Numerical modelling of gas migration from a deep geologic repository in Ontario's sedimentary rocks

There is currently much active research on how gas generation and migration can affect the long-term safety of underground repositories for radioactive waste. The objective of these research activities is to better understand the mechanisms of gas generation and migration, and their impact on mechanical and hydraulic stability of the geosphere. The CNSC is contributing to this study to obtain independent tools and data for assessing gas generation and migration from proposed geological repositories.

Significant quantities of gas could be generated in underground repositories for radioactive waste from several processes, such as the degradation of waste forms or corrosion of waste containers. These gases could migrate through engineered and natural geological barrier systems. If sufficiently large, the gas pressure buildup could cause the formation of microcracks or macrocracks that would affect the integrity of the barriers and geosphere as a long-term contaminant barrier. Furthermore, these gases could significantly impact the biosphere and groundwater. Thus, the assessment of the long-term safety of a repository for nuclear waste in deep geological rock formation requires a good understanding of the mechanisms of gas generation and migration, as well as their effects on the mechanical and hydraulic stability of the geosphere.

The study consists of the following steps:

  • A review of current experimental and theoretical research on thermal-hydraulic-mechanical-chemical (TMHC) disturbances in sedimentary rock due to gas generation from a repository in sedimentary rock formations: Particular attention is given to state-of-the-art conceptual and mathematical models for simulation of gas generation and transport.
  • Compilation of published geochemical, mechanical, hydraulic and thermal data on sedimentary rock formations in Ontario, which are relevant to the study of gas generation and transport: Data are compared to those from European sedimentary rocks. The usefulness and transferability of the research results (from Europe to the Ontario situation) are compared and analyzed.
  • Development of constitutive relationships and mathematical models to determine the THMC disturbances of sedimentary rock due to gas generation from a repository: The development of the models should be calibrated with results from in-situ tests performed in underground research laboratories, such as Tournemire in France, and Mont-Terri in Switzerland.
  • Use of the developed model to perform mathematical modelling of gas generation's impact on a hypothetical repository in Ontario sedimentary rocks.

The report describes the results of the study, and will be organized as follows. Chapter 1 provides an introduction to the issues that will be addressed and the objectives of the project. Chapter 2 includes: (i) a presentation of the mechanisms of gas generation and transport in underground repositories for radioactive waste; (ii) a review of the main concepts and modelling approaches with respect to the simulation of gas generation and transport through engineered and natural barriers; (iii) a discussion and presentation of relevant case studies carried out in laboratories and/or in European underground research laboratories. Chapter 3 presents a review and analysis of the existing geological and geotechnical data from Ontario's sedimentary rocks, which are relevant to the study of gas generation and transport. Furthermore, the aforementioned data are compared with those from European sedimentary rocks. This is followed by the discussion about usefulness and transferability of the research results in Europe to the Ontario situation. Chapter 4 deals with the development and verification of a mathematical model for the prediction of gas migration from a deep geologic repository (DGR) for nuclear waste as well as for the assessment of the impacts of gas migration on sedimentary rocks. Chapter 5 presents the application of the developed mathematical model to numerically analyze the gas migration from a proposed DGR for low- and intermediate-level waste in Ontario's sedimentary rocks and the disturbance of the sedimentary rocks by the gas generation and transport. Chapter 6 presents the conclusion.

Read the RSP-0277 Final Report (PDF)

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RSP-0282 – The study of diffusion dominant solute transport in solid host rock for nuclear waste disposal

Ontario Power Generation is proposing the construction of a deep geologic repository (DGR) at the Bruce nuclear site in southern Ontario for the disposal of low- and intermediate-level nuclear waste. The repository will be constructed at a depth of about 680 m below ground surface in the Paleozoic argillaceous limestone of the Cobourg Formation.

This report presents a study supported by the Canadian Nuclear and Safety Commission on the interpretation of paleo-hydrogeological conditions at three sites, using numerical simulation of the distribution of natural tracers. The study tested the hypothesis that diffusion is the dominant means of solute transport in the low-permeability formations of the Bruce site. Researchers combined composition profiles of δ18O and δ2H measured in porewater and groundwater samples, and the understanding of the paleo-hydrogeologic conditions to test this conceptual model. To do this, they simulated groundwater flow systems developed due to hydraulic head generated from glacial loading and unloading cycles and the effects of this on advective solute transport. The hydro-mechanical loading of the Pleistocene glacial cycles on the Michigan sedimentary basin was assessed using numerical analysis of coupled stress and porewater pressure. The effects of several factors were considered in the analysis of different glacial loading scenarios, including the number of loading cycles, the effect of a wet/dry glacial-soil interface, the effect of glacial advance direction and the effect of the Cambrian aquifer on the development of anomalous pressure heads. The analyses show the change in total head with time within the formations of the Michigan basin under different loading scenarios.

The impact of glaciation and deglaciation on the groundwater flow system was investigated for single 100,000-yr cycles and multiple-cycle scenarios. The results showed high porewater pressure developed within the formations during loading periods, and followed by the development of underpressure during the interstadial periods, especially in the lowest-permeability formations. Results also showed the formations have not reached hydrostatic conditions at the present time because of loading cycles that ended around 12,000 years ago. Results illustrated the difference in generated total heads in the rock formations, between applying mechanical loads on land surface and applying an equivalent hydraulic head. The base case simulations with wet based on glacial loading cycle show that currently, regions of underpressures occur in the upper Ordovician and lower Silurian formations characterized by very low hydraulic conductivity and adjacent to the Cambrian aquifer. These results were verified by comparison to measured environmental heads from the Bruce site. To achieve this simulation result, it was necessary to allow draining from aquitards to the Cambrian layer by having an outcrop to the Cambrian layer to dissipate high porewater pressure and the use of wet base glacial loading. This study presents a tool to evaluate the v effects of future glacial events on long-term performance of a nuclear waste deep geological repository.

Furthermore, upscaling flow and solute transport parameters, measured at field or lab scale, to the large spatial and temporal scale of nuclear waste disposal in deep geological formations means the use of these parameters to model the evolution time of natural isotopes profiles. The evolution time should fall in a plausible hydro-geological range. Natural isotopes of water (δ18O and δ2H) were determined from rock samples extracted from six deep boreholes at the Bruce site (Southern Ontario, Canada) by using vacuum distillation at 150°C (Hobbs et al., 2008). Also diffusion coefficients were measured using X-ray radiographic technique and diffusion cells and hydraulic testing (pulse, slug, and drill-stem tests) for hydraulic conductivities (Raven et al., 2011). The Bruce site falls at the eastern edge of the Michigan basin. The domain Paleozoic sedimentary rocks have a thickness of 860 m and are divided into 38 different layers of a sequence of dolostones, limestones and shale (Raven et al., 2011). The domain is characterized by anomalous pressures measured in different deep boreholes. The domain is bounded by the Cambrian aquifer from the bottom and 150 m of conductive layers at the top, and characterized by conductive horizontal layers at depths of 180 and 320 m below ground surface.

A series of diffusion and advection and diffusion models was performed, and results compared with the natural isotopes profiles. Initial and boundary conditions evolution time agreed with hydro-geological history. This confirms that parameters measured at small scales are plausible for formation scale. The results also showed the important of advection on solute transport from the upper and lower boundaries. The activation time was reduced by an order of magnitude when taking the effect of advection transport.

The report has five sections: The first section presents an introduction to the issues addressed in the report and the objectives of the study. The second section presents the hydro-mechanical model analyses of the Michigan Basin under past glacial loading cycles. The third section presents the solute transport model and analyses for the Bruce site. The fourth section presents an analytical model for diffusion dominant solute transport in a finite domain that can be used to assist in the discretization of numerical models. The fifth section presents the main findings, conclusions, and recommended future work. The report also includes five appendices. The first appendix presents the verification of the hydro-mechanical model, the second shows a verification of the numerical solute transport model, the third presents the mathematical details of the new analytical model, and the fourth includes a FORtrAN model built on the analytical model and its verification. The final appendix includes raw data of natural isotopes measured in deep boreholes.

Read the RSP-0282 Final Report (PDF)

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