Canadian Nuclear Industry Research Articles

ORCA: A Tool for Radiological Consequences for Accidental Releases

Evaluating atmospheric dispersion and radiological doses in the vicinity of buildings is required for small modular reactors (SMRs) because of the reduced size of their exclusion area boundary. The current Canadian nuclear industry tool for these calculations implements the methodology defined in CSA Standard N288.2-M91, which was written to support large Canada Deuterium Uranium (CANDU) nuclear reactors as opposed to SMRs. The ORCA (On/offsite Radiological Consequences of Accidents) code has been developed to address this technical concern in addition to evaluating atmospheric dispersion and doses in the far field. The code calculates worker and public doses following an airborne release of radioactive material into the atmosphere under postulated accident conditions at a nuclear facility. The current paper presents the key assumptions and methods utilized in ORCA and discusses qualification of the software to the requirements of CSA Standard N286.7-16. The new model is applicable to SMRs and existing reactor designs and reduces conservatisms in the near field (i.e., <1 km from the source) relative to the methods in CSA N288.2-M91.

Steady-State Subchannel Thermalhydraulic Assessment of a Full-Scale Pressurized Water Reactor-Small Modular Reactor Fuel Assembly With Conventional and Advanced Fuels

Abstract Thermalhydraulic analyses using subchannel codes (e.g., ASSERT-PV) are performed as a support tool to evaluate safety margins and the key parameters. Advanced fuels have recently attracted the international community's interest to improve safety margins during normal operation and accident scenarios by utilizing special coatings and barriers in a heterogeneous, multiregion, multicoating, multiclad annular fuel. In addition, advanced fuels improve the performance characteristics such as a higher burnup and better uranium utilization. Therefore, there is a need to understand the implications of such advanced unconventional fuels for the landscape of the Canadian nuclear industry and Canadian policy for energy development. In this work, subchannel thermalhydraulic analysis of a small modular reactor based on pressurized water reactor technology (PWR-small modular reactor (SMR)) core is performed using ASSERT-PV. A benchmark of a 17 × 17 fuel assembly with conventional fuel elements in comparison to PWR-SMR specification was conducted. The advanced fuel element system is also investigated and compared with the conventional one. The results indicated that the advanced fuel achieves a significant reduction in fuel element temperature by 15%. In addition, the results revealed that the proposed advanced fuel could achieve a minimum critical heat flux ratio (MCHFR) higher than the conventional fuel by 17%.

Selling off the Crown Jewels: Socialization of Costs and Privatization of Profits in the Canadian Isotopes Industry

This article explores the privatization process of a nuclear Crown corporation, the Radiochemical Company (rcc), under Brian Mulroney’s two Progressive Conservative governments. The rcc specialized in processing and commercializing radioactive isotopes for the medical market, an industry in which it was a major global player. Between 1984 and 1991, several stakeholders, including the rcc’s parent company Atomic Energy of Canada Limited, the Ministry of Energy, Mines and Resources, the Ministry of Finance, and the Canada Development and Investment Corporation, tried to define and shape the terms of the privatization. As a result, the process did not follow a straightforward path since it triggered conflicting political and strategic interests. Through the rcc’s case, this article provides a comprehensive account of the institutional mechanisms that were at play during the application of the Conservatives’ divestiture policy. Contrary to what the government’s rhetoric suggested, privatizing the rcc did not put an end to public spending in the medical isotopes business. The opposite actually happened as the rcc’s sale resulted in socializing the costs and privatizing the profits of this lucrative branch of the Canadian nuclear industry.

Overview of hydrogen combustion experiments performed in a large scale vented vessel at Canadian Nuclear Laboratories

Hydrogen combustion inside a post-accident containment environment may pose a threat to the integrity of the containment structure and equipment. In CANDU reactors, combustion-generated overpressures in an accident will be relieved by venting to adjacent from one compartment volumes through relief panels or existing openings. To have a better understanding of the fundamental mechanisms that control the overpressures resulting from a vented hydrogen deflagration, a 120 m3 large scale vented combustion test facility was constructed at Canadian Nuclear Laboratories. A number of experiments have been completed in this facility to investigate hydrogen combustion behavior simulated in a post-accident reactor containment environment. A selected number of experimental results are presented in this paper to demonstrate several parameter effects on vented combustion dynamics by varying the number of chambers (single, two or three interconnected), volume and vent size, initial turbulence (quiescent vs. turbulent), initial temperature and steam concentration, mixture uniformity (well mixed vs. highly stratified), vent locations, and number of igniters. These results have provided useful insights for hydrogen management for the Canadian nuclear industry and CANDU utilities participating in the CANDU Owners Group research program. Most experimental data have been used for validation of combustion models and development of user guidelines.

Burnup-dependent effect of lattice-level homogenization and group condensation on calculated kinetics parameters for CANDU-type lattices

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 not 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 at different burnups, for three fuel types: natural-uranium (NU) fuel, low-void reactivity (LVR) fuel and Advanced CANDU Reactor (ACR) fuel. Results show that the use of a two-group lattice-homogenized adjoint consistently overestimates the effective delayed neutron fraction by approximately 5% for all three fuel types and over a wide burnup range. The use of a two-group lattice-homogenized adjoint also introduces errors in the effective neutron generation time, but these are at most 1.3% (and their sign changes with burnup). Errors tend to vary with burnup by approximately 1% (of the individual parameter value). If a 69-group lattice-homogenized adjoint is used, the errors drop to approximately 2% for the effective delayed neutron fractions and 0.5% for the effective neutron generation time, for all three fuel types and over the entire studied burnup range. The variation of errors with burnup is also reduced.

Assessment of ASSERT-PV for prediction of post-dryout heat transfer in CANDU bundles

Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The recently released ASSERT-PV 3.2 provides enhanced models for improved predictions of subchannel flow distribution, critical heat flux (CHF), and post-dryout (PDO) heat transfer in horizontal CANDU fuel channels. This paper presents results of an assessment of the new code version against PDO tests performed during five full-size CANDU bundle experiments conducted between 1992 and 2009 by Stern Laboratories (SL), using 28-, 37- and 43-element bundles. A total of 10 PDO test series with varying pressure-tube creep and/or bearing-pad height were analyzed. The SL experiments encompassed the bundle geometries and range of flow conditions for the intended ASSERT-PV applications for existing CANDU reactors. Code predictions of maximum PDO fuel-sheath temperature were compared against measurements from the SL PDO tests to quantify the code's prediction accuracy. The prediction statistics using the recommended model set of ASSERT-PV 3.2 were compared to those from previous code versions. Furthermore, separate-effects sensitivity studies quantified the contribution of each PDO model change or enhancement to the improvement in PDO heat transfer prediction. Overall, the assessment demonstrated significant improvement in prediction of PDO sheath temperature in horizontal fuel channels containing CANDU bundles.

Assessment of ASSERT-PV for prediction of critical heat flux in CANDU bundles

Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The recently released ASSERT-PV 3.2 provides enhanced models for improved predictions of flow distribution, critical heat flux (CHF), and post-dryout (PDO) heat transfer in horizontal CANDU fuel channels. This paper presents results of an assessment of the new code version against five full-scale CANDU bundle experiments conducted in 1990s and in 2009 by Stern Laboratories (SL), using 28-, 37- and 43-element (CANFLEX) bundles. A total of 15 CHF test series with varying pressure-tube creep and/or bearing-pad height were analyzed. The SL experiments encompassed the bundle geometries and range of flow conditions for the intended ASSERT-PV applications for CANDU reactors. Code predictions of channel dryout power and axial and radial CHF locations were compared against measurements from the SL CHF tests to quantify the code prediction accuracy. The prediction statistics using the recommended model set of ASSERT-PV 3.2 were compared to those from previous code versions. Furthermore, the sensitivity studies evaluated the contribution of each CHF model change or enhancement to the improvement in CHF prediction. Overall, the assessment demonstrated significant improvement in prediction of channel dryout power and axial and radial CHF locations in horizontal fuel channels containing CANDU bundles.

ASSERT-PV 3.2: Advanced subchannel thermalhydraulics code for CANDU fuel bundles

Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The most recent release version, ASSERT-PV 3.2 has enhanced phenomenon models for improved predictions of flow distribution, dryout power and CHF location, and post-dryout (PDO) sheath temperature in horizontal CANDU fuel bundles. The focus of the improvements is mainly on modeling considerations for the unique features of CANDU bundles such as horizontal flows, small pitch to diameter ratios, high mass fluxes, and mixed and irregular subchannel geometries, compared to PWR/BWR fuel assemblies. This paper provides a general introduction to ASSERT-PV 3.2, and describes the model changes or additions in the new version to improve predictions of flow distribution, dryout power and CHF location, and PDO sheath temperatures in CANDU fuel bundles.

Nuclear worker studies: promise and pitfalls

Nuclear worker studies: promise and pitfalls

Effect of Lattice-Level Homogenization and Group Condensation on Calculated Kinetics Parameters of a Natural-Uranium-Fueled Equilibrium CANDU Core

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 quasi-static (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 not 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 calculated effective kinetics parameters of an equilibrium CANDU core.Results show that homogenization combined with group condensation introduces a positive bias of ˜5% in the effective delayed neutron fraction over a wide range of discharge burnups. Homogenization alone induces a positive bias of only ˜2%.The bias in the effective generation time is <1% for all studied discharge burnups, and its effect on the results of a positive-reactivity transient is found to be negligible, with differences being caused solely by the effective delayed neutron fraction bias. The fractional delayed neutron fraction bias for the equilibrium core is found to be very close to that for a fresh-fuel core. However, because of the lower effective delayed neutron fraction of the equilibrium core, the effects of the bias are larger for the equilibrium core than for the fresh-fuel core. For a sample positive-reactivity transient, the maximum power is found to be underestimated by 9% for the fresh core and by 14% for the equilibrium core as a consequence of homogenization and group condensation.

The Anatomy of a Brand: Shaping Nuclear Discourse to Create the Perception of a “Clean World”

AbstractWe examine how over the past 25 years the Canadian nuclear industry has used various strategies to diminish or remove perceptions of any environmental health risks that emanate from its practice and activities. Using both industry and critical website materials, we demonstrate how the industry “removes” risk by emphasizing its own safety culture in a complex process, its green credentials, its role in producing local energy options, and the scientific ignorance of its critics, the suppression of opposing views, and a narrow risk assessment approach to potential environmental and health hazards in the marketing of a specific brand. We suggest that the same strategies are being used after the recent Japanese nuclear disaster.

Knowledge management initiatives in the Canadian nuclear industry

The Canadian nuclear industry has been faced with many challenges in the area of Knowledge Management (KM) over the past decade. Increasing levels of retirement have come at a time when the industry is facing increased demand for experienced nuclear scientists and engineers for refurbishment, decommissioning and new builds. In the mid-1990s, the industry was faced with the prospect of a growing shortage of specialist skills and the loss of important technical knowledge and experience. The recognition of the importance of the problem and the need to act prompted the Canadian nuclear utilities, Atomic Energy Canada Ltd (AECL) and the CANDU Owner's Group (COG), to undertake several proactive measures to address the problem. This paper outlines several of their initiatives that have helped to address these challenges.

Citizenship and the production of landscape and knowledge in contemporary Canadian nuclear fuel waste management

This article considers the relationship between Aboriginal peoples in Canada and the nuclear industry in the contemporary geography of Canadian Nuclear Fuel Waste management. It explores the ways in which the knowledge produced by the Canadian nuclear industry, through the work of the Nuclear Waste Management Organization about nuclear waste, its management and its effects, gains primacy over the knowledge and experience of the Serpent River First Nation (SRFN). I identify a discourse of citizenship as instrumental to pursuing and maintaining industry control over knowledge produced about this policy issue. Of particular interest is the manner in which this discourse operates to disqualify and subjugate the alternative experiences of, and knowledge about, nuclear waste and radioactivity contained in the oral histories and testimonies of the SRFN. I suggest that the discourse marginalizes the knowledge of the SRFN through the use of scaled representations of identity and place, to create a particular ‘Canadian’ account of the fuel chain and its effects.

Risky Business: Nuclear Power and Public Protest in Canada

Chapter 1 Introduction: The Democracy-Technocracy Quandary Chapter 2 Risk Assessment in the Public Policy Domain Chapter 3 Nuclear Power and Public Protest Chapter 4 The Canadian Nuclear Industry Chapter 5 The Battle Over the Pickering Nuclear Generating Station Chapter 6 David and Goliath Meet on the Shore of Lake Ontario Chapter 7 The Future of Nuclear Power in Canada

Software tools for safety-critical software development

We briefly present a software methodology for safety-critical software, developed over many years to cope with industrial safety-critical applications in the Canadian nuclear industry. Following this we present discussion on software tools that have been used to support this methodology, and software tools that could be used, but have not been used for a variety of reasons. Based on our experience, we also present and motivate a list of high-level requirements for tools that would facilitate the development of safety-critical software using the presented methods, together with a small number of tools that we believe are worth developing in the future.

Observed and Modeled Tritium Concentrations in the Terrestrial Food Chain near a Continuous Atmospheric Source

Tritium concentrations were measured in a large number of environmental and biological samples collected during 2002 at two dairy farms and a hobby farm near Pickering Nuclear Generating Station in Ontario, Canada. The data cover most compartments of the terrestrial food chain in an agricultural setting and include detailed information on the diets of the local farm animals. Ratios of plant OBT concentration to air moisture HTO varied between 0.12 and 0.56, and were generally higher for the forage crops collected at the dairy farms than for the garden vegetables sampled at the hobby farm. Animal OBT to air HTO ratios were more uniform, ranging from 0.18 to 0.45, and were generally higher for the milk and beef samples from the dairy farms than for the chicken products from the hobby farm. The observed OBT concentrations in plants and animals were compared with predictions of IMPACT, the model used by the Canadian nuclear industry to calculate annual average doses due to routine releases. The model performed well on average for the animal endpoints but overestimated concentrations in plants by a factor of 2.

Canadian Scientists and Nuclear Industry Plan Joint Research Reactor

Canadian Scientists and Nuclear Industry Plan Joint Research Reactor

Highly Reliable Systems - Hardwired or Computerized

In three cases in the Canadian nuclear industry, computerized versions of control systems encountered serious problems where similar hardwired systems had previously been completely successful. Study of these aberrations makes it possible to discern the underlying causes, which lie almost entirely in the realm of system design but are aggravated by deficiencies in conventional software practices. Remedial changes in procedure and technology are proposed

Neutron Scattering Facilities at Chalk River

ABSTRACTThe Chalk River Laboratories of AECL Research provides neutron beams for research with the NRU reactor. The NRU reactor has eight reactor loops for engineering test experiments, 30 isotope irradiation sites and beam tubes, six of which feed the neutron scattering instruments. The peak thermal flux is 3 x 1014n cm−2s−1. The neutron spectrometers are operated as national facilities for Canadian neutron scattering research. Since the research requirements for the Canadian nuclear industry are changing, and since the NRU reactor is unlikely to operate much beyond the year 2000, a new Irradiation Research Facility (IRF) is being considered for start-up in the first decade of the next century. An outline is given of this proposed new neutron source.

Development of a Human Factors Engineering Program for the Canadian Nuclear Industry

This paper discusses the development of a new planning document for the Canadian nuclear industry called the Human Factors Engineering Program Plan. This document is to provide a central location for human factors requirements, complete with mechanisms for their implementation as applied to a given project. Human factors knowledge has been incorporated to varying degrees into such areas as control centre design, maintainability and human reliability analyses; but to date there has been no broadly based, formal human factors program in the Canadian nuclear industry. The HFEPP will provide a broader, integrated approach to human factors activities. Background on the development of the plan along with format and content requirements are discussed.