Research report summaries 2014–2015
Contractors' reports are only available in the language in which they are submitted to CNSC.
- RSP-0299 – Survey of design and regulatory requirements for new small reactors
- RSP-0300 – Effects of chronic exposure to alpha-emitting radionuclides on health and reproductive fitness of biota
- RSP-0300a: AECL – Effects of chronic exposure to low levels of alpha-emitting radionuclides on health and reproductive fitness of mammals: Final report
- RSP-0300b: McMaster – Final report on fish studies
- RSP-0302 – Interfacing seismological description of strong ground motion with engineering analysis of soil structure interaction for nuclear power plants
- RSP-0303 – Study on international decommissioning practice
- RSP-0304 – Incorporating aging effects into probabilistic safety assessments
- RSP-0305 – Loading of steam generator tubes during main steam line breaks
- RSP-0306 – Third party review of PRAISE-CANDU probabilistic fracture mechanics code
- RSP-0307 – Constitutive modelling of Tournemire shale
- RSP-0308 – Updated analysis of the Ontario Miners’ cohort
- RSP-0313 – Medical fitness standards for non-nuclear response force program support personnel with firearms and ammunition related duties
- RSP-0314 – Urine drug testing practices
- RSP-0315 – The forensic toxicology of alcohol and best practices for alcohol testing in the workplace
RSP-0299 – Survey of design and regulatory requirements for new small reactors
The objectives of this report are to perform a design survey of small modular reactors (SMRs) with near-term deployment potential, with a particular emphasis on identifying their innovative safety features, and to review the Canadian nuclear regulatory framework to assess whether the current and proposed regulatory documents adequately address SMR licensing challenges.
SMRs are being designed to lower the initial financing cost of a nuclear power plant or to supply electricity in small grids (often in remote areas) which cannot accommodate large nuclear power plants (NPPs). The majority of the advanced SMR designs is based on pressurized water reactor (PWR) technology, while some non-PWR Generation IV technologies (e.g., gas-cooled reactor, lead-cooled reactor, sodium-cooled fast reactor, etc.) are also being pursued. Several international nuclear technology providers are advancing their small nuclear reactor development programs; for example, standard design approval was given for the SMART reactor in South Korea, and several others are currently under construction (e.g., CAREM in Argentina, KLT-40S in Russia) or in various licensing stages (NuScale and mPower in the USA, VBER-300 in Russia). These six advanced water-cooled reactor designs were studied in this report along with a gas-cooled reactor (StarCore) which is proposed by a Canadian company.
The SMR designs, and the ways in which they incorporate inherent and passive safety characteristics, are summarized in this report. The report also identifies, to the extent that information is publicly available, the vendors’ research and development efforts in validating their designs. In the light of the Canadian Nuclear Safety Commission’s interest in how SMRs may cope with a Fukushima-type event, the author attempted to identify the vendors’ post-Fukushima responses. Caveat emptor that not all claims that were made by the vendors could be verified, since detailed SMR design information in the public domain is scarce.
The challenges, issues and considerations in licensing small nuclear reactors in Canada (from the authors’ point of view) are discussed. Then, REGDOC-2.5.2, the proposed regulatory document for new nuclear power plants, along with other related regulatory documents, was reviewed against these licensing issues.
This report asserts that the SMR designs examined here could be grouped into 2 categories regarding licensing: one category being a smaller version of conventional reactor technology with some safety design enhancements, the other category being SMRs heavily relying on inherent and passive safety characteristics. While the regulatory framework in Canada appears adequate for licensing the first category of SMRs, the second category of SMRs may pose some regulatory challenges; i.e., there are no clear instructions or protocols on how inherent and passive safety characteristics are credited.
In order to address this potential gap, the following recommendations are proposed to the CNSC:
- The existing regulatory framework is generally adequate for the large mostly-conventional SMRs such as KLT-40S and VBER-300, with a couple of major exceptions. These exceptions are a ship-based design, and the exclusion of LBLOCA from the design basis. The authors believe that these exceptions can be treated on a case-by-case basis.
- For advanced SMRs, much of REGDOC-2.5.2 applies. However, it lacks a framework for applying risk-informed regulatory judgements on innovative features. A supplement to REGDOC-2.5.2 applicable to advanced SMRs would address this gap and deal with the high-level issues identified in Section 8. The supplement should suffice to put a pre-project design review on a good foundation.
- REGDOC-2.5.2 should include requirements and guidance on applying a graded approach to advanced SMR designs.
- For advanced SMRs, some of the increased safety is achieved not through inherent means (e.g., low core power) but by engineered passive systems. These may be given much more credit in the safety case than active systems, in terms of reliability – i.e., they may not (need to) be redundant. While this might be justified on a case-by-case basis, formal guidance on credit for passive systems in SMRs would be a useful regulatory tool.
- If an operator proceeds to a construction licence, more detailed review criteria will be needed. The US is developing specific Standard Review Plans for each SMR design it is asked to review. The same model could also be used in Canada – i.e., it might be similar to the "Guidance" sections in REGDOC-2.5.2 except parts would be design-specific.
- Since all but one of the designs that are examined in this report have been developed outside of Canada, the CNSC could liaise with the regulator of the country where the design is to be deployed first, or where it has had pre-project licensing review (e.g. Korea, Argentina, and the U.S.). The purpose would be to fast-track an understanding of the design and the regulatory issues addressed in the vendor’s or first-adopter’s country.
RSP-0300 – Effects of chronic exposure to alpha-emitting radionuclides on health and reproductive fitness of biota
Atomic Energy of Canada Limited (AECL) and McMaster University formed a partnership to provide data to the Canadian Nuclear Safety Commission (CNSC) on the multigenerational effects of chronic exposure to low levels of alpha-emitting radionuclides on health and reproductive fitness of fish and mammals. The mammalian study of the research project was done at AECL and the fish study was done at McMaster University.
RSP-0300a: AECL – Effects of chronic exposure to low levels of alpha-emitting radionuclides on health and reproductive fitness of mammals: final report
This laboratory study will contribute to improve characterization of the ecological risks of the release of alpha-emitting radionuclides and will inform current policies that are relevant to uranium mining and milling in Canada. The long-term (including multigenerational) effects of chronic internal exposure to low-dose alpha-emitting radionuclides were studied using target concentrations of 0.012, 0.076, 0.78 and 8.0 becquerels/litre (Bq/L) of radium-226 (Ra-226) in drinking water. The overall objective of these experiments was to verify if such levels have the potential to cause health and/or reproductive effects when organisms are exposed for multiple generations. This multigenerational study did not reveal any detrimental effects of Ra-226 exposure (up to 8.0 Bq/L) through drinking water on health, growth and reproductive fitness of the mammals tested.
RSP-0300b: McMaster – Final report on fish studies
This document reports the results of the study of the multigenerational effects of an alpha emitter (radium-226 (Ra-226)) on fish. The species chosen was the fathead minnow (FHM), but some data are also presented for zebrafish (Zeb). The fish were exposed for their entire lifecycle to chronic doses of environmentally relevant levels (levels similar to those found in the actual environment) of radium administered in the food. The FHM lifetime experiments were extremely difficult and required two years of carefully controlled and recorded exposure. In the end no FHM breeding was achieved. The Zebs, established as a fallback due to the non-breeding of FHM, had a much shorter generation time, and it was possible to get data for F0 (first generation) and F1 (next generation) on fertility and fecundity with and without continuing Ra‑226 feeding in the F1 cohort.
Despite the lack of breeding success – which was unrelated to radium exposure – the FHM study is the first time a controlled laboratory experiment involving alpha radiation exposure has been done for the entire lifetime of FHM or any other fish species. The data were also generated following an environmentally relevant low dose exposure. This is important because many studies using toxic or potentially toxic substances use high doses and try to extrapolate back to low doses assuming a linear dose response. For instance, if a dose of 10 Bq/L results in x response, the assumption is that a dose of 1 Bq/L gives a response value one-tenth of x. However, this assumption rarely holds true in the low dose range. It was concluded that Ra-226 has no deleterious effects in FHM exposed, either acutely by injection or chronically via their diet, to environmentally relevant activities or to activities up to 1,000 times the relevant levels measured in fish. While the FHM fish did not breed, the dosimetric analyses show no Ra-226 was retained in the fish by the time they reached maturity; despite continuous feeding, the fish were free from Ra-226.
RSP-0302 – Interfacing seismological description of strong ground motion with engineering analysis of soil structure interaction for nuclear power plants
As described in the statement of the proposed work the purpose of this contract has been to formulate methodology for seismic wave excitation of numerical models of Nuclear Power Plant (NPP) structures, and specifically to provide detailed input ground motions for computations of seismic soil-structure interaction (SSI) in three dimensions (3D). This is illustrated in Fig. 1. Assuming that the structure, foundation and the surrounding soil are modeled numerically, this task can be accomplished by specifying the components of strong ground motion at a discrete mesh of points in the five surfaces of the "box" ABCD, which represents the boundary and the contact surfaces between the numerical representation of the model inside the box with the elastic, continuum mechanics representation of the site outside the box.
Traditional approach to solving the above problem has been conceptually same as what will be presented in this report, except that it was based on a simplification, which is that the seismic energy arrives to the site as vertically propagating waves (Fig. 2).
The geological structure near ground surface typically consists of "softer" deposits and therefore the seismic wave velocity decreases as the waves approach the surface. In such media, as a consequence of Snell’s Law, the ray path of body waves becomes progressively steeper (progressively closer to the vertical) (see Fig. 3). This fact has been
used to justify the simplified representation of ground shaking at a site in terms of vertically incident waves, and the associated one-dimensional (1D) site representation (Fig. 2).The wave front of a plane wave with incident angle γ, relative to vertical, will intersect the free surface at A and the vertical axis at B. As it continues to propagate, the points A and B will travel to right and up with phase velocities Cx and Cz (Fig. 4).
It is seen from Fig. 4, that for γ = 0 the horizontal phase velocity Cx will become infinite (i.e. all points on ground surface will move synchronously) and the model will reduce to 1D representation shown Fig. 2. However, γ is never zero and so the incident plane wave will arrive at the site and excite the structure by propagating horizontally with velocity Cx and with depth dependence described by cos(ωz / Cz ), as shown in Fig. 4. Therefore, a body wave pulse will move the ground surface as a horizontally propagating wave, as illustrated in Fig. 5. For nearly vertical wave arrival Cx motion will still be a horizontally propagating wave.
Soon after the arrival of body waves, surface waves will begin to arrive propagating horizontally through the wave-guide represented by parallel layers (Fig. 6).
Frequency components of surface waves will propagate with different phase velocities, which depend on the frequency of motion and the mode shape number. Those represent respectively the characteristic values and characteristic functions of the boundary value problem of wave motion in the layered half space. The wave amplitudes (i.e. the wave energy) will propagate with the frequency dependent group velocity. Near ground surface, where Cx is larger than the corresponding material velocity, the surface wave mode shapes will have periodic dependence versus depth (sines and cosines), but below the depth at which Cx becomes smaller than the material velocities the mode shape amplitudes will decay exponentially with depth (Fig. 3).
Horizontally propagating SH and Love waves will contribute torsional excitation of the foundation in addition to the out-of-plane translations. P, SV and Rayleigh waves, in addition to horizontal and vertical translations, will force the structure to rock in the (x,z) plane. This is illustrated in Fig. 7, which shows the structure excited by incident Rayleigh waves.
Strains and curvatures of ground deformation are also necessary for complete description of motions driving the walls of the numerical box ABCD in Fig. 1. Examples of computed strains and curvatures will be shown in Chapter 2.
To accomplish the task of describing the ground motion which consists of body and surface waves, and which can be used to describe the motion at all points bounding the numerical box, it is necessary to formulate algorithms which describe (a) strong ground motion along a HORIZONTAL line extending (radialy) from the earthquake source towards the site of the structure (e.g. along BC in Fig.1), and (b) along a VERTICAL line (e.g. along AB or CD in Fig. 1) in the layered half space. Once these motions have been formulated, the complete motion can be specified at any desired number of points in the five surfaces of the box ABCD.
The algorithms for description of strong earthquake motion along any horizontal line (HL) are described in Chapter 1. The essential feature of the algorithms for describing the motions along HL is the consistent use of phase delays based on the site- specific dispersion of the wave motions through the soil layers described in terms of layered half space.
For description of consistent motions along any vertical line (VL), it is necessary to work with decomposition of wave motion in terms of the frequency dependent mode shapes of body waves and of surface waves, which correspond to the characteristic functions of wave motion in the wave-guide in the layered half space. The mathematical formulation for the algorithms we have developed for the formulation of motions along VL is described in Chapter 2. At the time of this writing this method has been tested, and fully verified in terms of the basic physics of surface wave-guides. However, because it is now available for the first time, it will be necessary to continue to further refine and generalize its output formats, so that it is easily and conveniently consistent with requirements of different numerical models, which will be specified by future users. We will continue to monitor how it is used, by those who work with finite element and with finite difference models, and will further refine its output formats as required.
All of the above has been formulated for the site geology represented by parallel and flat layers. However, in a realistic geological environment, these layers can be quite irregular. Chapter 3 describes the scattering and diffraction of Love surface waves and of body SH waves by layers, which have irregular surface and interfaces. The scattered and diffracted waves, which are caused by irregular layer geometry, will interfere with the input motions and may lead to amplification and deamplification of incident motions, and to concentration of stresses and strains on and below the ground surface.
Chapter 4 extends the results of Chapter 3 for excitation by Rayleigh surface waves and body P and SV waves. These in-plane motions require more complex analyses than what is described in Chapter 3, as there are mode conversions between longitudinal P- and shear SV- wave motions.
Chapter 5 presents a manual describing the computer program SYNAC, which is used to compute the time series representing all components of ground motion (translations, rotations, strains and curvatures) for excitation of the numerical box models for SSI calculations.
RSP-0303 – Study on international decommissioning practice
Candesco – Division of Kinectrics was retained to conduct literature research on international decommissioning strategies, regulatory requirements and lessons learned from decommissioning nuclear facilities and provide a gap analysis between the current Canadian and International regulatory framework. Seven countries were included in this review: Finland, France, Germany, Italy, Sweden, the United Kingdom and the United States. International requirements and recommendations of the International Atomic Energy Agency (IAEA) and the European Commission (EC) were also considered.
The Canadian regulatory approach to the planning for decommissioning, decommissioning cost estimating and provision of funds for decommissioning is similar to the approach adopted in most of the other countries considered in this review. Canadian regulators address these matters through a combination of the use of statutory authority granted to the CNSC (financial guarantees), regulations (the requirement to include plans for decommissioning in an application for a Class I nuclear facility licence), regulatory documents (G-206 and G-219), licence conditions and code and standards (CSA N294-09 and CSA N286). Other countries also address these issues through a combination of statues (Italy), regulations (Italy and the United States), regulatory documents (Finland, Sweden and the United States), licence conditions (United Kingdom), government policy or the policies of national decommissioning agencies (France, Italy and the United Kingdom).
Some gaps were noted when Canadian regulatory practice was compared to the relevant IAEA Safety Requirements and regulatory practice in the other seven countries that were reviewed. The notable inter-country differences were between regulatory practices in Canada and the United States. The American regulatory system governing decommissioning (both NRC and DOE) is more highly developed that the system in any of the other countries considered in this review probably due to the number and variety of decommissioning projects that have already been completed in the United States.
Based on the results of this review it is recommended that:
- Licensees should be provided with guidance on the preferred or acceptable strategic approaches to decommissioning. It is recommended that this guidance should be consistent with the IAEA recommendation that "The preferred decommissioning strategy shall be immediate dismantling. There may, however, be situations where immediate dismantling is not a practical strategy when all relevant factors are considered."
- Licensees should be provided with clear guidance on if (or when) it would be acceptable to base decommissioning plans on an 'In-Situ Confinement' decommissioning strategy.
- Licensees should be required to give formal notice of permanent shutdown in advance of, or within a reasonable time after permanent shut down for decommissioning.
- The power of the CNSC to order a facility to decommission and the responsibility for decommissioning in the event that the owner/operator is unwilling or unable to perform the work should be clarified.
- A definition of the activities that may be performed under a Licence to Operate in anticipation of decommissioning should be provided.
- The schedule for submission of a Detailed Decommissioning Plan should be clarified.
- Guidance on the acceptable duration of decommissioning should be provided.
- Guidance on the Stabilization Activities and Storage and Surveillance Activities given in RD/GD-360 should be consistent with the guidance given in the CSA N294 standard. This may require revision of RD/GD-360, CSA N294-09 or both.
- Guidance on the content of a Storage with Surveillance Plan should be provided. This guidance could also be included in an amendment of the CSA N294 standard.
- Guidance on the content of an Interim End State Report (and a Characterization Report if that will be a separate document) should be provided. This guidance could also be included in an amendment of the CSA N294 standard.
- The acceptability of institutional control or restricted release following issuance of a License to Abandon and the procedures for implementing them should be clarified.
RSP-0304 – Incorporating aging effects into probabilistic safety assessments
This report supports the Canadian Nuclear Safety Commission’s (CNSC) ability to address aging effects in probabilistic safety assessments (PSAs), and in using risk-informed decision making in the area of aging management. It documents an aging PSA case study and its results; the case study was done for the CNSC using a CANDU baseline Level 1 and Level 2 PSA, along with inputs from phases 1 and 2 of the CNSC project to incorporate aging effects into PSAs.
The PSA is one of the most effective tools for risk-informed decision making. Current PSAs, however, do not explicitly account for the aging of nuclear power plant (NPP) components, which may lead to increased failure rates. Aging-related models and the effects of test and maintenance activities in controlling degradation, due to aging, of safety-critical NPP components should be taken into account in PSAs, to better reflect real situations. A PSA that explicitly incorporates aging effects and is capable of generating an age-dependent risk profile of the plant is referred to as an "aging PSA (APSA)". The APSA expands the existing PSA scope by adding new active and passive components. Current worldwide experience in the APSA area is limited. This case study is one of the first attempts to incorporate aging effects into an integrated, full-scope, plant-specific PSA model, and to observe its impact on the overall PSA results through future projections.
The case study documented in this report includes: screening of structures, systems and components (SSCs) and associated aging mechanisms (to incorporate them into the PSA model); establishing time-dependent models for component aging failure rates; PSA modeling of SSCs and related aging mechanisms in the baseline PSA provided by the CNSC; and quantification of the updated PSA model and a comparative analysis of results (with and without aging effects modelling).
A linear, time-dependent model for aging failure rates was chosen for the study. A sensitivity case done with an exponential model showed a slightly higher "severe core damage frequency" and "large release frequency" than the results based on the linear model. The key parameter in the linear, time-dependent, aging-failure-rate model is "aging failure acceleration". The aging failure accelerations for all relevant components were established. To quantitatively account for the impact of aging management activities on time-dependent failure rates, the proportional age reduction model was applied using two additional parameters: "age improvement factor" and "aging management activity time period".
The main results of PSA modeling of component aging include:
- In the reference PSA risk profile, the initiating event with the largest contribution to core damage risk is the total loss of service water.
- Results are very sensitive to changes in aging improvement factors and aging acceleration rates (i.e., the effectiveness of aging management).
- Results indicate risk importance of aging cables and the positive effect of an effective aging management program.
- Results provide only an indication – rather than a projection – of "severe core damage frequency" and "large release frequency" into the future, as they are conditional on a large number of assumptions; the most important issue is the lack of sufficient plant data for establishing plant-specific, time-dependent failure rates.
- In general, it is not considered possible to make accurate projections of quantitative risk measures into the longer-term future.
The results of this case study should be considered with caution, given the nature of the plant-specific failure rates and the uncertainties associated with the overall model.
Taking into account the lessons learned from the project, the report presents suggestions to CNSC staff for possible future work to develop a practical APSA model that could be used for quantifying the impact of aging on NPP safety. These include APSAs aimed at a) predicting the aging risk profile (significant aging risk contributors) in the reasonably near future, by using relevant operational data; and b) exploring the effectiveness of component-specific aging management programs, by using relevant operational data (e.g., from system health reports).
RSP-0305 – Loading of steam generator tubes during main steam line breaks
The overall objective of this research is to study the effects of a postulated Main Steam Line Break (MSLB) accident and the consequent steam generator blowdown on the transient loading of nuclear steam generator tubes. An improved understanding of the physical processes involved during transient two-phase fluid blowdown across tube bundles permits the development of improved design tools to ensure steam generator safety during such events.
To perform the required experiments, an experimental facility was designed and built. The experimental apparatus was equipped such that thermodynamic phenomena could be investigated through pressure and temperature measurements. In addition, dynamic load cells were installed on a model CANDU design tube bundle test section for transient tube loading measurements.
Both this experimental rig and the instrumentation system were successfully commissioned, and, following remedial steps taken to establish instrument credibility, a two-phase experimental program was developed and initiated. Using R-134a as the working fluid, measurements of temperature, pressure and tube loading, as well as simultaneous high-speed flow visualizations, were taken in conditions simulating a full-scale commercial steam generator.
This report includes an analysis of the experimental project findings to date, as well as a discussion of the strategy formulated to develop a predictive methodology for transient blowdown loading on tube bundles based on parametric investigation techniques.
RSP-0306 – Third party review of PRAISE-CANDU probabilistic fracture mechanics code
The report is a review of the documents generated as part of a probabilistic fracture mechanics code and is entitled "Third Party Review of PRAISE-CANDU Probabilistic Fracture Mechanics Code" by Lloyds Register Martec for the Canadian Nuclear Safety Commission. The objective of the work is to provide an independent third-party evaluation of the PRAISE-CANDU Probabilistic Fracture Mechanics (PFM) code development process carried out for the Composite Analytical Approach to address Large Break Loss of Coolant Accident safety margins. The scope of work involves technical evaluation of the CANDU Owner’s Group (COG) reports on PRAISE-CANDU in order to address and offer recommendations on the suitability and adequacy of the PFM code degradation models, the validation and verification (V&V) and benchmarking activities and the probabilistic simulation and uncertainty analyses techniques. In addition, a literature review of the acceptance of PFM codes in regulatory decision making in other countries is covered.
The goals of the review include: (i) determination the applicability of degradation assessment models used in PRAISE-CANDU and the potential gaps in the models; (ii) the adequacy of the theory presented to describe the basis for the selection of input parameters to ensure that consistent results would be generated by different users; (iii) the suitability of PRAISE-CANDU input parameters by answering whether the input parameters accurately describe the loading, environmental conditions which could lead to crack initiation and subsequent pipe breaks; and (iv) The suitability of the methodology used to address uncertainties in input and output. Six documents were reviewed individually and the pertinent issues emphasised in each document were studied to determine the adequacy of the PRAISE-CANDU Tool for use in regulatory decision-making.
It was concluded that the deterministic models and probabilistic calculation methods used in the development of PRAISE-CANDU are current and versatile, but additional documentation and justification of some assumptions and the selection process for some of the models would be beneficial. Further validation and verification exercises should also be undertaken. If used appropriately, the code has the potential to provide useful insights into piping system failures that can be incorporated into regulatory Risk Informed Decision Making (RIDM) activities.
An exhaustive review of the application of probabilistic fracture mechanics to nuclear power plant components and systems in other regulatory jurisdictions indicated that PFM has been used for RIDM, along with expert opinion, engineering judgement and experience, to assess the structural integrity of two components, reactor pressure vessels and the piping systems.
RSP-0307 – Constitutive modelling of Tournemire shale
This report describes the results of a project (contract # 870055-13-0331) related to constitutive modeling of the mechanical behavior of Tournemire shale. In the first part, an overview of the basic trends in the mechanical response of the shale is provided. This includes a discussion on the results of experimental tests performed under different loading conditions, including axial tension, cyclic compression and creep. The focus is on the sensitivity of material characteristics to the orientation of the bedding planes. Subsequently, the constitutive relation incorporating the notion of a scalar anisotropy parameter, which is a function of a mixed invariants of the stress and microstructure tensors, is derived. Here, an implicit integration scheme is outlined and the problem of identification of material functions/parameters is addressed in depth. The performance of the framework is verified by simulating a set of experimental results described earlier. The last part of this report is focus on the description of localized deformation that is associated with formation of macrocracks. A mathematical formulation of the problem is outlined and the proposed approach is incorporated in a finite element code. An illustrative example is provided which deals with assessment of damage formation/propagation due to excavation of a deep tunnel in Tournemire shale.
RSP-0308 – Updated analysis of the Ontario Miners’ cohort
This study is an updated analysis of mortality and cancer incidence for a cohort of Ontario uranium miners exposed to radon decay products (RDP). The cohort had been created previously, using the Ontario Mining Master File (1954–1986) and data from the National Dose Registry (1954–2004). For this update, the mortality follow-up of the cohort between 1954 and 2007 was expanded by linking to records in the Canadian Mortality Database. Similarly, cancer incidence from 1969 to 2005 was ascertained by linking to the Canadian Cancer Database at Statistics Canada. Annual exposure to radon, in working level months, was available for each cohort member.
This update examines the risk of lung cancer incidence and mortality in the Ontario uranium miners’ cohort as a result of RDP exposure. The larger cohort size provides greater precision in estimating lung cancer risk from exposure to RDP. While previous studies of this cohort have focused on lung cancer mortality, the present update also looked at cancer incidence and at studying other cancers of interest, such as stomach cancers and leukemia. Non-cancer mortality was also examined. Exploratory analyses were also conducted to investigate associations between cumulative exposure to RDP and cancer incidence and mortality for cancers other than lung. No excesses or clear dose-response relationships between RDP and other cancers were apparent. Similarly, no clear associations were seen between cardiovascular disease mortality and cumulative RDP exposure.
RSP-0313 – Medical fitness standards for non-nuclear response force program support personnel with firearms and ammunition related duties
Establishing medical, physical, and psychological employment fitness standards is vital to most occupational health and safety programs and is required under legislation in most Canadian jurisdictions. In October 2008, the Canadian Nuclear Safety Commission (CNSC) approved Regulatory Document RD-363 – Nuclear Security Officer Medical, Physical, and Psychological Fitness, which applied to all Nuclear Security Officers, including Nuclear Response Force personnel. However, RD-363 did not include non-Nuclear Response Force program support groups – such as armourers, instructor/trainers, and controlling authorities – who also perform firearms and ammunition related duties.
The primary objective of this report is to establish minimum medical fitness standards for non-Nuclear Response Force program support groups that will ensure job safety while at the same time preserve optimum productivity. Secondary objectives of the report are to recommend improvements to the current medical evaluation and selection process (if warranted), to make recommendations regarding the scope and periodicity of medical re-evaluation, and to assess the necessity for establishing specific physical and psychological fitness standards for non-Nuclear Response Force program support groups.
Representatives of several non-Nuclear Response Force program support groups at Ontario Power Generation, Bruce Power, and Atomic Energy of Canada Laboratories were interviewed and preliminary workplace analyses were completed to identify essential tasks likely to be impacted by decrements in medical fitness. A comprehensive literature review was then conducted to compare these findings to published medical standards from other national and international organizations having workers employed in similar roles. Once completed, a consensus list of recommended minimum medical standards for non-Nuclear Response Force program support groups was compiled.
A template for a Medical Examination Report for examining physicians was created to help improve the comprehensiveness, quality, and transfer of medical information between examining physicians and Company medical support teams. The Medical Examination Report included three sections: Part A – Past Medical Issues, with two sub-sections, Personal Medical History and Family Medical History; Part B – Current Medical Issues, with three sub-sections, Functional Review, Immunizations and Supplements, and Lifestyle Review; and Part C – Physician Contact, with three sub-sections, Personal Metrics, Physical Examination, and Ancillary Testing. Once completed, the Medical Examination Report will be used to standardize the medical information provided to Company medical support teams who will then render a final decision on medical fitness to Company management. Medical documentation will be retained at the medical clinic and accessed only by medical staff on an as required basis.
Finally, the requirement for specific physical and psychological fitness testing for non-Nuclear Response Force support groups, similar to that outlined in Regulatory Document RD-363 – Nuclear Security Officer Medical, Physical, and Psychological Fitness, was assessed and recommendations were provided.
RSP-0314 – Urine drug testing practices
Laboratory testing for drug use by workers in government and industry has been implemented in many countries over the past 25 – 30 years. Testing urine for drug consumption is one objective indicator of recent drug use. Urine drug testing, however, does not measure drug related impairment of a worker but does provide an indication of recent drug use. These programmes have a very specific drug testing menu and are not used to screen for all drugs which may be in a donor’s urine specimen.
Each of the components of a workplace drug testing programme have been developed in a legally defensible manner, from the specimen collection site, transportation of the specimens to the testing laboratory, receipt of specimen at the forensic laboratory, individual donor demographics, actual testing - screening and confirmation for drugs and/or metabolites.
The technical aspects of urine drug testing has a solid scientific basis and forensic laboratories performing workplace drug testing are certified by an external governmental agency in the US which provides workplace laboratory certification in the US and Canada. Rigorous quality assurance and on-site inspection teams visiting laboratories every six months ensures reliability of the testing. All aspects of this testing follows the standard approach used in forensic testing programmes of initially employing a screening test (designed to detect a specific drug or drug class) and a second (confirmation) test for all specimens that screen positive in the initial testing.
The drugs or drug classes that are generally part of workplace testing programmes include cannabinoids (marijuana), cocaine, opiates – codeine, morphine and heroin metabolite, phencyclidine and amphetamines. It is recommended that the Canadian Nuclear Safety Commission (CNSC) include these drugs or drug classes with the following exception. It is recommended that the CNSC not include phencyclidine in the testing programme due to low prevalence of this drug in Canada. Two additions to the testing programme are recommended. It is recommended that the CNSC have a broader testing menu in the opiates sub-category including - hydromorphone, hydrocodone, oxycodone and methadone. In addition, it is strongly recommended that the prescription medications – the benzodiazepines be incorporated in the workplace testing programme. The CNSC should develop a process to revise the drug menu for the drug testing programme periodically.
Due to the widespread use of drugs of abuse in our society, it is strongly recommended that CNSC develop a workplace drug testing programme as a deterrent to inappropriate drug use/abuse and to provide an objective indicator of drug use by workers in the industry.
RSP-0315 – The forensic toxicology of alcohol and best practices for alcohol testing in the workplace
Alcohol is currently the most common and serious drug that can affect safety in the workplace. The forensic toxicology of alcohol, including its absorption, distribution and elimination and blood alcohol concentration (BAC) calculations, is briefly discussed in this report. Alcohol is a depressant drug and can impair human performance at BACs as low as 20 mg/100mL. This impairment increases with increasing BAC.
The extensive scientific literature confirms that the proposed BAC limits for the workplace of 20 to 39 mg/100mL (resulting in temporary removal of a safety sensitive worker from duties) and 40 mg/100mL or greater (resulting in a policy violation and removal of the worker from duties) are scientifically valid. As shown by BAC calculations these BACs (20 mg/100mL or greater) will not affect social drinkers who have several glasses of wine with dinner or several bottles of beer in the evening and go to work the next morning.
The best and most objective method of determining impairment of human performance due to alcohol is by determining the BAC. The best method and practice of determining BACs in the workplace is by evidential breath alcohol testing. Initial screening for alcohol may be conducted rapidly and efficiently using a passive alcohol sensor.
Breath alcohol testing using an evidential breath testing instrument, operated by a qualified breath alcohol technician using the proper procedure, will provide the most reliable, rapid and noninvasive results and is the best practice for alcohol testing in the workplace. Urine alcohol testing and standardized field sobriety tests are not recommended.
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