Research report summaries 2001–2002

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RSP-0134 – Establish a sound process for determining the adequacy of the design of uranium facilities

Hazard identification and risk assessment is a complex, often imperfect process, that when used with understanding, can help to sort out real from trivial risks, and thus where additional risk reduction measures can bring most benefit. The process of hazard identification and risk assessment can be helpful for ranking risks as a basis for setting priorities. In short, it is a tool that provides assurance that the major hazards inherent in the design have been characterized and so facilitates risk management.

Confidence in the accuracy of a risk assessment will be improved if it is based on sound data, it contains a clear understanding of the properties of any models used and an explicit statement of the assumptions used to accommodate uncertainties when there is no scientific basis for choosing one approach over another. Ideally, any risk assessment should include an analysis of those uncertainties and their effect on the overall risk estimate. Where quantitative approaches are used, the calculated risks should be expressed as a range or distribution rather than as a single numerical estimate of risk.

Hazard identification and risk assessment is an iterative process. The process should be revisited and repeated, particularly when specific risks appear high or when new hazard or dose response or better exposure data become available.

No one methodology can claim to adequately identify all hazards, assess the associated risks and justify the safeguards for the effective elimination or control of the hazards. Integration of several methodologies is required to ensure, so far as is reasonably practicable, that hazards have been identified and risks assessed, and that safeguards are adequate. This should result in the residual risks being as low as reasonably practicable (ALARP).

The list of methodologies can be broken down into three types. Those that should be used principally for hazard identification, those that can be used for hazard identification and can also provide some risk assessment and those that should be used for risk assessment. These break down into the following:

Principally for hazard identification

  • checklists
  • hazard studies – Consequence based
  • hazard studies – Deviation based
  • hazard studies – Periodic
  • safety reviews
  • fault analysis
  • expert judgment

Hazard identification and risk assessment (coarse)

  • audits (principally identification, but can provide some validation of safeguards)
  • hazard indices (principally comparative risk assessment, but can identify areas of hazard safeguard concern i.e., against good practice)
  • fault trees and Event trees (principally use to work back from top hazardous event to identify those causes leading up to that hazardous top event. They are also used to provide a quantified assessment of likelihood to feed into QRAs)
  • environmental assessments (principally a coarse risk assessment based on potential environmental consequences, but will also identify environmental factors both moderating and escalating the consequences specific to environmental hazards)
  • human factors (principally risk assessment, but will also identify underlying human factor causes and consequences specific to human errors)

Principally for risk assessment

  • reliability (validation for reliability of safeguard systems i.e., trips and alarms)
  • risk ranking
  • hazard/risk assessment (QRA)

Hazard identification is often seen as the most important part of the overall hazard identification and risk assessment process. This is because a) to a large extent a hazard known is a hazard controlled and b) without the confidence that all hazards have been identified, how can any assessment of the risks be ensured, as only the risks associated with known hazards will be assessed?

The development of hazardous events can be broken down into the following components:

  1. Causes. These are the causes of possible deviations from intended operation.
  2. Deviation. This is the resultant deviation from intended operation following from the cause.
  3. Consequences. These are the undesired consequences resulting from the deviation from intended operation.

The hazard identification methodologies can use this break down to assist in the identification of hazards. Hazard identification is therefore possible from the bottom up – identification of causes and deviations, or from the top down – identification of undesired consequences, working back to identified possible deviations and causes.

Two of the methodologies namely, human error analysis and environmental assessments, are specific to particular hazards that are not readily assessed using other methodologies, although they have common features with other methodologies. They use in-depth knowledge to identify potential causes, mitigating and escalating factors specific to human error and environmental consequences.

Finally, a common factor in both hazard identification and risk assessment is the level of training, expertise and experience on the part of the assessor. This will have a considerable impact on the quality of the resultant assessment outputs. Most of the review sections refer to the appropriate use of an expert or a required level of expertise.

A competency model using skills, capabilities, training and experience as criteria usually gives the required characteristics of an expert. Knowledge should be deep in the chosen domain and sufficiently broad to allow articulation into other fields. People skills, though less important than domain knowledge, are nevertheless key to an expert's impact. It no good having significant domain knowledge and experience if these factors can not be imparted to and applied by less expert personnel to be incorporated into the process design. Such deep knowledge of a domain will usually have been acquired over a significant e.g., > 5–10 years, period of time.

This report attempts to provide an insight into some of the issues surrounding each of the methodologies. This is in order to enable personnel with little (or no) experience in particular methodologies to be able to review the way in which the methodologies have been applied. To question the quality of the results from assessments and the degree to which recommendations for changes are accepted and relied upon.

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RSP-0135 – Cancer surveillance in proximity to nuclear facilities – Phase 1

Some reports of excess cancer incidence/mortality near nuclear facilities have been published. Although controversy regarding potential excess childhood leukaemia appears to be largely settled, the issue of possible health risk has not been ruled out absolutely. Because of design limitations of existing studies carried out to date and the possibility of site specific emission issues, the scientific community cannot emphatically preclude the possibility of excess leukaemia or other cancers. To address the resulting continued public concerns, recommendations have been made that further epidemiological studies be carried out and that these focus on cancer incidence surveillance and modelling of radiation exposure.

A surveillance system will ensure that public concern about the health impact of nuclear facilities is addressed effectively and comprehensively. When implemented, this system will monitor cancer patterns in the vicinity of all major Canadian nuclear facilities, and will identify geographical and temporal trends in cancer incidence proximal to these facilities. The resulting information will assist both the CNSC and Health Canada in meeting their respective mandates pertaining to the protection of the health of the public.

Data resulting from this surveillance activity will be useful for ensuring the sufficiency of regulatory parameters for guiding public exposures and for addressing public concerns. It is unlikely that the results for individual sites will lead to the enhancement of scientific knowledge regarding the radiation dose-response relation because of the relatively small populations in the vicinity of most sites, the rarity of many cancer sites, and the uncertainties in exposure estimates. However, the results from similar facilities can be combined.

A pilot project by Cancer Care Ontario has been initiated for the Pickering area to determine the feasibility of the surveillance system and provide information on selecting the methods for subsequent work. The Pickering area was chosen since the nuclear power plant there has been in operation since 1971.

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RSP-0136 – Study of the potential implications of adopting single failure/severe accident licensing requirements

The Canadian approach to accident analysis, which involves single and dual failure analyses, is quite different from that used in other countries. The process used elsewhere is based in large part on the U.S.-styled design basis accidents/severe accidents approach. CNSC staff identified some areas of difficulty with the current Canadian practice and this report was commissioned to investigate the advantages and disadvantages of each approach. This information will be used as input in a CNSC staff review to determine if changes to the current Canadian practice are warranted.

The background to the current Canadian single/dual failure approach was reviewed and found to have a strong technical basis. The advantages and disadvantages of both it and the design basis accidents/severe accidents approach were reviewed and some fundamental differences identified. The current requirements for validation of accident analyses for both the single and dual failure accidents were reviewed and the approach being taken to severe accidents analyses elsewhere is discussed.

The review finds that a complete change to the design basis accidents/severe accidents approach is not recommended, but that a change in emphasis in the current approach should be considered. Recommendations are made for reclassifying the current list of accidents into three categories; design basis accidents, dual failure accidents and severe accidents. The design basis accidents for all CANDU stations should be more clearly defined and tighter compliance with requirements instituted. It is recommended that the emphasis for the study of severe accidents should be to provide a better understanding of the phenomena involved so that severe accident management guidelines can be prepared to help the operators minimize potential offsite doses. Analyses should use best estimate with uncertainty approach with recovery actions and more realistic assumptions incorporated. This approach will result in overall risk reduction by focusing the operator's attention on severe Accident prevention and on mitigation through more technically based emergency operating procedures.

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RSP-0137 – The combined effects of alpha radiation, nickel and arsenic – Exposure on human fibroblasts

Exposure to high doses of ionizing radiation is well known to increase carcinogenic risk, although this has not been demonstrated at low doses and low dose rates in either humans or animals. At low doses and low dose rates of low LET radiation, considerable evidence indicates that cells respond by increasing their capacity for DNA repair, the so-called adaptive response, a process which decreases, rather than increases, carcinogenic risk (1, 2). Exposure to high LET radiation is considered to be more hazardous per unit dose than exposure to low LET radiation. Exposure to alpha radiation constitutes about half of the dose from natural background, mainly from inhalation of radon and its daughters. Occupational exposure occurs mainly to miners, particularly uranium miners, and results from a combination of inhaled radon daughters and particulate uranium ore dust. A recent study of the risk of inhaled uranium ore dust in rats indicated that lung cancer risk from this alpha radiation exposure was not correlated with dose but was directly correlated with dose rate (3). This result suggested that the risk was dependent only on the DNA repair capacity of the lung cells, that is, the time available to completely repair the DNA damage from one alpha track before the cell received another track.

Compounds of nickel and arsenic are carcinogenic to humans. Chronic exposure to arsenic compounds, as either As(III) or As(V), leads to lung and skin cancer in humans but not experimental animals, (4). The mechanisms involved are unclear, since As does not show mutagenic responses in mammalian cells in culture (5). Arsenic does however produce chromatid type chromosomal aberrations and sister chromatid exchanges (6, 7). There are several studies indicating that the interaction with DNA repair processes might be the predominant mechanism in arsenic-induced genotoxicity, as compared with direct DNA damage (5, 6).

Nickel compounds (Ni II) are carcinogens in humans and experimental animals. DNA protein cross links as well as chromosomal aberrations are induced in mainly heterochromatic regions (8). This may be due to oxidative DNA damage resulting from the inactivation of cellular defense systems (5). Ni(II) has also been shown to interfere with the repair of UV- and X-ray induced DNA damage in human cells (9). The DNA damage enhancing effects and the DNA repair inhibition effects both occur at non-cytotoxic concentrations (5).

The two mechanisms indicated above for these metals – enhanced DNA damage and repair inhibition – could account for the carcinogenic potential of the metals. However, it should be remembered that epigenetic mechanisms, such as interference with processes involved with tumor promotion or progression, may also contribute to their action.

While the risk from alpha radiation exposure of miners is well documented, these persons may also be exposed to minerals containing Ni or As, and the possible interactive effects of these metals with high LET radiation has not been tested. The purpose of this investigation was therefore to test alpha irradiation in combination with Ni (II) or As for additive or synergistic effects in normal human fibroblast cells. The tests employed were intended to detect effects at the level of DNA or chromosomal damage and repair.

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RSP-0138 – Review of the coverage limit in the Canadian Nuclear Liability Act – Phase 1 Development of the methodology

The Nuclear Liability Act (Canada, 1970) limits the liability of nuclear reactor operators in Canada to the first $75 million of any offsite damage resulting from an accident. The Government of Canada recently decided to increase this limit to bring it in line with international trends. In preparation for this, Magellan Engineering Consultant Inc. (Magellan) was asked to examine the adequacy of the proposed new liability limit for nuclear power plants. Magellan was also asked to propose a consistent methodology for establishing liability limits for nuclear and radiological practices other than nuclear power plants and to look at how the NLA operations manual, which provides guidelines for the implementation of the NLA, should be modified in light of the review performed.

The project was divided into several phases. Phase 1 consists of developing the methodology. Subsequent phases will address the refinement and implementation of the proposed methodology. This report addresses phase 1.

The report first examines issues and challenges that are faced in trying to project the cost of accidents. While it is possible to assign values to several cost components, some issues are process-related or deal with cost quantification that is complex, subjective and intangible. To address those issues, it is proposed to hold a workshop with senior representatives from key federal departments involved in the implementation of the NLA.

The key steps in the proposed approach are as follows:

  • Choose a reference nuclear power plant.
  • Develop a list of reference accidents and reference weather scenarios for which calculating a cost is desired.
  • Identify the cost components that are to be included in the cost calculations.
  • Quantify the cost components.
  • Calculate the costs of the reference accidents.
  • Estimate the coverage fraction of the proposed maximum basic coverage (liability limit).
  • Validate and adjust the approach.
  • Identify key parameters of a revised parametric equation for determining maximum coverage requirements.
  • Perform a sensitivity analysis to those key parameters.
  • Develop and validate the parametric equations.
  • Define other categories of nuclear facilities.
  • Apply the same methodology to other facilities.

It is proposed that the operations manual be revised only once the cost evaluations and parametric equations are complete.

Several issues related to process and cost will have to be discussed before assumptions are finalized. Hence, the recommended process involves iterations and a high level of consultation.

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RSP-0139 – Analysis of Pu-239/240 and Am-241 in urine samples: Report on a method development project conducted for the Canadian Nuclear Safety Commission

The Bioassay Section of the Bureau was asked by the CNSC to develop a method for the preparation and measurement of urine samples spiked with plutonium (Pu-239 or Pu-240) and americium (Am-241). Once determined to be sufficiently accurate and precise, the method will be utilized for the subsequent establishment of an intercomparison program that will assist Internal Dosimetry Service laboratories meet the requirements of Regulatory Standard S-106 [1] for Pu/Am urinalysis.

Urine samples, preserved with concentrated HCl, were spiked with Pu-239 and Am-241 at several concentrations chosen to test the method's capability to detect and measure the urinary excretion of an intake at the 1 ALI level at various time intervals following exposure. Pu-242 and Am-243 were used as tracers, respectively, to monitor radionuclide recovery during the sample preparation process. All standards and tracers were purchased from US NIST.

Plutonium and americium were pre-concentrated together by calcium phosphate co-precipitation, and separated later by extraction chromatography using Tru-Spec columns purchased from Eichrom, Inc. While the method developed by RPB's Bioassay Section is adapted from the Eichrom procedure [2], the wet-ashing procedure, which precedes chromatography, was considerably modified through the use of a different combination of oxidants as well as much larger amounts of these reagents to achieve a more thorough oxidation of interfering materials.

Each radionuclide was prepared for alpha-spectroscopy by cerium fluoride precipitation rather than electro-deposition. The alpha-spectrometre used was EG& G ORTEC's OCTÊTE PC Alpha Spectrometer. Counting and analysis software packages used were ORTEC'S MAEStrO and AlphaVision (ver 4.01).

The mean tracer recoveries from this analytical procedure at the spike concentrations tested compare favorably with those reported by the Eichrom technique from which the present method was adapted. Eichrom Inc. [7] reported recoveries of 74 percent and 95 percent, respectively, for Am-241 and Pu 242, from "a combination of 20+ real and synthetic urine". The values reported above for the present project are comparable for Pu-239 (98.5 percent) and much higher for Am-241(98.3 percent), using natural human urine which is more difficult to work with because of the unpredictable variation of interference levels from one sample to the next.

The three-day counting time may be cumbersome but is acceptable modus operandi among industry bioassay laboratories. At concentrations closer to the MDA, we would propose doing longer counts to improve precision and bias within, of course, the usual constraints of promptness-of-feedback requirements.

The method developed here has adequate sensitivity for the establishment of a urinalysis intercomparison program for Am-241 and Type M Pu-239/240 compounds. For Type S Pu-239/240, this capability is restricted to urinary excretion corresponding to ³ 1 ALI at the times post intake discussed above.

Because a much larger fraction of intake is excreted in faeces, the alpha spectrometric procedure developed here can be used for faecal monitoring of intakes ³ 1 ALI as well as fractions of an ALI corresponding to a selected investigational level (IL). However, faecal analysis is rarely used for routine monitoring. Its main use is in the investigation of known or suspected intakes.

If urine monitoring is the preferred routine monitoring method, a more sensitive technique such as ICP/MS is recommended. Pickford and Cox [8] report a detection limit for Pu of 4 fg/L or 9.2 μBq/L. Whereas some investigators such as Wyse and Fisher [9] report the need for sample preparation similar to that undertaken in the method developed for this project, a saving in time is still realized in the measurement step where the ICP/MS technique allows a processing rate of up to 100 samples or more per day.

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RSP-0140 – Evaluation of safety technician training programs at Ontario Power Generation Inc. Bruce NGS "B" Facility

During the week of January 22, 2001 the training programs for Ontario Power Generation (OPG) Radiation Protection and Safety Technicians at Bruce NGS "B" were evaluated. For this evaluation a subset of the CNSC's "Objectives and Criteria for Regulatory Evaluation of training programs," revision 1. January 1997 was used to determine whether or not these programs followed the principles of a systematic approach to training. (SAT)

The evaluation team reviewed OPG training policies and procedures; reviewed course training material; conducted sixteen (16) interviews; observed the delivery of five (5) courses and examined the training records system.

The training programs possessed many features of SAT-based training. The technicians all had a clear idea of what is expected of them, both during training and as job incumbents. The instructors demonstrated good knowledge of the subject matter for the courses observed by the evaluation team. In general, most of the analysis, design, development, implementation and evaluation phases are in place for the courses that were reviewed.

The evaluation team found the Ontario Hydro Nuclear (OHN) procedure "Safety Technician Training and Qualification Description" (N-PROC-tr-0034) document did not reflect the existing technician training programs. This document described the previous training program for all safety technicians at OHN facilities. New technician training programs are being developed and new "Training and Qualification Descriptions" (TQD) documents for technician positions are being produced. OPG provided the evaluation team with TQD's for the radiation protection technician (N-TQD-402-00001, Rev. 01) and the nuclear conventional safety technician (N-TQD-411-00001, Rev. 00); however the emergency preparedness safety technician TQD document was not available.

Training records for the radiation protection (RP) and safety technicians are available electronically, but only the RP technician training records correspond directly to the training requirements stated in the TQD document for their specific job positions. In some cases, technicians did not meet the entry-level criteria for their position and were performing work before completing the necessary training.

The training matrix for each of the technician job types were found to be incomplete. Some of the tasks, knowledge and skills, and training courses were not identified. Overall, the training programs are not fully developed and the radiation protection and safety technician programs are being reviewed and revised.

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RSP-0141 – Physical and numerical modelling of an in-pit tailings management facility

An in-pit tailings management facility (TMF) is an excavated pit into which tailings are deposited. At decommissioning the pit is capped, the water table allowed to recover and the tailings are left submerged. The benefits of the isolated, geologically stable, and anaerobic environment are considered to outweigh the negative impacts associated with the creation of a long-term source of groundwater contamination. The performance of an in-pit TMF can thus be expressed in terms of the rate of solute expulsion from the facility and the solute concentration generated down-gradient of it. A design feature intended to impact this performance is the so called "pervious surround". A pervious surround is a pit liner of high hydraulic conductivity material which causes groundwater to be diverted around the tailings rather than through them. If no pervious surround is included, the pit is said to have a "natural surround".

Two scale models were constructed of an in-pit TMF and a volume of the surrounding host rock. The models consisted of a 0.9 m diameter cylindrical pit penetrating 0.45 m into a 2.3 m long x 1.35 m wide x 0.9 m thick volume of host rock. In one model (the EPM model) a mortar sand was used to represent a ubiquitously fractured host rock. In the other model (the DF model) the apertures between five stacked rock slabs represented four horizontal discrete fractures. The pit was lined with a 2 cm thickness of either a filter gravel to represent a pervious surround, or a mortar sand to represent a natural surround. The lined pit was then further lined with a 5 cm thickness of crushed silica containing fluorescent dye to represent solute- filled tailings. The interior of the pit was occupied by a steel drum.

Experiments were conducted to characterize groundwater flow and solute transport through the four host rock/surround combinations. To begin each experiment a steady state groundwater flow field was established across the model. Groundwater samples were then collected during the next 7-55 days with which to characterize the movement of the tracer from the tailings source to the down-gradient end of the model. Groundwater samples were also collected from the tailings at the end of each experiment with which to characterize the residual tracer distribution.

The rate at which tracer was expelled from the tailings peaked at the start of each experiment, and then fell off towards some steady-state value. In general, the performance of the models in this regard was insensitive to the type of host rock, or the type of surround. This insensitivity was attributed to the fact that the steel drum prevented advective transport of solute within and out of the tailings. Once expelled from the tailings, however, the movement of solute through the surround and host rock was strongly controlled by the different steady-state flow regimes.

The numerical model FRAC3DVS was used to simulate groundwater flow and solute transport in the four experiments. The numerical model was not able to match precisely the rate of solute expulsion from the tailings at early time, but achieved a very good match at a later time. The early time discrepancy was attributed to the use of an overly coarse discretization in the outer few mm of the tailings source. The model was not able to match precisely the movement of solute through the surround and host rock. The discrepancies reflected an inaccurately simulated groundwater flow regime, and the limitations of the dispersion term used within FRAC3DVS. Despite the observed discrepancies, however, the numerical model was extremely useful as a tool for the analysis of the physical model results, and as a tool for the further predictive modelling.

It was concluded, based on the results of the physical and numerical modelling, that solute expulsion from a pit lined with a pervious surround will be dominated by diffusion, and that the rate of the diffusive flux will decrease asymptotically with time. In contrast, solute expulsion from a pit with a natural surround will be dominated by the rate of advective flux from the center of the tailings towards the down-gradient edge, and this rate will be relatively steady with time.

A combination numerical/analytical solution was used to estimate the rate of solute expulsion from an idealized full-scale facility with either a pervious or a natural surround. The analysis showed that when the tailings were ten times less permeable than the host rock, the rate of solute expulsion from the pervious surround system dropped below the rate for the natural surround within approximately one year after decommissioning. If the tailings were 100 times less permeable than the host rock it took 250 years for this to occur.

Further investigation was recommended to:

  1. investigate more thoroughly the process of matrix diffusion as it will effect plume migration
  2. ascertain more precisely why the numerical model was not able to better simulate the experiments
  3. investigate how geochemical phenomenon will effect solute expulsion from the tailings
  4. revisit the modelling used to predict the performance of three uranium mine in-pit TMFs currently under operation

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RSP-0142 – Survey of Canadian portable gauge users

A survey was designed to improve CNSC understanding of the work performed by portable gauge users in Canada. The survey focused on aspects of the work that will impact on the dose the users might receive from using the gauge. Issues discussed in the survey included general demographics, number of times gauges are used, transportation habits and training. Statistics were generated from the survey results and are presented in the report. The statistics are intended to aid the CNSC with future policy direction related to portable gauge users.

The premiere result of this survey is that portable gauge users reside in two groups based on usage rates. The light user group (55 percent of the sample) use the gauges less than 1000 times per year whereas the heavy user group (45 percent of the sample) use the gauge from 1556 to 20880 times per year.

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RSP-0143 – PHT system acoustics: Validation of code assumptions

This report analyzes the acoustic behaviour of a centrifugal pump in a pipeline. Current acoustic two-port models of the pump are reviewed that permit the extraction of the contribution of the pump to the acoustic far-field observed in the pipeline. A form of the acoustic scattering matrix model is selected for the current analysis on a single volute, single discharge centrifugal pump.

Verification of the elements in the matrix is carried out on the test apparatus through the use of a piezoelectric, water-resistant speaker that introduces a secondary acoustic field into the pipeline. The results are in good agreement with those predicted using the full form of the acoustic transfer matrix of the pump found in the literature. Observed scatter in the data is attributed to pipeline acoustics and vibrations.

Source data for the pump is extracted using the scattering matrix for various flow rates and pump speeds. Results suggest the pump behaves as a mixed type acoustic source, displaying both monopole and dipole characteristics. Furthermore, possible acoustic feedback may lead to amplification of one or more of the sources present in the pump. Lastly, pump loading is observed to have a significant impact on the source magnitude and phase.

The report concludes that existing theoretical models to predict the source behaviour of a centrifugal pump are inadequate based on the above observations. Recommendations for further study are provided that includes improved measurements and the development of a more precise theoretical model for the pump as an acoustic source in a pipeline.

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RSP-0144 – Community relations exploratory research focus groups

General views about nuclear energy/facilities in Canada

  • The majority of participants indicated that nuclear energy is a topic they only occasionally think about. It is not an issue uppermost in their minds or a frequent subject of conversation except on those occasions when a nuclear-related event makes the news.
  • Most participants also confessed that they have very little knowledge about nuclear energy and nuclear facilities, including a few who said they did not know that nuclear energy is used in Canada or that we have nuclear plants in Canada.
  • Many felt they had no direct involvement with nuclear energy or facilities, i.e., it was not something in their backyards. And in fact, even those who live close to a nuclear power generating plant (i.e., Toronto, Moncton, Trois Rivieres) expressed a relatively low level of concern about any aspect of nuclear energy.
  • Many of the opinions and views expressed in the sessions seemed to be shaped by events such as Chernobyl, Three Mile Island and the recent Russian submarine incident.
  • When asked what kinds of things they think about regarding nuclear energy, for most participants the thoughts expressed took the form of concerns rather than benefits. Most of them focused on issues related to nuclear waste and nuclear plants.
  • When asked what types of information they would like to receive, many wanted answers to fundamental questions about nuclear energy ( e.g., what is nuclear energy and how does it work? What are the uses of nuclear energy? The number and location of nuclear facilities in Canada, etc.) as well as answers to more specific questions about waste disposal and nuclear plants.
  • As one would expect, given both the low level of concern and knowledge stated about nuclear energy, very few said they have ever looked for information on this topic.

Awareness/knowledge of nuclear energy regulation in Canada

  • There is no almost no awareness of the CNSC, either unaided or aided. The AECB appears to be somewhat better known although it too was just a 'name' to most participants – i.e., they could not provide any details about the organization or its responsibilities.
  • Most felt that the government was the body responsible for the regulation of the production and use of nuclear energy in Canada. However, they were almost equally split on whether the federal government (or a department of the federal government such as Natural Resources or Environment Canada) was responsible or their respective provincial government had overall responsibility. A sizeable number attributed the responsibility to their hydro provider while others said they did not know.
  • In the sessions outside of Québec, many of those who eventually settled on the federal government as the regulatory body did so through inference or deductive reasoning rather than because they knew for a fact that this was the case.
  • In Québec, far more participants clearly identified the Government of Canada as the regulatory body for nuclear energy in Canada.

Involvement with CNSC's processes

  • For many, their perceived lack of knowledge coupled with their low level of concern about nuclear energy impacts on how and the extent to which they opt to be involved with the CNSC. Specifically, many believe they cannot either contribute a great deal or even learn from involvement with the CNSC's consultation and licensing decision-making processes. Therefore, most choose a passive involvement (i.e., information-sharing) rather than what they believe to be more active roles (i.e., consultation and involvement in the licensing decision-making processes).
  • Some individuals also felt that their opinions would carry little weight in the consultation and licensing decision-making processes and thus chose not to be involved. Their reasoning was two-fold:
"Why should they listen to me given how little I know about the subject? " The perceptions that decisions have already been made prior to input from the public; "I as an individual cannot influence the outcome. There is too much at stake and too much time has already been invested by the nuclear industry. "
  • Only a few participants seemed to have had any previous experience with consultations or licensing hearings on any issue. As a result, they tended to view these activities as not only foreign to them but also quite intimidating.
  • It was also obvious from the discussions that quite a few people do not understand the role and function of a regulatory body in general. As an example of this, many wanted information that went beyond the mandate of the CNSC. Again this is not surprising in view of the limited knowledge people have, pointing to the need for the CNSC to clearly communicate the specifics and the extent of its mandate with regard to the control and regulation of nuclear energy in Canada.
  • The vast majority of participants indicated that they wanted to be involved with CNSC in the information-sharing process. Their topics of interest fall into three broad categories and each includes a long list of questions that people have:
    • Nuclear energy in Canada
    • The safety of nuclear energy
    • Controls over nuclear energy
    These are the suggested means for communicating this information:
    • a brochure/flyer sent in the mail
    • through electronic media including scientific programs on regular and specialty channels to educate the public and to keep them updated
    • the internet
    • public meetings
    • 1-800#
    • email
  • Compared to the information-sharing process, considerably fewer participants stated that they would want to be involved with the CNSC in the consultation process. The areas people identified for consultation are as follows:
    • the construction of any new facilities “in my area”, although people focused more on power plants and waste disposal sites than they did on medical/research facilities
    • the safety measures put in place at power plants to ensure that there are no accidents
    • the transportation and disposal of nuclear waste
    • the importation of nuclear waste from other countries
    • the standards set for the accountability of the industry and the reliability of people in charge of nuclear facilities
The favoured methods for consultation were:
  • public meetings/town hall type meetings
  • a commission that travels from city to city to hold public meetings – Surveys/reply forms sent in the mail
  • through a 1-800# that allows people to ask questions and provide their comments to a live operator
  • an interactive website or chat room

Relative to both the information-sharing and consultation processes, far fewer participants indicated interest in being involved with the CNSC in the licensing decision-making process.

The areas or issues participants identified for involvement with the CNSC's decision-making are:

  • the construction/licensing of new facilities involved with nuclear energy or waste disposal, most particularly those located close to their homes. Notably, the same type of discussion took place here about what people consider to be in their 'backyards' or close to home as we noted in the section on consultation. Again, there was no consensus reached on this issue.
  • safety measures in place for the use of nuclear energy and disposal of nuclear waste.
  • any project that will have an impact on the health and safety of people and the environment.

The main means participants identified for getting them involved with the licensing decision-making were largely the same as those listed for the consultation process.

Many who wanted to be involved in the decision-making process stated that since they are currently ill-informed about nuclear energy, they would expect the CNSC to provide them with a complete overview of the topic under discussion prior to their involvement in any consultation or decision-making process. They judge this to be essential. Without this background information (that they expect would present the pros and cons of a given project), they felt they would not be able to form an opinion and therefore, could not contribute to the processes.

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RSP-0145 – Study of geographic information systems as a corporate tool for CNSC

Geographic information system (GIS) technology has become a routinely important tool for many public and private sector organizations worldwide. A large portion of all data in the world has some sort of spatial aspect to it. Street addresses, branch locations, inventory items, utility plant layout, logistical traffic, events, environmental measurements are all examples of data elements that have a spatial component. GIS provides a means of studying, analyzing and visualizing spatial data in a way that is not readily accessible with other data processing tools.

The personnel of the CNSC that have put this project forward have shown significant leadership in the future development of GIS applications in the government regulatory sector and we applaud their efforts. While the CNSC has taken some advantage of GIS technology, additional possibilities do exist, but have yet to be realized.

In this study, the opportunity for evolving the current GIS implementations into a corporate tool is explored. Through staff interviews, reviews of existing documentation and discussions with other government departments that have implemented GIS, the project team has determined that opportunities for broader access to GIS functionality within the CNSC can result in benefits to each of the divisions interviewed. Benefits include improved communication capabilities through a graphic interface, increased understanding of the status of licenses, faster response to queries and emergency situations and improved management of site and measurement data, to name a few.

It is difficult to rationalize the acquisition of a tool such as GIS for any particular group within the CNSC. In the case of the CNSC, benefits are accrued in small measures of time saved, better information to staff and stakeholders, increased insight into data holdings, and new or simplified methods for integrating disparate databases. These types of benefits can be realized by a number of groups throughout the organization and when aggregated justify adopting GIS as a corporate tool. Having GIS available as a corporate tool will help the organization as a whole in the regulation of the nuclear sector through improved communication by way of enhanced visualization, as well as better management and analysis of its key data assets such as stakeholder information, environmental data and documentation.

The introduction of GIS within an organization often leads to expanded use within a short period of time. Initial requests tend to be for the visualization of data but soon migrate toward requests for geographic analysis. For this reason, the recommended approach is to create a comprehensive GIS framework to support the evolution of GIS technology use at the CNSC. The framework consists of a steering committee, a working group of trained GIS specialists, technology and data. Once established, the technical committee and the working group of GIS specialists must focus on completing a business plan that supports the expanded investment in GIS technology. To this end, it is recommended that an Intranet-based application that provides basic geographic-based visualization and analysis capabilities of the LOUIS license database be considered. For a moderate investment, such an application will have the broadest reach within the organization and will form the basis for discussions on the expanded use of GIS at the CNSC.

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RSP-0146 – 129I in the environment: Phase II – The fate of atmospheric 129I in a shallow sand aquifer system at Sturgeon Falls, Ontario, Canada

The concentrations of stable iodine and 129I have been measured in groundwaters, soil and snow at Sturgeon Falls, Ontario, Canada, in an effort to examine the transport of this fission-product radionuclide product in a natural setting. Tritium and 14CDIC in groundwaters were also measured to assess recharge of the thermonuclear bomb pulse. Groundwater infiltration at this site moves vertically downward through 35 m of fine sand and silt, to an overlying basal aquifer of coarser sand. The thermonuclear bomb-pulse is clearly preserved in the tritium profile. The 1963 peak at 12.7 mbgs provides a calculated vertical velocity of 0.35 m/year.

The concentrations of 129I in snow (8.544 x 107 atoms/L) and in the upper organic litter layer and underlying soil (4.3 x 108 and 5.6 x 107, respectively) are high due to uptake of atmospheric 129I since the thermonuclear weapons testing era. However, groundwaters at the depth of the 3H bomb pulse have 129I concentrations of only 1.9 x 106 atoms/L, values much lower than regional surface waters (107 to 109 atoms per litre). This is attributed to partitioning of 129I from infiltration waters onto soil organics during recharge. Stable iodine concentrations (< 1 ppb) do not correlate with 129I indicating that any pre-nuclear era 129I/I equilibrium no longer exists. The 14C bomb pulse in dissolved inorganic carbon (DIC) is also found to be retained in the near surface, with release to infiltrating groundwaters via oxidation of organics. A similar mechanism may be responsible for the very slow downward migration of 129I observed in the shallow groundwaters.

Tritium-free groundwater in the basal aquifer, considered to be pre-bomb in age, has 2 to 4 x 105 atoms 129I per litre (129I/I = 2.2 to 2.8 x 10-10). These values are considerably greater than published estimates of pre-bomb 129I (2 x 104 atoms/L, 129I/I = 6 x 10-13) and indicate a minor contribution from the basal granite bedrock.

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RSP-0147 – Review of CANDU steam generator fitness-for-service guidelines and Darlington life cycle management plan

The objective of the project was to determine the validity of using OPG's FFSG to evaluate and disposition flaws which do not satisfy the criteria of acceptance (CAN/CSA-N285.4-94). This was to be done by reviewing the two OPG documents and provide an expert opinion on their validity, completeness and accuracy. The tasks to be performed were as the following:

Perform a detailed review of the FFSG and Darlington LCM Plan, including:
  • verification of equations and calculated results, with emphasis on the areas of most importance to safety
  • comparison of the FFSG and Darlington LCM Plan against the requirements and guidance, with suitable consideration given to the differences between CANDU steam generators and PWR steam generators

Provide expert opinion on:

  • the adequacy of the FFSG in terms of predicting the behavior of flaws in-reactor, given the methodology has been supported by in-laboratory experiments on EDM notches and machined flaws in unirradiated steam generator tube material
  • the scientific soundness of using the FFSG as an engineering procedure to perform a condition monitoring assessment of flaws and Operational Assessment of flaws
  • the validity of the FFSG for flaw characterization and evaluation of detected flaws
  • the adequacy and engineering soundness of the FFSG' set of acceptance criteria, for both prohibiting and permitting leakage. This shall include opinion on validation data for maximum tolerable fret size (MTFS) and flaw at risk of leaking (FAROL)
  • the adequacy of the FFSG to provide structural integrity assessment of entire tube population in a nuclear generating unit
  • the adequacy of the FFSG for steam generator tube repair and material removal
  • the capability of the FFSG to account for common degradation mechanisms and degradation morphology
  • the capability of the FFSG to account for tube-to-tube material variability of flaw initiation and propagation
  • the completeness of the FFSG and capability to cover inspection performance requirements

The main conclusions developed by the work performed for this project are described in section 1. It includes responses to the set of nine specific topics listed above that were raised by CNSC regarding the FFSG. Section 2 provides introductory material describing the objectives and scope of the project, and how the report is organized. Sections 3, 4 and 5 contain results of detailed reviews of the FFSG and Darlington LCM Plan against the requirements of NEI 97-06 and its invoked set of EPRI guidance documents and against the suggestions included in AECB-INFO-0572. The main conclusions from these detailed reviews are covered in section 1. Chapter 6 contains the results of an engineering review of the Darlington LCM Plan.

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RSP-0148 – Safety report review guide for research and radioisotope production reactor facilities – Phase 2

This report describes the methodology and processes used to develop a safety report review guide for research and radioisotope production reactor facilities. The background and requirements for the guide are described and the input information used in the preparation for the guide is recorded. The trade-offs and alternatives considered during the development of the guide are described. Appendices provide details of the incorporation for the requirements of the CNSC regulations in the guide, comparisons made with other guides, the development of the hazard classification process and accident sequences for existing and hypothetical facilities. Recommendations are provided for maintaining the guide current and developing safety risk criteria.

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RSP-0149 – Ontario miners database feasibility study

A feasibility study was conducted to explore the three data files on Ontario uranium miners available at the CNSC, NDR and WSIB to determine which of the data file(s) is most suitable for the future population-based matched case-control study. The type of identifying information on the subjects in the WSIB and NDR files were surname, first given name, second given name, date of birth (year, month and day), place of birth and SIN. The status of these identifying variables was similar in both data files except for the place of birth, which was more complete in the WSIB file than in the NDR file. In addition, the WSIB file had an alternate surname and alternate given names. Both WSIB and NDR files had a very high proportion of subjects with the minimum required identifying variables for a data linkage to the case-control file (98.76 percent and 99.10 percent, respectively). However, since the case-control file had place of birth as an additional identifying variable, the data linkage to the case-control file using the WSIB file may have higher probability of data linkage than the NDR file due to availability of the type and completeness of the identifying variables.

The total number of cohorts was somewhere between 25,320 and 26,809, with the NDR file having a higher number of cohorts. On the radon exposure measurements, there were conflicting findings for the radon doses; the WSIB file had a higher radon doses than the NDR file on individual comparisons whereas the NDR file had higher radon doses than the WSIB file on aggregate comparison. However, the NDR file showed statistically significantly longer duration of radon exposure. This pattern that the NDR file having longer years of follow-up was also seen in the gamma exposure measurement analysis where there was overwhelming evidence that the subjects in the NDR file had longer years of employment than the subjects in the WSIB file. The WSIB file may be incomplete compared to the NDR file because the WSIB file was collected for another study conducted in the early 1990s.

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RSP-0150 – Assessing the uncertainty in the pressure tube ballooning predictions using safety code

The final objective of this project is to determine if the safety analysis code MUL TI-SMARTT adequately predicts the pressure tube ballooning observed in tests. This report is for the first phase of work, which was to provide a detailed work plan to achieve this objective, using a probabilistic assessment approach.

The safety analysis code, related test data, and prior analytical work, are reviewed. Prior studies that have performed deterministic comparisons of predicted values (using code) and test values are also discussed.

Each task of the work plan for the second phase of this project is explained. The work plan consists of a probabilistic evaluation of safety analysis code predictions using existing test data. Methods to test the statistical significance of grooves, hydrogen and iodine on pressure ballooning predictions are also discussed.

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