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Research report summaries 2002–2003

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Archived information is provided for reference, research or recordkeeping purposes. It is not subject to the Government of Canada Web Standards and has not been altered or updated since it was archived. Please contact us to request a format other than those available.

Contractors' reports are only available in the language in which they are submitted to Canadian Nuclear Safety Commission (CNSC).

RSP-0152 – International Common Cause Data Exchange (ICDE) Project: Preparation of data to be submitted to the clearing house

This report describes the activities and methods used for Canada's initial data submission to the International Common Cause Data Exchange (ICDE) project. The International ICDE project was established by the Organization for Economic Cooperation and Development OECD to encourage multilateral cooperation in the collection and analysis of data relating to common cause failure (CCF) events at nuclear power stations. The CNSC established is own project to collect data on Canadian nuclear power plants and fulfill its obligations under the ICDE agreement. The report describes the data sources used for collecting data from Canadian nuclear power plants, the methods employed to analyse the data and code it into the ICDE database, and some summary statistics on the data.

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RSP-0153-1 – Tritium in the Canadian environment: Levels and health effects

Tritium, a radioactive form of hydrogen, occurs naturally in the environment as a result of cosmic ray interactions in the atmosphere. There is also residual tritium in the environment from the testing of nuclear weapons in the atmosphere in the 1950s and 1960s. Tritium is currently produced in nuclear reactors by the fissioning of nuclear fuel and by the neutron irradiation of heavy water used in nuclear research and power reactors. Emissions from these nuclear facilities, and from some industrial facilities that use tritium, add to the naturally and weapons-produced tritium in the environment.

Away from nuclear facilities in Canada, concentrations of tritium in air and in water are typically 0.1 becquerel per cubic metre (Bq/m3) and 5 becquerels per litre (Bq/L) respectively. In the regions around nuclear facilities these values are typically 1 Bq/m3 and 30 Bq/L. Also, the concentration of tritium in the water of some garden produce may reach 100 Bq/L. Very close to some nuclear facilities, air and water concentrations can reach 10 Bq/m3 and 100 Bq/L, respectively, with the concentrations in some locally-produced garden produce reaching 3,000 Bq/L.

The annual radiation doses to Canadians can be estimated from the measured concentrations of tritium in air, water and foods. The annual doses range from less than one microsievert up to 25 microsieverts. Most annual doses from tritium are less than a few microsieverts and few members of the general public receive doses close to the upper end of this range.

The radiation doses to Canadians from tritium in the environment, even close to nuclear facilities, are too low to have any observable health consequences.

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RSP-0153-2 – Tritium in the Canadian environment: Questions and answers

Tritium, a radioactive form of hydrogen, occurs naturally in the environment as a result of cosmic ray interactions with the atmosphere. Tritium is also present in the environment as a result of releases from nuclear weapons that were tested in the atmosphere in the 1950s and 1960s. Emissions from nuclear power and research facilities, and from some industrial facilities that use tritium, add to the naturally-occurring and weapons-produced tritium in the environment.

Concern has been expressed about the levels of tritium that are present in the Canadian environment and about possible effects on health. This brochure provides answers to a series of questions on the characteristics of tritium and its radiation, on the amounts of tritium in air, water and food as well as the radiation doses that Canadians receive from these amounts, and finally on the possible health effects from such radiation doses.

The overall conclusion is that the levels of tritium in the environment, even close to nuclear facilities, do not have any impact on public health in Canada.

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RSP-0154 – Canadian waste site database – Inventory of landfills, hazardous waste disposal sites and scrap metal yards in Canada

INTERA Engineering Ltd. (formerly Duke Engineering and Services) was retained by the CNSC to construct a database of landfills, hazardous waste disposal sites and scrap metal yards throughout Canada.

The Nuclear Safety and Control Act (NSCA) came into force 1997. The new NSCA expands the regulatory authority of CNSC, such that previously unlicensed (from a CNSC perspective) sites across Canada may come under regulatory control of CNSC. In particular landfills, hazardous waste disposal sites and scrap metal yards currently exempt from licensing requirements may require CNSC licensing. To assist the CNSC in decisions related to licensing of such facilities, INTERA has compiled an inventory of site and regulatory information pertaining to landfills, hazardous waste disposal sites and scrap metal yards in Canada called the Canadian Waste Site DataBase (CWSDB).

Compilation of the CWSDB began through contact with government agencies on the federal, provincial/territorial and municipal level to determine the regulatory agency and applicable acts or regulations governing the licencing and operation of landfills, hazardous waste disposal sites and scrap metal yards. Listings of landfills and hazardous waste disposal sites were obtained from various sources, including directly from provincial/territorial contacts, through ordering paper or electronic copies of information from government agencies, and though listings provided on the internet. Listings of scrap metal yards were not available from government agencies, and were obtained through national computerized business listings such as the 2001 Edition Select Phone CD. Database information was collected and assembled in Microsoft EXCEL spreadsheets, which were subsequently entered into the ACCESS database. The database was created in Microsoft ACCESS 2002 by Computer Business Systems, a subcontractor to INTERA.

Through contact with authorities, and review of the acts and regulations pertaining to landfills, it was determined that the licencing and operation of landfills and hazardous waste disposal sites are regulated at the provincial/territorial level, and that regulations do not reference the CNSC or the NSCA or their predecessors. Scrap metal yards appear to be largely unregulated, with the exception of municipal zoning by-laws.

The database includes 7,003 site entries for landfills, hazardous waste disposal sites and scrap metal yards. Of the 7,003 site entries, there are 6,044 landfills, 137 hazardous waste disposal sites and 822 scrap metal yards identified in the CWSDB. The CWSDB should be considered accurate and up to date with respect to open/active landfill sites, hazardous waste disposal sites and scrap metal yards. The CWSDB likely lists only a small percentage of closed sites for certain provinces, as this information was not available from provincial/territorial sources.

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RSP-0155 – Comparison and validation of methods that determine neutron dose equivalent

The neutron dose equivalent produced by photonuclear reactions in and around medical linear accelerators was studied. A series of experiments at five separate facilities, involving a total of seven accelerators was performed. The data was compared to analytical methods commonly used for estimating neutron doses in the mazes of accelerators, and was also compared to the results of Monte Carlo calculations.

Recommendations regarding appropriate techniques for estimating the neutron dose equivalent are presented, and a number of observations regarding the significance of neutron dose equivalent and factors which affect the neutron production are discussed.

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RSP-0156 – Guidelines for the assessment of electromagnetic interference in the CANDU plant – Phase 2

This study was designed to investigate and characterize in more detail than had been previously available the electric and magnetic fields associated with the higher amplitude EMI events in the frequency range from 1 kHz to 10 MHz in a CANDU nuclear power station. Although conducted EMI voltages and currents are considered equally important in this frequency range, permission was not obtained for the connection of the necessary probes to safety-related equipment terminals.

The station selected was Darlington Nuclear Generating Station (DNGS), operated by Ontario Power Generation (OPG). A computerized data acquisition (DAC) system equipped with electric and magnetic field sensors designed to respond to fields at frequencies from 1 kHz to 10 MHz was designed and was placed in DNGS for a total period of six and one half months.

The equipment was located for about two months at each of three survey locations, rooms S322, S318, and T305. The first two rooms, next to the Unit 3 control room and largely filled with cable termination and distribution racks, had been used as sites for an earlier EMI survey at the same station, while the third was selected as representing an equipment location closer to plant machinery.

The DAC equipment was configured to record real-time electric and magnetic field waveforms by concurrently triggering four analog to digital channels on an actual event. Each trigger produced a file containing four sensor voltage versus time records, each with 10,000 sample points. Two channels, one electric field and one magnetic field, were sampled at a sampling frequency of 20 Ms/s (defined as the HF channels). The remaining channels, one electric field and one magnetic field, were sampled at a sampling frequency of 1 Ms/s (defined as the LF channels). Each file was date and time stamped.

The general findings of this survey may be summarized as follows.

  • Each of the three survey locations has a distinct EMI environment, suggesting that the main sources of EMI observed in each location are close to the point of observation.
  • Electric field amplitudes are highest in the control computer equipment room (Room S322) next to the Unit 3 control room, while magnetic field amplitudes are highest near the turbine floor (Room T305).
  • All of the recorded events consisted of short (1 to 500 µs) time-limited transients occurring either in isolation or in a sequence. While there is a quasi-continuous background of transient fields, the EMI environment is bounded by a relatively small number of high-amplitude events.
  • Electric field events are mainly oscillatory transients with characteristic frequencies in the 1 to 10 MHz frequency range. However, the presence of some large unipolar impulses extends the frequency range down into the low kilohertz region.
  • Magnetic field events are oscillatory transients with characteristic frequencies occurring throughout the 1 kHz to 10 MHz range of observation. Individual transients can have a very broad frequency spectrum. The trend is to decreasing peak amplitude with increasing characteristic frequency.
  • The ratio of peak electric field to peak magnetic field for individual transients is highly variable. While this ratio is also dependent on the polarization of the incident fields relative to the sensitive axes of the sensors, the variability suggests that the larger transients are mainly observed under near field conditions.

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RSP-0157 – Power reactor regulatory standards development – Phase 1

In executing this project, three main areas were investigated:

  1. The experience of stakeholders to date with regards to standards and guidelines for nuclear power plant regulation.
  2. Areas in which standards and guides are needed for nuclear power plant regulation.
  3. The current national and international collection of standards and guides used for nuclear power plant regulation.

Collectively, this project is important in its scope of interviews, identification of standards and offering of recommendations.

In this project we have found:

Experience of stakeholders

In completing the first area of investigation, a total of 29 stakeholders were interviewed (19 internal to the CNSC and 10 external). A full range of stakeholders were contacted, with input from:

  • CNSC Directorate management
  • CNSC specialists
  • CNSC site staff
  • licensees
  • industry

Common feedback from the stakeholders was noted and summarized, and numerous recommendations are offered based on this feedback.

A number of the common concerns from the stakeholders regarding process issues surrounding the development of regulatory standards and guidelines were noted.

Need for standards and guides

The stakeholders identified a need for over 40 documents, including high-priority need for top-level documents on:

  • general safety principles
  • environmental goals
  • security

A stakeholder preference to utilize existing IAEA documents in part or in whole as the basis for developing new standards and guides was identified.

For the top-level documents, CNSC stakeholders favoured an approach which would develop documents that would cross all regulatory areas of the CNSC (i.e., were not limited in scope to power reactor regulation).

Existing national and international standards and guides

A very large number of existing documents (in excess of 400) were identified. These documents included:

  • national standards and guidelines used for CANDU reactors
  • international standards and guides used by other countries for CANDU reactors
  • international standards and guides used by other countries for other reactor types

A detailed review of each document could not be completed due to the sheer number of documents identified. Based on the stakeholder input, recommendations were focused on using existing IAEA standards and guidelines, and on identifying a process which would allow systematic identification of the documents which were needed.

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RSP-0158 – Doses to transport workers: Phase 2

The CNSC is required by legislation to develop guidelines and criteria for companies to establish a written radiation protection program for carriers, consignors and consignees of radioactive materials (RAM). This requirement comes into force on June 1, 2004. The goal of the current study, Phase 2 of a multi-phase project, is to provide information that the CNSC can use to develop these guidelines and criteria.

The major tasks of the study were to:

  • measure doses of ionizing radiation received by a selected population of transport workers over a defined period of time
  • collect shipping documents related to the period of time in which the doses are measured
  • collect dose rate records from transport companies having such records
  • collect radiation protection policies, etc. from a representative sample of companies
  • develop a correlation between the doses received and the radioactive packages transported

Seventeen companies participated in the project at 25 sites in four provinces. Over 250 workers were monitored. Participating companies included:

  • a courier company
  • 8 trucking companies
  • a provincial Department of Highways whose workers transported and used moisture gauges containing RAM
  • a manufacturer
  • shipping/receiving workers at hospitals and a university, involved in internal transport
  • air cargo terminals
  • a railway
  • a port

The report documents the methodology and results of the study. It also includes examinations and statistical analyses of several variables that may affect workers doses, e.g., the presence of a radiation protection program, the job category of the worker and the sum of the transport index of packages handled. It provides a demonstration of how recommendations of the CNSC and the International Atomic Energy Agency might be used with data such as that collected, to determine the degree of monitoring required. Reasons for the high participation rate and general success of the study are discussed, as well as some of the difficulties encountered. Recommendations are made for further studies in support of the goals of this project.

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RSP-0159 – Paleothermometry of Canadian Shield groundwaters

The origin and history of groundwaters in the crystalline environment of the Canadian Shield have safety-relevant implications for the burial of radioactive waste, and are of interest to the CNSC. Of particular interest is to determine whether the noble gas concentrations in deep shield brines hold information that could discriminate between an origin as hyper-evaporated Paleozoic seawater or brines created by refreezing of seawater in a Pleistocene periglacial setting. This report summarizes the methodology and results of a noble gas research study carried out by the University of Ottawa, financed by a research contract with teh CNSC (contract no. 99-172). The research program involved the design and building of a noble gas extraction line with quadrupole mass spectrometer analyser. The analysis design includes the addition of isotope spikes for isotope dilution calculations of noble gas concentrations

Analysis of the noble gases concentrations of groundwaters from the 3,500 level of the Con Mine show atmospheric ratios for Ne, Ar, Kr and Xe. Their concentrations correspond to recharge a recharge temperature of 5°C.

Deeper groundwaters from the 4,900 and 5,300 level show evidence of subsurface contributions for He, Ar and Xe. In particular, the results for 53-A and 49-A show a considerable noble gas over-pressuring in excess of atmospherically equilibrated components. Measured gas volumes have been corrected for contributions of atmospherically derived noble gases and are expressed as concentrations relative to those for an air-saturated brine. The brine on the 5,300 level, which represents 70 percent of the discharging water, is over-pressured in Kr by a factor of 4, in Xe by a factor of 17 and in Ar by a factor of 141. For the sample from the 4,900 level (28 percent brine) a measurement of He over-pressuring was possible, giving a concentration of He = 0.01 cm3 g-1 (851,000 x atmospheric equilibrium). Accordingly, no paleo-temperature calculations could be made with these data.

The high dissolved He concentration is near the upper limit of the range of published values for crustal fluids. The measured value normalized to 100 percent brine provides an estimated subsurface residence time for the brine of 230 ± 50 percent for a porosity range of 0.01 (0.005 to 0.02). This suggests recharge of these brines during the Paleozoic.

The lower He concentrations for sample 35-B (1.72E-4 cm3 g-1, with <1 percent brine) is nonetheless over-pressured in He by a factor of over 3000, likely due to diffusion from the rock matrix and from the high He source in deeper brines.

The 40Ar/36Ar ratio in the Con Mine brine (~3000) is one of the highest ratios measured in crustal fluids. The ratio of He to radiogenic Ar (4He/40Ar*) is 2.7, a value lower than those reported for brines in other geological settings. This over-pressuring in radiogenic 40Ar is attributed to crustal degassing with additional release of radiogenic Ar during alteration of K-bearing minerals in the Con mine rocks.

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RSP-0160 – Independent expert peer review: Canadian industry best estimate analysis and uncertainty methodology

Best estimate and uncertainty analysis methods, for design basis accidents in CANDU reactors, have been developed and demonstrated by Atomic Energy of Canada Ltd (AECL) and Ontario Power Generation Inc (OPG). The CNSC has issued a statement of their expectations for the applicability of such analyses in licensing submissions.

The CANDU Owners' Group commissioned an expert peer review of documents describing the AECL and OPG work. This report contains the peer review report on the methodology.

The reasons for moving away from a pessimistic limit of operating envelope to a best estimate and uncertainty (BEAU) method are said to be:

  • The plant operates mostly well away from the LOE conditions so LOE predictions are less realistic.
  • The LOE analysis calculates small safety margins relating to improbable plant conditions. More usual conditions have larger margins that are not quantified.
  • BEAU can give more realistic predictions at a predetermined confidence level.
  • BEAU can help resolve some outstanding safety issues by predicting more benign outcomes.
  • BEAU can help operators diagnose events by predicting more likely behavior.
  • BEAU can relax operating restrictions so reactors can run at higher power and with larger safe operating envelopes (SOEs).

The effort needed can exceed expectations. Developing a best-estimate and uncertainty method is a complex, time-consuming task which requires the integration of several different technical disciplines and the input from all interested parties including the research teams, licensees and the regulator. The Canadian approach in which all of these parties actively participate in the development, should produce a product that will be a benefit to the Canadian nuclear sector and the public. From the documentation it is clear that a substantial amount of work has been performed which is state-of-the-art for the CANDU design and that a determined start has been made. The complexity of the underlying problem and the many issues that have to be correct in the uncertainty analysis make an independent review timely. The approach being taken by Canadian parties is timely and consistent with the safety analysis approach being developed and used in the world today.

Our main findings are set out below. Specific conclusions and recommendations will be found in section 10 of this report.

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RSP-0161 – Proceedings – Second Workshop on Remediation of Idle Uranium Mines in Canada, Elliot Lake, Ontario

The purpose of the workshop was to bring together a broader group of potential stakeholders including other regulators who had an interest/involvement in remedial work at legacy uranium mines in order to:

  • provide further insight into requirements under the NSCA and regulations, and put the issues faced at a licensee's site into broader context
  • encourage a positive dialogue between the CNSC and federal provincial departments responsible for legacy uranium mine sites
  • provide CNSC staff with a better understanding of the technical, public and other issues related to legacy mine sites across Canada and specific to individual sites

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RSP-0162 – Detection of radiation damage in sub-populations of human white blood cells

Ionizing radiation induces cellular damage by directly or indirectly affecting genetic material. A damaged cell triggers a cellular response to either repair the detected damage or induce programmed cell death (apoptosis). It is now well established that ionizing radiation induces differential cell killing through apoptosis of human lymphocyte sub-populations. The neutral comet assay has been well established as a method for measuring apoptosis in human leukocytes.

In this study, the neutral comet assay was used as a comparison for developing a rapid flow cytometry method for measuring apoptosis. This method can measure apoptosis in several leukocyte subpopulations simultaneously, removing the need for separating the cell types prior to analysis. It greatly increases the speed of measuring apoptosis and has potential as a high throughput biological dosimeter.

The flow cytometry method, along with the neutral comet assay, was used to measure apoptosis in patients who had received 131I treatment. The purpose was to determine whether radiation induced apoptosis could be used as a biological indicator of damage in situations where the time of exposure was not precisely known, and sample collection may be days to months post exposure. Moreover, this investigation examined the feasibility of measuring apoptosis after exposure to internalized radio nuclides such as 131I, which deliver a chronic dose of radiation.

Unfortunately, neither the neutral comet assay nor flow cytometry was able to detect radiation-induced apoptosis after chronic exposures to radiation. These assays, however, still have potential as biological dosimeters in cases where an acute dose of radiation is received. This would be useful after an accidental exposure where a large population receives an acute exposure within a known time frame. Flow cytometry would allow rapid triage of these individuals which, would aid in determining what treatment would be useful.

The field of flow cytometry, however, offers unlimited possibilities for new methods of biological dosimetry. It is feasible that a chromosome analysis assay could be developed using flow cytometry, which would increase the speed of the classical dicentric chromosome assay. Since the damage measured with this assay is specific to radiation, it would be an ideal biological dosimeter.

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RSP-0163 – Independence expert panel review of reactor physics uncertainties

An independent expert panel reviewed three types of reactor physics parameters and their uncertainties: coolant void reactivity worth, fuel temperature reactivity coefficient and delayed neutron parameters. The intent was to examine the sufficiency and conclusiveness of the evidence provided by the Canadian nuclear sector to confirm the basis for the parameters presented to the panel. The review consists of individual reports from panel members with advice (but not necessarily consent) from the other members.

The panel could not reach a consensus regarding the bias and uncertainty in WIMS-IST calculations of coolant void reactivity worth and the fuel temperature reactivity coefficient. In both cases, one reviewer found that the assessment of the available evidence was insufficient and/or inconclusive. This conclusion was justified with detailed discussion of the industry analysis and by considering issues presented to the panel by a working group consisting of representatives of the industry and the CNSC. In both cases another reviewer found the industry position to be supported by the evidence. The arguments for both positions are found herein and it is the hope of the panel that the reader can come to an appropriate conclusion by reading the discussion of both viewpoints.

For the void reactivity one review was critical of the analysis methods: The CNl analysis is invalid for 28-element fuel and dubious for 37-element fuel. Experimental data for MOX fuel used in the analysis may not represent mid-burnup fuel found in CANDU reactors. The effect of errors introduced by using SCM has not been quantified. The analysis using MCNP should be improved by doing more calculations and by improving the analysis with existing calculations. Yet, an attempt by the reviewer at using the provided material in an alternative way (which needs refinement) has shown that the proposed biases and uncertainties for both 28- and 37-element fuel are probably reasonable.

A second review focused on the experimental methods which were found to be acceptable. The buckling change on voiding discrepancy (BCVD) was found to be remarkably good, with a standard deviation of :f:0.Ol5 m-2 for experiments with a wide range of fuel types. The conclusion is that the substitution method can be used for the 28-element UO2, 37-element UO2, and 37-element MOX fuels. It should be noted however, that the :f:0.Ol5 m-2 standard deviation corresponds to about :i:O.6 mk, which is large relative to the size of the void reactivity discrepancy (VRD) or bias of the 28-element and 37-element fuels.

A third review was in agreement with the industry recommended values for bias and uncertainty for the 37-element fuel; namely, 1.6:f:l.l mk. However, for the 28-element fuel, the preference was for 0.7:f:l.l mk rather than the industry value of 2.O:I:l.l mk. However, when one considers the uncertainties, there is no disagreement. Specifically, 0.7:f: 1.1 mk overlaps the industry value of 2.O:I:l.l mk such that the two are essentially the same within the statistical uncertainties. Hence, the industry recommended value is acceptable.

As stated above, the conclusion with regard to the bias and uncertainty used for the fuel temperature coefficient is also mixed. One reviewer concluded that the estimates of uncertainty are not justified and may be non-conservative and recommends that additional analysis be undertaken. For this reviewer, the principal problems are: the analysis for mid-burnup fuel is used as the basis for the error (both bias and uncertainty); the random component should be in absolute units (~k/K) rather than relative (%) units; the uncertainty introduced in the SCM calculation is not taken into account; and there is no analysis for 28-element fuel. The industry position is that mid-burnup fuel represents the average for a reactor whereas the reviewer expects that the bias and uncertainty should be dependent on local burnup or a bounding value should be justified. The reviewer shows that the limited number of experimental data points can be analyzed differently to come up with more conservative (larger) random errors (uncertainty).

The second reviewer did not consider the above problems to be significant and concluded that the results derived by the industry are reasonable-in particular for the random error, which is based on conservatively increasing the derived error of :1:7. 5% to :1: 10%. However, that reviewer does suggest that the industry might want to consider an even larger relative error in the FTC bias. The argument is also made that the uncertainty is almost irrelevant because the FTC is not directly used in CANDU transient analyses.

With regard to the delayed neutron data the conclusion of one reviewer is that the delayed neutron data currently used in CANDU transient simulations appears to be adequate. If new data is contemplated being used in the future, then validation exercises should be repeated. The main contributors to uncertainty have been identified and assessed relative to the delayed neutron data recommended for use in modeling CANDU transients. The approach taken compares well with international practice. Predictions of total delayed neutron fraction for CANDU core conditions have been adequately assessed and established. The resultant uncertainty in beta is on the order of 7 percent. It is recommended that additional uncertainty analyses be performed to identify the major factors contributing to this uncertainty and attempt to reduce this uncertainty to 3–4 percent.

Another reviewer is in agreement that the delayed neutron data is adequate but, rather than base this conclusion on an examination of the source of data and the way in which it is analyzed, it is based on comparisons of transient results with plant data. These measurements cover abroad range of transients and therefore are expected to be sufficient to test the delayed neutron data.

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