Research report summaries 2006–2007
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Contractors' reports are only available in the language in which they are submitted to the CNSC.
- RSP-0207A – Ownership of the idle uranium mine site being part of Bicroft properties in the Township of Cardiff now Bicroft in the province of Ontario
- RSP-0207B – Ownership of the idle uranium mine site being part of the Madawaska mine site in the township of Faraday in the Province of Ontario
- RSP-0208 – Fuel failure mechanisms under accident conditions
- RSP-0209A – Consultant report on general guidelines for off-site emergency preparedness and response nuclear
International Safety Research
- RSP-0209B – Consultant report on general guidelines for off-site emergency preparedness and response radiological
International Safety Research
- RSP-0210 – Radio bioassay and dose assessment of intakes of radiocarbon
CNSC Working Group on Internal Dosimetry
- RSP-0211 – International activities in development and application of best estimate analysis methods
Dr. H. Glaeser
- RSP-0212 – Technical support for implementation of IAEA periodic safety review in support of life extension of NPPs
- RSP-0213 – Ontario uranium miner update study protocol
SENES Consultants Limited
- RSP-0214 – The Canadian delegation's notes & comments to the 54th Session of UNSCEAR
- RSP-0215 – Neutrons from electron accelerators with beam energies from 6 to 10 MeV
National Research Council, C. Ross, P. Saull, D. Tchmshkyan
- RSP-0216 – Review omments on NRU severe accident studies-related reports
- RSP-0217 – Investigation into the role of pipe breaks in the licensing of CANDU reactors with positive void reactivity feedback and the credible application of early detection (leak-before-break)
- RSP-0218 – Guidance on meeting regulatory expectations for the engineering safety aspects of protection from malevolent events
ASMIS Consulting Inc.
- RSP-0219 – Review of liner and cover design and construction plans; The Port Hope Long-term Low-level Radioactive Waste Management Project preliminary engineering design report volume 1. Design and operations plan LLRWMO-1340-PDD-12001, revision 0
R. Kerry Rowe Inc.
- RSP-0220 – Scaling of RD-14M large LOCA experiments GAI 00G01: A review of COG-04-2023: A scaling assessment of RD-14M for coolant voiding during the power pulse phase of a postulated large-break LOCA scenario: Application to a 20% reactor inlet header break in the Point Lepreau reactor
N.U. Aydemir, W. Wulff
RSP-0207A – Ownership of the idle uranium mine site being part of Bicroft properties in the township of Cardiff now Bicroft in the province of Ontario
RSP-0207B – Ownership of the idle uranium mine site being part of the Madawaska mine site in the township of Faraday in the province of Ontario
The CNSC is responsible for regulating uranium mines in Canada from siting through to post-closure. Many of the mines which operated in Canada are now closed, leaving only some mines in northern Saskatchewan as active, producing mines.
In order to properly regulate these sites the current ownership needs to be understood. CNSC staff understands the ownership at most of these sites; however, the Bicroft and Madawaska sites are more complex, having had multiple ownership changes, both spatially and through time.
In these two reports, the ownership history of the two sites in question is identified including the distinction between surface and mineral leases for the sites.
Constrained axial expansion of the fuel string within a CANDU fuel channel is typically associated with fuel heatup after a large break LOCA or a fast LOR. This mechanism may create large axial stresses in the fuel element that could lead to sheath or end cap failures. The nuclear sector standard codes for modelling fuel behaviour during normal operation and a LOCA or LOR event are based on a 1D axi-symmetric model of a fuel element in which the end cap is not modelled in detail. The purpose of the present contract is to develop a methodology for conducting 2D analyses of the fuel element under large break or other tools that realistically describe the behaviour of physical processes in a component or system. LOCA and fast LOR conditions to assess possible failure in the end cap region.
In order to perform this 2D coupled thermal-mechanical analysis efficiently, a computational procedure, which involved a transient heat transfer model and a thermal-stress analysis model, has been developed. The heat transfer model was first utilized to predict the transient temperature field in the fuel element after the onset of a LOCA or a LOR event, and the thermal-stress model was then used to compute the time-dependent stresses and strains in the structure based on the transient temperature distributions. The 2D heat transfer model of the fuel element included heat conduction in both the fuel and sheath materials with temperature-dependent material properties, transient convection heat transfer to describe coolant conditions, as well as time-dependent and non-uniform energy release rate. A steady-state heat transfer analysis was first performed to generate the temperature field under normal operation. This steady-state heat transfer solution was subsequently utilized as initial condition for the transient analysis. In the thermal-stress model, the fuel stack was modelled using linear elastic material properties with temperature-dependent thermal expansion coefficient and elastic modules. The thermal mechanical behaviour of the Zircaloy-4 sheath material was characterized using a highly complicated thermal visco-plastic material model involving phase transformations and various components of creep deformations. The unique thermal strain property of the Zircaloy-4 material was taken into account. Also included in this 2D numerical model were special finite elements suitable for representing thermal and mechanical interactions between the fuel stack and the sheath/end cap materials.
In this report, the computational procedure developed in this contract for 2D transient nonlinear finite element analysis of the fuel element was first presented in detail. The temperature-dependent material properties, such as thermal conductivity, specific heat, thermal strain and elastic modulus, which were collected in the current contract for the fuel and sheath materials, were then discussed. The development and verification of the heat transfer interface element and the gap element, which were required to represent thermal and mechanical interactions between the fuel and sheath/end cap, were then presented. This presentation was followed by the implementation and validation of the thermal visco-plastic material model for high temperature behaviour of the Zircaloy-4 material. At this point, the computational capability for 2D finite element analyses of nuclear fuel elements was fully established.
In order to perform 2D finite element analyses of the fuel element in large break LOCA and fast LOR events, initial and boundary conditions must be defined. These transient conditions included time histories of coolant pressure, coolant temperature and coolant-sheath heat transfer coefficient. A power pulse history was required to describe the variation of the energy release rate in time. In this report, these initial and boundary conditions for typical large break LOCA and fast LOR cases were presented. To complete the finite element model of fuel element, transient fuel-sheath heat transfer coefficient and internal gas pressure were also required. In the present analyses, these conditions were obtained by using the industry standard ELEStrES and ELOCA programs. The ELEStrES code also provided radial distribution of the energy release rate, which was taken into account in the present simulation along with the end flux peaking phenomenon. In addition, the ELEStrES and ELOCA solutions also provided a means for partial verification of the present 2D finite element solutions.
In the present analysis, the finite element model of the end cap was based on a generic design, which was carefully constructed to preserve all the essential geometric details in a real end-cap structure. Both the fuel and sheath/end cap were discretized using the eight-noded axi-symmetric solid elements to achieve high accuracy in finite element calculations. The radial temperature distributions at various axial locations and time histories of temperatures and stresses predicted by the present 2D finite element analyses at the middle section of the fuel elements were compared with the ELEStrES and ELOCA solutions. Very good agreement was observed.
For the large break LOCA condition, the present 2D calculation predicted that the highest visco-plastic strain occurred in the sheath about 8 mm away from the sheath-end cap interface because this part of the sheath material was hotter than the sheath at the middle due to the end flux peaking, so entered the beta phase first. High stresses were also identified at the sheath-end cap interface and the inner corner of the notch in the end cap area. For the fast LOR condition, extremely high stresses and athermal strains were found at the sheath-end cap interface and inner corner of the notch, leading to potential failure at these locations.
RSP-0209A – Consultant report on general guidelines for off-site emergency preparedness and response nuclear
The guidelines proposed in this report address accidents involving large nuclear facilities in Canada or in the United States along the Canada/United States border, in particular nuclear power plants and large research reactors (such as MAPLE). They are also applicable to nuclear powered vessel accidents.
The guidelines were developed in consultation with representatives from federal and provincial emergency preparedness organizations and are intended for people and organizations that may have a key role in the development, review, audit and evaluation of emergency response plans and arrangements. The previous version of this document was extensively reviewed and discussed during a national meeting organized by the CNSC in Ottawa on 27 March 2000. The current version incorporates comments from this national meeting.
These guidelines are divided into preparedness (infrastructure) and response (functional) requirements. The requirements consist of fundamental objectives that are supported by planning and response objectives. The objectives are generic so that they can be applied to various provinces and/or different organizational structures without imposing unjustified constraints on how they can or should be met. However, in some cases, additional guidance is provided on how to achieve the objectives.
A companion document entitled General guidelines for off-site emergency preparedness and response radiological (ISR-R-1083-2) addresses accidents at smaller facilities and those involving radioactive sources.
RSP-0209B – Consultant report on general guidelines for off-site emergency preparedness and response radiological
The guidelines proposed in this report address radiological accidents other than those involving the release of fission products from large nuclear facilities. This includes, for example, accidents at research reactors such as slowpoke reactors, radioisotope storage facilities, research laboratories, pharmaceutical facilities, medical institutions, industrial non-destructive testing facilities (including mobile facilities), transportation accidents and orphan sources. It also attempts to cover special cases such as nuclear powered satellite re-entry and contamination accidents resulting from intentional or unintentional spills in the environment.
The guidelines were developed in consultation with representatives from federal and provincial emergency preparedness organizations and are intended for people and organizations that may have a key role in the development, review, audit and evaluation of offsite emergency response plans and arrangements. The previous version of this document was extensively reviewed and discussed during a federal-provincial meeting organized by the CNSC in Ottawa on 27 March 2000. The current version incorporates comments from this meeting. These guidelines are divided into preparedness (infrastructure) and response (functional) requirements. The requirements consist of fundamental objectives that are supported by planning and response objectives. The objectives are generic so that they can be applied to various provinces and/or different organizational structures without imposing unjustified constraints on how they can or should be met. However, in some cases, additional notes are provided on how to achieve the objectives. A companion document entitled General guidelines for off-site emergency preparedness and response nuclear (ISR-R-1083-1) addresses accidents specifically involving the release of fission products from large nuclear facilities (including nuclear powered vessels and large research reactors such as MAPLE).
This report describes the basis for the design, implementation and management of a radiobioassay monitoring program for intakes of radiocarbon, including the assessment of doses. It includes criteria for selecting workers who should participate in the program, and for choosing the monitoring frequency. Descriptions of the biokinetic models necessary to design the monitoring programs and to assess doses are provided in appendices. Other appendices describe quality control processes for the liquid scintillation and in-vivo counting systems most commonly used in monitoring. Worked examples demonstrate the interpretation of monitoring results in preparing the dose assessment.
Although carbon-11 is found in workplaces that prepare or research diagnostic materials, because of its very short half-life (~20 minutes) bioassay is not practical. Carbon-14, on the other hand, is generated during CANDU reactor operation, and is also used extensively in biological research. During employment as such facilities workers may be exposed to a variety of radiocarbon compounds that can be inhaled, ingested, or absorbed through intact skin or open wounds. Metabolic models for selected carbon-14 materials are presented a s a basis for the design of monitoring programmes, and the interpretation of results. Monitoring methods are described for common carbon-14 compounds, and monitoring frequencies are recommended.
The report was prepared by the Working Group on Internal Dosimetry of the Canadian Nuclear Safety Commission. The working group is composed of experts in the field of internal dosimetry from industry, health sciences and government.
RSP-0211 – International activities in development and application of best estimate analysis methods
Evaluating nuclear power plant performance during transient and accident conditions has been the main issue of safety researches in the thermal-hydraulic area carried out all over the world since the beginning of the exploitation of nuclear energy for producing electricity in the 1950s.
Safety analyses are defined as analytical investigations by, which is demonstrated how safety requirements, such as ensuring the integrity of barriers against radioactive releases and various other acceptance criteria, are met for initiating events occurring in a broad range of operational states and accident conditions, and in other circumstances, such as varying availability of the systems.
Conservative assumptions may predict misleading progression of events, unrealistic time-scales or may miss some physical phenomena. Therefore, the most safety important sequence of events may be overlooked. In addition, a conservative approach often does not show margins that in reality could be utilized for greater operational flexibility.
In order to prevent such results a best estimate approach may be better suited and the uncertainty of such a calculation should be evaluated to compare with acceptance criteria. A best estimate approach provides more realistic information about the physical behaviour, and focuses on the most relevant safety issues. For important bounding scenarios a best estimate approach with detailed uncertainty evaluation should be performed.
For scenarios with large margins to acceptance criteria it is still appropriate to use conservative analysis where no detailed evaluation of uncertainties is performed. When best estimate analyses are used to demonstrate compliance with acceptance criteria in licensing applications they must be supplemented by uncertainty evaluation. Best estimate analysis can provide information about the existing margins between calculation results and acceptance criteria. This analysis is referred to as best estimate and uncertainty (BEAU) approach. Essential for a best estimate analysis is the use of a best estimate code or other tools that realistically describe the behaviour of physical processes in a component or system.
RSP-0212 – Technical support for implementation of IAEA periodic safety review in support of life extension of NPPs
Nuclear units in Canada are now reaching the point where they must be refurbished and upgraded if they are to continue operating (life extension) or be taken out of service if this is not feasible. Since major decisions about the future of a nuclear power plant may generate public concerns, there is a need for accurate information to be available to the public about plant life extension activities.
To inform the licensees of the Canadian nuclear power plants (NPPs) and the general public about its expectations relating to licensing of NPP life extension, the CNSC has issued in 2005 a position paper that forms the basis for the regulatory approach for life extension of nuclear power plants in Canada. The key features of the regulatory approach are that the licensees perform an environmental assessment (EA), carry out an integrated safety review (ISR) in accordance with the IAEA periodic safety review (PSR) guidance and, based on the results of the EA and ISR, develop an integrated implementation plan for safety improvements.
This report was prepared to assist the CNSC in the application and implementation of the IAEA Periodic Safety Review in the regulatory oversight of nuclear power plants in Canada.
PSR is part of the regulatory system in many countries. It is a key regulatory instrument for maintaining the safety of long term plant operation and for addressing requests by licensees for plant life extension. Recognizing that PSR is a regulatory instrument, the regulatory organization plays a leading role in the implementation of a PSR, although the licensee has the primary responsibility for performing the PSR.
Through the PSR, a licensee can show, and the regulator can verify, that there continues to be a valid licensing basis, that ageing processes are being properly managed, and that the plant safety level is comparable to modern plants.
Lung cancer has long been known to be an occupational hazard for underground miners with high exposures to radon and its radioactive decay products (RDP). In fact, RDP is the second leading cause of lung cancer after smoking. Although this risk of lung cancer is generally associated with uranium mining, in fact many other (non-uranium) underground mines and human dwellings in the temperate zones have potential exposure to elevated levels of radon and its decay products and the attendant risks. In particular, recent North American and European pooled analyses of residential radon studies indicate a lung cancer risk at natural background radon levels as low as 100Bq/m3. The World Health Organization (WHO) has an international initiative to evaluate lung cancer risk worldwide and Health Canada has recently proposed a new guideline for indoor radon exposure in Canada from 800 Bq/m3 to 200 Bq/m3. All this emphasizes the need for improved precision in the estimates of the risks arising from exposure to radon and RDP and improved understanding of factors that modify the estimation of risk.
Studies of underground miner cohorts are one of the main sources of information on the potential health risks from radon. The Ontario Uranium Miner Study is of great interest, due to the large number of miners in this cohort and the fact that Ontario uranium miners were exposed to lower levels of radon than in most miner cohorts. Also, their exposure information is of better quality than other miner studies.
After 20 years since the last update of this cohort, a study update is needed now for a number of reasons. Earlier updates of this cohort did not capture all the effects (e.g., increased lung cancers) on miners that entered the cohort later, due to the long latency for development of solid tumours. Similarly, the modifying effects of the RDP and lung cancer relationship is not well understood. For cancers with relatively good survival (e.g., prostate cancer), mortality studies are inferior to incidence studies because these cancers are comparatively unlikely to appear in death certificates as the cause of death. The previous studies did not examine the association between gamma radiation and cancer or other causes of death. Finally, miners who entered the cohort later had lower exposures and better quality exposure information. The additional 20 years of follow-up will approximately double the person years of experience and permit the study of RDP and lung cancer at low doses with more certainty (greater power) than in the past.
In general terms, the objectives for the proposed update and re-analysis of Ontario Uranium Miners cohort are:
- update Cohort's mortality (1955–present) and add cancer experience (1969–present)
- evaluate lung cancer risk in miners related to RDP
- estimate effect of modifying factors
- control for confounders
This report contains notes and comments by members of the Delegation of Canada to the 54th session of the United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEAR), held May 29 to June 2, 2006 in Vienna.
The report comprises two parts. Part I is the consolidated report by members of the delegation concerning the documents discussed in draft during the 54th session, of which five were approved for publication, pending incorporation of final changes directed by the committee. This was coordinated and assembled by Kevin Bundy (CNSC).
The second part is provided solely by the Canadian representative, who also currently serves as Vice-Chairman of UNSCEAR, being nominated to this position at the 2005 meeting by the other representatives. Part II addresses additional aspects of the 2006 meeting, including: the Committee's 2006 Report to the General Assembly and the ensuing discussion of this report in the Fourth Committee in UNHQ in New York; corrections authorized by the executive of UNSCEAR; celebration of the Committee's fiftieth anniversary, with congratulatory words from Kofi Annan, UN Secretary-General, and Dr. Hans Blix, keynote speaker; and further comments on specific documents.
A memorandum of understanding (MOU) between the National Research Council (NRC) and the CNSC directed members of the Ionizing Radiation Standards (IRS) Group to study the question of neutron production by low-energy electron accelerators. This report summarizes the results of the work carried out under the terms of this MOU. We used the NRC research accelerator to irradiate targets of Be, Al, Cu and W with an electron beam with various energies from 9 to 20MeV. Bubble detectors were used to measure the neutron dose-equivalent rate in the vicinity of the targets. Detectors were individually calibrated in terms of neutron dose-equivalent using an AmBe neutron source. The measured neutron dose rates were compared to those predicted by the Monte Carlo codes MCNP, FLUKA and GEANT. Differences between the Monte Carlo code predictions can be very large and depend on both the target material and electron energy. For a W target the codes agree to better than a factor of two and the GEANT predictions agree best with the measured values. Assuming a W target with an X-ray dose rate of 30Gy/min at 1m, the maximum neutron dose-equivalent rate at the edge of an exclusion zone 15m from the target is estimated to be about 84µSv/hr if there is no neutron shielding. Ordinary concrete with a thickness of about 20cm would reduce the dose-equivalent rate below 25µSv/hr. We include a survey of the relevant literature.
Review comments are presented which are related to the following documents offered for review during a visit to CNSC headquarters in November, 2006:
- NRULE-03600-ASD-001, -001V1 through -001V10
- AECL NRU Reactor Safety Analysis Reports (Volumes 1 and 2)
- draft / internal CNSC document dated Feb. 10, 2006 and Dec. 5, 2006
Since the visit during November, 2006, several items were specifically noted and have been addressed in the review report along with additional supporting comments and assessments:
- Development of technical information and bases for judging the adequacy of the assumptions made in the NRULE reports (V1 to V10) in relation to exothermic energy releases from molten Al-U3Si fuel interactions. The relevant databases of information have been accessed and now highlighted in my report. This includes information related to the extent and rate of energy releases. During face-to-face meetings called by the CNSC in 11/2006 it was revealed by NRU contractor (C. Blahnik) that his assessment neglected this component of energy input under the belief that the rate of reaction is extremely slow (i.e., takes place over an hour or more). Based on the explanation and clarifications provided in the attached draft report, this needs to be revised and inclusion of fairly rapid (seconds to minutes) completion of the said reaction and the impact it will have on resulting core melt progression and energetic explosive reactions need to be taken into account.
- During the 11/2006 review meeting with NRU representatives (R. Leung and C. Blahnik) it was discussed that the onset of potential steam explosions should consider the triggering sources and the relative propensity for initiating an explosive interaction between aluminium and water. Along these lines, to provide guidance the contractor has retrieved the available published documentation of highly-focused experimentation related to triggering of control rod melting related steam explosions under severe accident conditions prevalent in the Savannah River production (SRP) reactors. This information is included in this report and should be of assistance for guiding decisions related to resolving potential safety related threats from the onset of steam explosions in the NRU facility in a more structured manner.
- Hydrogen generation during steam explosions has been ruled out in the NRULE documents that I reviewed. While this may be reasonable to assume for general non-explosive scenarios, it is certainly not the case for cases involving ignition-coupled steam explosions. Generation of hydrogen, deflagration and detonation phenomenology is important for any safety case during severe meltdown accidents. I have compiled relevant experimentally based information developed for USDOE production reactor programs to offer input into this potential threat and the same is included in the draft report. This information indicates that significant hydrogen generation and, of course, the potential impact on safety is non-negligible for NRU safety studies and must be incorporated as advised in the attached report.
- Recriticality issues for the NRU were identified during my visit as worthwhile to revisit and to develop a structure to address. I have already provided considerable technical material to CNSC staff. A simple but straightforward framework for assessing this threat to the NRU facility during severe accidents is now included in this report. This framework appears to indicate that recriticality in NRU can not be dismissed a priori. The presence of multi-kg quantity of U-235 in a D2O environment is known from experimental studies of bare spheres alone to result in critical masses ranging from as low as 3 to 4 kg. Of course, the situation for NRU is more complex (including the potential impact for burnup fission products, B-10 control poison, etc.). However, the NRU system even at 130 MW incorporates close to 50 kg of U-235 and copious quantities of D2O moderator. In the absence of a modified core-melt progression assessment that includes various possible pathways for relocation of the melt, it is highly premature to conclude the lack of the potential for intermittent recriticality (as has been evident from review of the NRULE documents).
- Inquiries have been made per request in relation to the MELCOR code with the code developers at USDOE's SNL site as well as with past staff of the Savannah River production (SRP) reactor severe accident effort. This was motivated by the need to obtain relevant information on the availability and status of MELCOR as a systems-wide code to use for NRU studies (recognizing it's unique U-Al fuel). What I have found is that the MELCOR/SR code system for addressing U-Al reactors during severe accident conditions is presently available in the public domain. However, the code and associated detailed documentation has been archived at the Savannah River site. Investigations related to the MELCOR/SR code do indicate that the version of MELCOR/SR developed for successfully addressing severe accident issues for the SRP reactors (which led to successful restart of the SRP reactors for tritium production) is deposited in a repository at the Savannah River laboratory (SRL). It is presently unclear what level of effort will be required to gain access to the operating manuals and related electronic media since the K-Reactor complex was used for classified purposes. Finally, discussions with MELCOR code developers indicate, fortunately, that recent developments and refinements to MELCOR do appear to make it suitable in itself for modelling the NRU-specific structures for core-melt progression including axial heat transfer along with fission product release and with use of U-Al material properties with some restricted effort. Therefore, regardless of whether the MELCOR/SR code system is made available to AECL or to the CNSC for NRU studies, the USNRC-sponsored code system MELCOR may be able to serve a useful purpose for NRU's core melt progression studies as well as for radionuclide transport characterization through the confinement system as well as to integrally evaluate steaming-type loads.
Finally, per request from Mr. Leung (of AECL) and agreement with the CNSC pointed inquiries were made to confirm and back up my own knowledge of past expenditures of needed resources for the acceptable level of conduct of world-class severe accident safety studies in ~100 MW type research reactors in the USA. Due to differences in cost per man-year in the US vs Canada and elsewhere, I have offered these insights for their consideration (transmitted to me in confidence) in terms of person-time and level of effort. This information is included in the report. A direct comparison indicates that the present level of effort by AECL for NRU severe accident safety characterization is grossly lower in my opinion and evident from the NRULE documents that I have reviewed for the CNSC as part of this effort. I hasten to add that this comment and conclusion are not to be construed as passing a negative judgment on the technical competence of the AECL staff or their contractor which I find to be quite good; merely to indicate that the extremely restrictive effort level sponsored by AECL is very likely the cause of the inadequate level of safety assurance as evident from the limited studies and level of sophistication presented in their NRULE documents.
Based on my experience, the present state of evidence and methodologies as used and presented by AECL can not stand up to worldwide scrutiny for judging the safety of the NRU facility during severe accident conditions, but this should be possible to rectify with a renewed focused effort to correct for the deficiencies pointed out in this report. In this spirit, it is my hope that the review comments and guidance provided in the attached draft report will help guide the next round of improvements to bring the resulting submission to expected world standards. A world-class facility such as the NRU should require nothing less.
RSP-0217 – Investigation into the role of pipe breaks in the licensing of CANDU reactors with positive void reactivity feedback and the credible application of early detection (leak-before-break)
This report investigates the frequency of leaking of the reactor cooling systems (RCS). The approach assumed that the requirements of clause 6.1 of the draft regulatory document could be met if the probability of a leak large enough to void the reactor was so low that for all practical purposes positive void reactivity feedback wouldn't happen.
In the past statistics of failures, based mainly on fossil power plants or extrapolated from data on small leaks, did not indicate that the frequency of leaks was extremely low. Now decades later there is more data and better ways to analyze it. For CANDU the predicted frequency of leaks (10s to 1000s of gpm) per reactor year (ry) is 4.2E-7/ry for an 18ins PHT pipe and 3.6E-7/ry for a 22ins header. Some of these leaks would leak indefinitely; some would be expected to continue to grow to critical length and form a double ended guillotine break unless operator action was taken.
To avoid a power pulse, mitigating action should be taken before the leak rate exceeds the capacity of the make-up pumps and voiding starts; this is long before a crack has grown to a critical crack length. For a CANDU, voiding may start for a break size of 4ins diameter pipe or less. This encompasses feeder sizes, therefore the relationship between the make-up capacity and the diameter of feeders is critical.
The design of reactivity control, core cooling, and containment for a CANDU are based on postulates of double-ended guillotine breaks (DEGBs), usually header breaks or primary circuit breaks. Also, voiding can be caused by operator error and by component failure. Statistics for these were included with the estimates of leak rates and combined with the results of probabilistic fracture mechanics. Then the proportion of leaks that would propagate quickly to DEGBs was estimated, assuming fatigue would be the main failure mode. This gives DEGB failure rates of 7.7E-10/ry for 18ins pipe and 8.8e-10/ry for 22ins headers. Failure rates for smaller pipe are much higher.
If an acceptance criterion for the frequency of a DEGB was taken as extremely low, i.e., less than 1E-6, then all the PHT pipes greater than 7ins diameter would have extremely low failure rates if LBB were credited. If LBB was not credited, only pipes greater than 18ins. would have extremely low failure rates.
In CANDU, the leak detection threshold is very good, but the required response to a leak somewhat indefinite. If a station's response to leaks was reliably quick-detecting leaks before their rate exceeded the make-up capacity then LBB could be credited as another element of defence-in-depth. A potential power rise could be curtailed without waiting for, or relying on, the trip signals for the shutdown systems.
Summarizing: The frequency of leaking (10s to 1000s of gpm) from an 18ins pipe or larger is extremely low. Crediting effective response to leaking, LBB, would decrease the pipe size with extremely low failure rates to 7ins or larger.
Consequently, with legislated requirements for reliable operator response to leaks in less than one hour, leak before break could be credited for all primary heat transport piping greater than 7ins (18mm) diameter. Since a double ended guillotine break of pipes smaller than this can cause voiding, this would be insufficient assurance that positive void reactivity feedback was also extremely low.
What is of concern is the frequency of failure of pipes that could void the reactor is too high to ignore (3.3E-2) corresponding to a 2.5 percent RIH break whereas the criterion for an extremely low failure rate, so low it could be ignored, is a 6 percent RIH break. Therefore it is unlikely that a licensee could argue that voiding could be reasonably avoided. The arguments for licensing a reactor with positive void reactivity would have to rest on the likelihood of a tolerable degree of voiding
RSP-0218 – Guidance on meeting regulatory expectations for the engineering safety aspects of protection from malevolent events
Following the events on September 11, 2001, there have been a series of ongoing initiatives, at the CNSC that have served, and continue to serve, to significantly reduce the risks associated with malevolent threats. These initiatives have mandated several licensee assessments, analyzing not only the physical protection (security) aspects of malevolent event protection, but also the engineering safety aspects of this protection.
As part of the Robustness Project, staff at the CNSC has identified the need to document more formally the regulatory expectations relating to the engineering safety aspects of malevolent event protection. This report contributes to that initiative.
In developing this report we discuss possible malevolent events in the context of future nuclear power plants. The set of possible malevolent events that we include in this report are derived from our work in Chapter 1 on Review of International Practices. These are the set of threats, for which we feel a consensus is building in the international community that will be applied to future nuclear power reactors.
In Chapter 2 we list and then systematically parameterize these events to understand and illustrate the engineering safety consequences that these malevolent events could have on key structures, systems and components of future nuclear power plant. We have focused on the importance of protecting at least one success path (i.e., shutdown, removal of decay heat and containment of radioactive materials). We have highlighted issues such as incident damage progression (i.e., loss of structural integrity with time) that will be important from a rescue, firefighting and operational point of view. These issues could affect layout & defensive construction techniques such as the prevention of progressive collapse. Throughout the report we have included the importance of physical protection infrastructure to assist structural robustness concerns.
In Chapter 3 we present recommendations to facilitate the development of regulatory requirements to ensure that the postulated malevolent events will not cause unacceptable consequences. We recommend four sets of regulatory documents that we feel will allow the licensees to generate the corresponding programs that will achieve the end result of verifiable protection practices from malevolent events.
RSP-0219 – Review of liner and cover design and construction plans; The Port Hope Long-term Low-level Radioactive Waste Management Project preliminary engineering design report Volume 1. Design and operations plan LLRWMO-1340-PDD-12001, revision 0, R. Kerry Rowe Inc.
The conceptual design for the cover and liner proposed for the Port Hope Project Long Term Waste Management Facility, as detailed in the Preliminary Engineering Design Report [Volume 1, LLRWMO-1340-PDD-12001, Revision 0, July 2006] (the report) is considered to be generally sound. Provided the issues raised in this review are addressed and provided that the detailed design, specifications and construction drawings are subjected to a thorough peer review to confirm that the issues raised herein have been addressed, it is considered likely that the “less than 1mm/a seepage rate” out of the landfill assumed in the Environmental Assessment Study Report will be achieved over the 500-year period being considered.
The comments made in this review should be read in context. The report that was reviewed is a preliminary engineering design report and as such contains a level of detail adequate for conceptual review but not for fully assessing the eventual performance. Greater details would be expected to be provided in the detailed design report, including detailed specifications and construction quality control and assurance procedures. While this review identifies issues not addressed in the report that need to be addressed, this is not meant as criticism since these details would not generally be provided in a preliminary engineering design report. Rather they are provided as a check list to assist in the preparation of that report and in the review of that report. It is my opinion that provided these issues are addressed (and I consider it feasible to address all of these issues), this landfill can provide safe long-term containment of the proposed wastes.
There is no reason that a primary leachate collection system with a design life of many hundreds of years could not be designed for this site. However the current design does not, in this reviewer's opinion, provided for adequate assurance of long term-performance for the primary leachate collection system. Specific factors to be considered in the revision of the design of the primary leachate collection system and preparation of the detailed design report, construction drawings and specifications have been itemized.
Given the nature of the waste and provided that the design of the primary leachate collection system is revised to respond to the issues raised in this review, the design of the secondary leachate drainage layer with a hydraulic conductivity to exceed 1x10-4 m/s is considered reasonable. However it is recommended that redundancy is added with respect to the sump in the secondary layer (as well as in the primary layer).
The report provides some indication of factors to be considered with respect to the compacted clay liner. This is reasonable as far as it goes. However in order to prepare the final detailed design and specifications, additional work is needed as indicated in this review (including consideration of clay-leachate compatibility).
The report does not address the potential for cation exchange between the geosynthetic clay liner (GCL) and adjacent soil or the potential effect on GCL hydraulic conductivity. Nor does it adequately address the potential for freeze-thaw of the GCL in the landfill cover. To provide confidence that the assumed low hydraulic conductively of the GCL proposed for the cover will be achieved, it needs to be demonstrated that either: (a) frost will not reach the GCL; or (b) the geochemistry of any pore fluids in soil adjacent to the GCL is such that significant cation exchange is not likely.
The proponent has indicted that “it is anticipated that the potential heat generation from decomposition of the organic component of the co-mingled waste will be significantly less compared to typical MSW landfill sites” and that “there is no measurable heat generated due to radioactive decay of the waste. Based on the above, it is expected that the temperature of the liner system will be very similar to the natural ambient temperature of the surrounding ground.”
Presuming that this information is accurate, the liner temperature can be expected to be less than 20oC which will give an acceptable (greater than 500 years) service life for the geomembranes and minimizes any risks of desiccation of clay liners due heat generated by the waste. However, it is recommended that the liner temperature be monitored in each cell to confirm the validity of the assertion that the liner temperature will be very similar to the natural ambient temperature of the surrounding ground.
Both GCLs and compacted clay liners (CCLs) are susceptible to shrinkage and desiccation cracking, particularly when below a GM in a composite liner. The report does not adequately address measures to minimize the risk of desiccation of the CCL or shrinkage and panel opening of the GCL during construction. Thus the detailed design report should address: (a) the construction procedures that will be adopted to ensure that there will be no desiccation of the CCL before or after covering with the geomembrane and prior to placement of waste; and (b) specification of a GCL that has a low probability of shrinkage.
Leakage through the proposed composite liners at the LTWMF will be highly dependant on the presence and size of holes and the coincidence of the holes with wrinkles. Thus there is a need to (a) detect and repair holes formed during construction, and (b) minimize the potential development of holes during and after waste placement. The comments in the report regarding geomembrane installation and testing of seams are reasonable in the context of a preliminary report but will need elaboration in the detailed design report.
The detailed design report should adequately address:
- the performance a leak detection survey shortly after construction of the liner system and the repair of holes
- limits on gravel in the foundation soil to be placed below the geomembrane/composite liner;
- protection from damage or strains that could ultimately cause stress cracking due to indentation caused by gravel in the granular drainage layer
- wrinkles in the geomembrane
- construction methods and inspection procedures that minimize the risk of holes being formed in the geomembrane during placement of the drainage layers or the select waste
- procedures that minimize the risk of damage to the geomembrane during the operating life of the landfill (e.g., if there is a need to move waste or drums for any reason)
The current design has granular layers above the geomembrane (in one case separated from the geomembrane by a geocomposite and in once case by a geotextile cushion). Provided that these layers are specified to be sand that does not contain angular particles (or isolated gravel) that could perforate the geomembrane, they should provide adequate protection of the geomembrane assuming that appropriate construction specifications are adopted. The geocomposite above the primary geomembrane has the potential of inducing tensile strains in the underlying geomembrane. If this is retained in the design, testing would be required to confirm that the proposed geocomposite will not induce significant tensile strain in the geomembrane under simulated field conditions.
If the design is modified to address the concerns raised in this review regarding the design of the primary leachate collection system to include a coarse gravel drainage layer (as per O. Reg 232/98, Schedule 1), there will be a need to provide adequate protection of the geomembrane. A sand layer that does not contain angular particles (or isolated gravel) has the potential to provide this protection.
Wrinkles in a geomembrane increase the potential for contaminant migration through a hole in the geomembrane at or near the wrinkle and the potential for development of future holes. The report does not address the issues related to wrinkles. While it is difficult, and not essential, to construct without wrinkles, the construction specifications and the detailed design report should address the issue of minimizing wrinkles that will remain once the geomembrane is covered.
The calculations of leakage through composite liners used in the Environmental Effects Assessment Report (Appendix C) appear to be based on techniques that underestimate the observed leakage through composite liners and do not take account of the effects of wrinkles. In this reviewer's opinion, the leakages ultimately observed at the LTWMF are likely to exceed that indicated by the calculations in Appendix C even assuming good design, construction and construction quality control/assurance.
The consideration of both advective and diffusive contaminant transport and the use of the Program POLLUTE in the Environmental Effects Assessment Report is considered appropriate. However the selection of parameters and the modelling itself (as reported in section D8) was not reviewed since it is beyond the scope of the present assignment.
The long-term performance of a geomembrane will depend on the geomembrane properties, the tensile strains in the geomembrane, the exposure to chemicals in the leachate, and temperature. The detailed design report will need to provide detailed specification of geomembrane properties that will maximize the likely service life of the geomembrane. The geomembrane should meet the specifications of both GRI GM-13 and O. Reg 232/98 (Schedule 3). The tensile strains in the geomembrane can be minimized (and hence its service life maximized) by: (a) minimizing the potential for angular particles (e.g., gravel) in the soil below the geomembrane/GLC composites and in the CCL in the geomembrane/CCL composite; (b) having adequate protection above the geomembrane; and (c) minimizing wrinkles in the geomembrane.
The report indicates that “there are no maximum concentrations that would exclude material from being required to be managed at the LTMWF” (§8.1, p. 8-1). Thus it will be important that waste containing contaminants that would reduce the service life of the geomembrane, and especially waste with high concentrations of these contaminants (e.g., hydrocarbons, surfactants, trace metals etc.) be located in the waste pile well away from the liner system. This should be addressed in the detailed operations plan. It is also recommended that the leachate characteristics be monitored with time.
Assuming that: (a) the HDPE geomembranes specifications meet both GRI GM-13 and O. Reg 232/98 (Schedule 3), (b) the issues identified in this review are adequately addressed, (c) the temperature of the geomembranes does not exceed 20oC and (d) the leachate characteristics with respect to contaminants that can impact on the geomembrane service life are not higher than found in MSW leachate, the service life of the geomembranes in the primary and secondary liners are estimated to be of the order of 600 years and hence in excess of the 500 year design period for the facility.
The 2:1 berms inside the landfill are considered steep and may induce undesirable stresses in the liner system (especially the geomembranes). It is recommended that these slopes be reduced to 3:1.
The detailed design report should include details regarding the testing, inspection and monitoring programs to be conducted prior to, during and following construction of the liner, during operations and following closure of the facility. The report addresses a limited number of these issues. This review has highlighted a number of factors that should be considered in the detailed design report. There should also be an independent (third party) firm, experienced in CQA for lined landfill facilities with both geosynthetic and compacted clay liners, retained for construction quality assurance.
This review has identified a number of issues that need to be addressed in order to minimize leakage to the underlying groundwater. Provided that these issues are adequately addressed and appropriate third party construction quality assurance is provided during construction, it is considered likely that the long-term post closure annual average leakage out of the landfill to be less than 1mm/a assumed in the environmental assessment study report over the specified 500 year facility design life.
RSP-0220 – Scaling of RD-14M large LOCA experiments GAI 00G01: A review of COG-04-2023: A scaling assessment of RD-14M for coolant voiding during the power pulse phase of a postulated large-break LOCA scenario: application to a 20% reactor inlet header break in the Point Lepreau reactor
Positive void reactivity feedback of CANDU reactors causes a fission power pulse after a large-break loss of coolant accident (LBLOCA). The power pulse can severely challenge fuel integrity, depending on the rate and extent of coolant voiding.
The CNSC initiated generic action item (GAI) 00G01 in 2001 for all CANDU utilities to focus on the importance of accurate computer models for channel void predictions. It is therefore important for computer validation that all aspects important to the channel voiding behaviour during a Large-Break LOCA in a CANDU reactor are simulated in properly scaled experiments. The available test facility is called RD–14M and electrically heated. The subject report COG-04-2023 is the second scaling analysis submitted by the industry and based on the hierarchical two-tier scaling (H2TS) methodology.
The review shows that the analysis follows in principle the steps of H2TS except for two omissions. The report needs to be clarified and revised to provide scaling criteria for escalation of fission power, for fission power shut-off, for fuel response, for fission power feedback to vapour generation and possibly for pump coast-down. For completeness, the scaling analysis must be revised to include metrics of neglected agents of change, namely γ-heating and feedback from fission power to void generation. Scaling groups need to be physically interpreted and some apparent contradictions need to be resolved by explanation or revision of conclusions. Conclusions rightfully to be drawn from the H2TS analysis were used as starting assumptions without substantiation. That renders the present H2TS analysis to be not auditable. The subject scaling analysis depends strongly on experiments and computer simulations that are still to be qualified by the same scaling analysis.
Without reactor scaling criteria for fission power and fuel response, there is no way to say whether the non-nuclear RD-14M facility can simulate fission power excursion in a nuclear reactor; no scale distortion can be estimated. Void fraction in RD-14M should be sensed and the signal should be used to control the electrical heating in RD-14M heating elements.
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