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[en] This presentation discusses the identification and training of non-technical skills for enhancing resilience in emergency operations centre commanders. Currently there is focus on technical skills and seniority and no focus on non-technical skills. The research project the the Chalk River Nuclear Labs involves the development of taxonomy of non-technical skills, assessing the most important non-technical skills and training of most important non-technical skills.
[en] Highlights: • Adopting thermal-hydraulic phenomena (list of 116 phenomena – published). • Exploiting capabilities of existing numerical tools & existing experimental database. • Procedure to plan new-prioritized research in nuclear thermal-hydraulics. - Abstract: The difficulty in predicting locally and globally the transient evolution of two-phase or multiphase flows in complex systems is well recognized in nuclear thermal-hydraulics. Large efforts involving the expenditure of huge resources during the last three decades in previous century brought to the creation of giant databases (e.g. including experimental data and results of computer code calculations) and to the perception that the safety of nuclear reactors is guaranteed notwithstanding residual areas of unawareness. Nowadays, thousands of scientists continued to generate progress in the area having available much lower resources: more and more dead-ends for established research outcomes are experienced; the progress in knowledge resembles the slow expansion of a swamp rather than the fast moving of a river. In this paper a procedure is proposed to identify directions for research in nuclear thermal-hydraulics which are consistent with the needs in nuclear reactor safety. Two pillars for the procedure are constituted by the characterization of phenomena and by the application of qualified computational tools. Decision makers and scientists may prioritize research in areas where large impacts upon design and safety issues are identified in advance.
[en] Recent regulatory activities have demonstrated a shift in licensee expectations that healthy nuclear safety culture must include both excellence in nuclear security and human performance.While these aspects are integrated to varying degrees in the management system framework of Plan-Do-Check-Act; there are challenges in understanding the use of the language and the new performance indicators that may be useful to make the organization health visible to leaders in these areas. In this regard, such a framework would need to provide oversight assurance that both field and knowledge workers are cognisant of the expectations of nuclear safety and security in their work activities. This paper will explore our recent understanding and operating experience as an OEM vendor in relation to the above and proposed recent changes in REGDOC 2.1.2 Safety Culture . For context, we will provide an overview of our experience in using the Canadian Federal Industrial Security Manual as a basis of management system requirements to context nuclear security in a manner that supports our operations and proposed regulatory direction. (author)
[en] Highlights: • Thermal responses of a scaled down suppression pool are studied numerically. • Steam injection is simulated using direct contact condensation method and effective source method. • Comparing with experiment, direct contact condensation method produces better results. • Reasons for different thermal status are analyzed by flow field patterns. - Abstract: Thermal status or temperature distribution in terms of mixing or stratification inside a suppression pool (SP) is decisive for containment pressure and hence reactor safety. Numerical models were developed in this study to predict the temperature distribution inside a scaled-down SP with steam injected through a horizontal pipe. Regarding the modeling of injected steam, two methods were compared. Effective source (ES) method, which considers the effective heat and momentum source instead of calculating condensing steam directly, can predict the stratified thermal status but failed to reproduce the mixing thermal status. The reason may be attributed to the lack of an appropriate momentum source model. Direct contact condensation (DCC) method, which calculates the condensation process directly, can give agreeable results with experiments and successfully distinguish between mixing or stratification thermal status. In addition, the DCC method can also predict the transition between mixing and stratification. However, moment of the transition was found to deviate from measurement. It was assumed the measurement and proper modeling of the gas-liquid interface structures is necessary for improvement of the simulation. Besides, reasons for the above thermal statuses were analyzed by fluid velocity fields, and it was shown that stratification can be avoided if the injection location is moved downside. The numerical methods proposed can be used for design and improvement of thermal features of SP.
[en] Highlights: • This study presents a LBB leakage analysis code (LEABLE) for various conditions. • Criteria were employed in LEABLE to select the corresponding model. • Single-phase and two-phase critical flow models were proposed. • The LEABLE code was validated by experimental data. • The influences of different parameters on LBB leakage were studied by LEABLE. - Abstract: The leak before break (LBB) concept plays an important role in reactor safety. A code (LEABLE) for leak before break (LBB) leakage estimation was developed in this paper. Complicated crack morphologies and different conditions were taken into account, covering subcritical and supercritical fluids, single phase and multi-phases, choked and unchoked flow. Precise and reliable criteria were proposed for proper model selection. A modified two-phase critical flow model was presented and a single-phase critical flow model for superheat gas was established. Besides, modifications on flashing inception, flow resistance and particulate plugging were conducted in the LEABLE code. Moreover, the code was validated and verified by experimental data for both artificial and natural cracks. Reasonable agreements between experiment values and code predictions are shown for all the conditions. And in the present study the LEABLE shows a higher precision than existing models and commercial softwares (PICEP and SQUIRT). The parameter sensitivity study of LBB leakage was performed as well. The mass flux increases with crack opening crack opening displacement (COD), crack length and the entrance resistance coefficient, and decreases with global and local roughness, number and angle of turns. However, the entrance resistance coefficient and local roughness have limited effects on LBB leakage.
[en] Following the 2011 Fukushima earthquake and tsunami, available margins beyond the design basis earthquake (DBE) were questioned and required to be assessed by the nuclear industry. Provisions were required to be made to ensure that there are adequate margins in the nuclear power plant systems, structures and components (SSCs) to safely withstand beyond design basis earthquakes. This paper provides an overview of approaches taken to evaluate the available margins, the measures taken to enhance plant safety margins and the status of current recommended margins beyond DBE by the Canadian and international nuclear industry to cater for the beyond design basis earthquake also called Design Extension Conditions (DEC). (author)
[en] Highlights: • Conjugating: • ALARA. • BEPU. • Extended safety margin. • Independent assessment. - Abstract: The paper aims at fixing bases for possible strengthening of current Nuclear Reactor Safety (NRS) and safety analysis: this is done by combining the logical frameworks connected with the terms As-Low-As-Reasonably-Achievable (ALARA), Best-Estimate-Plus-Uncertainty (BEPU), Extended-Safety-Margin (E-SM) and Independent-Assessment (IA). ALARA is an early principle in Nuclear Reactor Safety: designers and operators must do their best to minimize doses to the humans. BEPU is an approach in Accident Analysis, part of NRS: one may state that BEPU implies the best use of computational tools to determine the safety of nuclear installations. Then, ALARA may be seen at the origin of BEPU. Furthermore, BEPU (and BEPU elements like V & V, Scaling, procedures of code application and code coupling, etc.) can be extended to all analytical parts of the Final Safety Analysis Report (FSAR). This brings to BEPU-FSAR. Safety Margins constitute an established concept in NRS: a few dozen SM values must be calculated in current safety analyses and demonstrated to be acceptable. The concept can be extended to everything part of the design, the operation and the environment for a Nuclear Power Plant (NPP) Unit, thus forming the E-SM. Here ‘the environment’ includes the personnel in charge of activities connected with the NPP. The E-SM implies the formulation of some ten-thousands safety margins values, which shall correspond to a similar number of monitored variables. IA is an early requirement in NRS: data ownership and system complexity prevented so far a comprehensive application of the requirement. IA analyses conflict with industry policies to keep proprietary data. IA based BEPU-FSAR analyses are essential to finalize the E-SM design. The implementation of the idea in the paper brings to an additional safety barrier for existing and future nuclear reactors which may reduce the probability of core melt, presumably at an affordable cost for the industry.
[en] Online leak repair by sealant injection employs a compound for forming an online moulded gasket injected at suitable pressure in an enclosure formed by valve/flange/pipeline and/or suitable clamps. With techniques prevalent in the industry, it can be done on steam, hydrocarbons or gas leaks from subzero temperatures to 6000°C and pressures up to 350 bar(g). The priority for every plant is to stay online with safety. Online sealing technique is a part of maintenance strategy by which any plant (organization) can save energy and money, reduce noise and emission levels, control erosion damage as well as improve plant safety, and most importantly, protect the environment