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Session Chair: Ram Srinivasan, Consultant, United States of America Session Chair: Hyun Gook Kang, RPI, United States of America
Single Failure Criterion Requirement in Canada
Sanja Simic, Robert Rulko
Canadian Nuclear Safety Commission
At the international level single failure criterion (SFC) requirements have existed for many years; however, in the early days of CANDUs, Canada did not have the explicit SFC requirements. Instead, Canada implicitly required that the special safety systems be designed with SFC in mind. The design was required to have sufficient redundancy such that no failure of any single component of a special safety system could result in impairment to an extent that the system would not meet its minimum allowable performance standards under accident conditions. Safety support systems that supply compressed air, electrical power or cooling water were also required to meet this requirement. More recently, the Canadian Nuclear Safety Commission (CNSC) changed its approach to SFC, in an effort to develop technologically neutral, versus CANDU specific, regulatory requirements. The new approach, outlined in REGDOCs 2.4.1 and 2.5.2, follows modern IAEA requirements found in SSR-2/1, which require that the SFC be applied to each safety group incorporated in plant design. Safety groups are defined by the IAEA as: “The assembly of equipment designated to perform all actions required for a particular initiating event to ensure that the limits specified in the design basis for anticipated operational occurrences and design basis accidents are not exceeded.” Therefore, the application of new Single Failure Criteria (SFC) drives redundancies and fail-safe design practices through a larger set of systems than in the past in Canada. The main impact of this is that a single failure must be considered in the safety support systems, not just in the special safety systems themselves. Challenges surrounding the application of single failure criteria are many and include single failures on passive versus active components, testing, maintenance, and operator actions. This paper provides an overview of the evolution of the single failure requirement in Canada.
Time Window Calculations of Operator Action During a MSLBI for a Typical 2-Loops PWR
Alireza Najafi1, A.S. Shirani1, F. Yousefpour2, K. Karimi3, A. Shahsavand4
1Shahid Beheshti University; 2Nuclear Science and Technology Research Institute; 3Islamic Azad University; 4Amirkabir University
In some cases of Success Criteria Analysis framework, it is needed to obtain the available time for human action, to achieve this, Time Window Calculations of operator action is carrying out, and it is intended for determining the realistic needed time of operator initiation action during different sequences of an accident. This study performs Time Window Calculations in the case of a Main Steam Line Break Accident Inside containment for a typical 2-loops PWR. During the accident, When the Emergency Feedwater is not available; the operator action for Primary Feed & Bleed is required to remove the decay heat. The objective of this study is to calculate the necessary time of operator action to initiate the Primary Feed & Bleed, thus preventing the core damage end state. The accident is simulated based on a verified and validated steady-state model. According to the event tree, the operator must take action during the 2nd and 7th sequences of the accident. This study calculates the necessary time of initiating Primary Feed & Bleed for both of these sequences. Results of the 2nd sequence showed that 55 minutes and 135 minutes are almost the maximum available time for the operator with and without the availability of a PSV, respectively. The results of the 7th sequence indicate that 50 minutes and 135 minutes are almost the maximum available time for the operator with and without the availability of a PSV, respectively.
A Computational Risk Assessment Approach for Nuclear Power Plants subjected to Earthquakes and Floods
Jieun Hur1, Halil Sezen1, Cutis Smith2, Tunc Aldemir1, Richard Denning1
1Ohio State University; 2Idaho National Laboratory
Probabilistic Risk Assessment (PRA) plays a major role in identifying vulnerabilities in nuclear power plant (NPP) designs, guiding their daily operations, risk-informing regulations, and prioritizing regulatory oversight. Historically, most of the emphasis of PRA has been on internal events – hazards that typically occur inside the plant such as transients and coolant leaks. Although the importance of external hazards risk analysis is now recognized, the methods for assessing the risk associated with low probability external hazards rely heavily on subjective judgment of specialists, often using conservative elements in the analysis, as is the case with the conservative Deterministic Failure Margin approach commonly used for seismic probabilistic risk assessment.
This study presents a computational risk assessment (CRA) approach to evaluate physical processes using mechanistic models, computer resources, and probabilistic scenarios in order to provide better confidence in the validity of the results. The CRA involves the development and application of methods for the high fidelity deterministic assessment of accident phenomenology, stochastic modeling including the development of probability density functions, sampling from uncertainty distributions, determination of initiating event frequencies, and assessment of system and component failure probabilities. The paper: i) illustrates the CRA, and its challenges and limitations, ii) offers a comprehensive framework for the estimation of risk to nuclear power plants (NPPs) subjected to external events such as earthquakes and floods, iii) enhances the understanding of component performances, and, iv) offers new insights about the physical impact from external hazards to NPPs and associated uncertainty quantifications using realistic case studies.
For the case studies, models were developed for structures, systems, and components (SSCs) at a NPP site including auxiliary building, containment building and condensate storage tanks. Probabilities of failure of SSCs were estimated considering seismic loading and internal flooding caused by seismic loading. Spatial dependency of equipment failures were also investigated.
Risk Assessment of Site Specific Natural Hazards of the UAE
Since the UAE has declared the building of its first civil nuclear power plant, the necessity for conducting natural hazards risk assessment has increased which is solely based on nuclear siting criteria. The potential effects of natural hazards can be mitigated by commencing a proper risk assessment that will inform the selection of the nuclear site. In this respect, the selection of the risk assessment methodology is crucial as it would support the decision makers in the stage of site selection. This study employs the method of risk matrix applied on natural hazards for seismic, flooding and tsunami. In addition to the former hazards, Dust-Sand Storms “DSS” and climate change have been assessed in a form of context discussion. DSS assessment in particular is a vital element in terms of evaluating the natural hazards. This study has been conducted using mainly the risk matrix as the primary assessment tool combining the probability of occurrences and severity. Several risk matrices are developed to accumulate all historical, scientific and other global data that is used for the comparison of risk levels between the UAE and the rest of the world. In this work, the susceptibility of regions to natural hazards has been taken into account under the aspects of historical events recorded with respect to the geographical location. Vulnerable regions are represented through hazard maps as a result of the risk assessment. The overall outputs show encouraging results in some regions of the UAE. Although the assessment presents lower risk regions, it should be emphasized that minimizing the uncertainty by acquiring more historical data is required.