scholarly journals A Method to Avoid Underestimated Risks in Seismic SUPSA and MUPSA for Nuclear Power Plants Caused by Partitioning Events

Energies ◽  
2021 ◽  
Vol 14 (8) ◽  
pp. 2150
Author(s):  
Woo Sik Jung
Keyword(s):  
Exact Solutions ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Fault Tree ◽  
Ground Acceleration ◽  
Special Operations ◽  
Core Damage ◽  
Full Power ◽  
Conditional Events

Seismic probabilistic safety assessment (PSA) models for nuclear power plants (NPPs) have many non-rare events whose failure probabilities are proportional to the seismic ground acceleration. It has been widely accepted that minimal cut sets (MCSs) that are calculated from the seismic PSA fault tree should be converted into exact solutions, such as binary decision diagrams (BDDs), and that the accurate seismic core damage frequency (CDF) should be calculated from the exact solutions. If the seismic CDF is calculated directly from seismic MCSs, it is drastically overestimated. Seismic single-unit PSA (SUPSA) models have random failures of alternating operation systems that are combined with seismic failures of components and structures. Similarly, seismic multi-unit PSA (MUPSA) models have failures of NPPs that undergo alternating operations between full power and low power and shutdown (LPSD). Their failures for alternating operations are modeled using fraction or partitioning events in seismic SUPSA and MUPSA fault trees. Since partitioning events for one system are mutually exclusive, their combinations should be excluded in exact solutions. However, it is difficult to eliminate the combinations of mutually exclusive events without modifying PSA tools for generating MCSs from a fault tree and converting MCSs into exact solutions. If the combinations of mutually exclusive events are not deleted, seismic CDF is underestimated. To avoid CDF underestimation in seismic SUPSAs and MUPSAs, this paper introduces a process of converting partitioning events into conditional events, and conditional events are then inserted explicitly inside a fault tree. With this conversion, accurate CDF can be calculated without modifying PSA tools. That is, this process does not require any other special operations or tools. It is strongly recommended that the method in this paper be employed for avoiding CDF underestimation in seismic SUPSAs and MUPSAs.

Site core damage frequency for multi-unit Nuclear Power Plants site

2017 ◽  
Vol 96 ◽  
pp. 56-61 ◽  
Author(s):  
Mahendra Prasad ◽  
Gopika Vinod ◽  
Avinash J. Gaikwad ◽  
A. Ramarao
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Core Damage

scholarly journals Uncertainty analysis of containment dose rate for core damage assessment in nuclear power plants

2018 ◽  
Vol 50 (5) ◽  
pp. 673-682
Author(s):  
Guohua Wu ◽  
Jiejuan Tong ◽  
Yan Gao ◽  
Liguo Zhang ◽  
Yunfei Zhao
Keyword(s):  
Uncertainty Analysis ◽  
Dose Rate ◽  
Damage Assessment ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Core Damage

scholarly journals AN OVERVIEW OF RISK QUANTIFICATION ISSUES FOR DIGITALIZED NUCLEAR POWER PLANTS USING A STATIC FAULT TREE

2009 ◽  
Vol 41 (6) ◽  
pp. 849-858 ◽  
Author(s):  
Hyun-Gook Kang ◽  
Man-Cheol Kim ◽  
Seung-Jun Lee ◽  
Ho-Jung Lee ◽  
Heung-Seop Eom ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Fault Tree ◽  
Risk Quantification

The Use of Risk-Informed Inspection Approaches of Ukraine

2013 ◽  
Author(s):  
Bogdan Hryshchenko ◽  
Mykhailo Polianskyi ◽  
Anatoliy Nosovskyy ◽  
Oleksandr Sevbo
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Operating Experience ◽  
State Inspection ◽  
Core Damage ◽  
Unit Design ◽  
Low Efficiency ◽  
International Financial Corporation

Currently in Ukraine inspection activity is based on deterministic conservative principles, operation experience and expert appraisal of the inspector. Possibility and benefits of PSA are used with low efficiency. Results of the study conducted by IFC (International Financial Corporation) indicate the absence of a risk-informed approach in the practice of inspections in Ukraine. Also, according to the World Bank appraisal of the investment climate in Ukraine it should be concluded that until regulatory authorities begin implementation of a risk-informed approach in planning inspections in Ukraine random unscheduled events will dominate and won’t provide the goal of State Inspection. Information which obtained from the PSA helps to direct human and financial resources to the problems research that are the most important for safety, and to eliminate or reduce the requirements, which will reduce expenses of solution for significant issues. Inspection in the planning, preparation, implementation and evaluation of the results of which, in addition to deterministic estimates, operating experience and expertise evaluating risk are used is called the risk-informed inspection. The use of risk-informed approaches allows to: focus inspections on design and operational aspects, which have dominant influence on the safety of nuclear power plants (NPP); improve the schedule of inspections (recording of risk assessments in determining the scope, frequency, and type of inspection); and use an additional source of information on the systems and the components of power units, personnel availability. Applying a risk-informed approach to inspection, inspectors can focus primarily on systems that make the largest contribution to core damage frequency, failure of which leads to significant increase of CDF. Risk-informed approach allows to select the most important elements to test systems that will increase the efficiency and quality of inspections. Based on the above it can be concluded that the experience of inspector, his knowledge of the power unit design, the process, the mechanisms of failure of equipment and of accident running, the use of information on the importance of components and systems for the safety of nuclear power plants, obtained from PSA — an effective way to achieve the best results in improving safety.

Improvements in the German Fire PSA Methodology Demonstrated at a German BWR Built to Earlier Standards

2009 ◽  
Author(s):  
M. Ro¨wekamp ◽  
O. Riese ◽  
H. P. Berg
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Critical Parameters ◽  
Guidance Document ◽  
Time Intervals ◽  
Screening Process ◽  
Systematic Calculation ◽  
Fire Propagation ◽  
Core Damage ◽  
Full Power

Fire PSA has to be performed as part of the Probabilistic Safety Analysis (PSA) for full power and low power and shutdown states within the mandatory comprehensive Safety Reviews for German nuclear power plants (NPP) at time intervals of ten years. The German guidance document on PSA methods requires a Fire PSA of several steps starting with alternatively a qualitative and quantitative or an advanced combined screening approach providing estimated values for damage frequencies. The existing combined approach has been enhanced by automating the screening process including a systematic calculation of fire propagation probabilities and standardized simplified fire simulations to roughly predict critical parameters. Further enhancements focus on fire induced cable failures and circuit faults and improvements in the uncertainty and sensitivity analysis. Thus the uncertainties in the results of fire PSA can be reduced as far as feasible and the predicted CDF values are more reliable. The results of a Fire PSA for a German NPP with boiling water reactor designed to earlier standards (BWR-69 type) with an overall fire induced core damage frequency of 1.9 E−06/a are outlined. This is in good agreement with results publicly available from Fire PSA for NPP in other countries. However, the CDF values are higher than those for some other German plants resulting from some pessimistic assumptions made.

scholarly journals Multi-state Risk-Based Maintenance Analysis of Redundant Safety Systems Using the Markov Model and Fault Tree Method

2021 ◽  
Vol 9 ◽  
Author(s):  
F. Mohammadhasani ◽  
A. Pirouzmand
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Fault Tree ◽  
System Level ◽  
Aging Effects ◽  
Maintenance Activities ◽  
Maintenance Policies ◽  
Safety Systems ◽  
Tree Method

The risk-based maintenance strategy has received special attention in the safe operation of nuclear power plants. Simultaneous quantification of the positive and negative effects of maintenance activities and components degradation effect makes it possible to accurately evaluate the risk criterion for safety systems of nuclear power plants. However, it is difficult to integrate the effects of maintenance and components degradation into the standard reliability approaches. A straightforward approach for considering components degradation and different maintenance policies is to make use of Markov maintenance models. In this article, the effectiveness of maintenance activities (including changes in the surveillance test intervals and alteration in the different maintenance policies) on the components unavailability with considering aging effects is quantified using Markov maintenance models and then by coupling these models and the fault tree method, the risk measure is upgraded from the component level to the system level. The proposed models are applied to evaluate the unavailability of two safety systems of VVER-1000/V446 nuclear power plants as case studies. The results show that the Markov method due to its multi-state nature is effective in the conservative evaluation of risk measures so that the unavailability computed by the coupling process is higher than the original unavailability (calculated by system fault tree using PSA data of nuclear power plants) for all maintenance policies. In addition, this study illustrates that the developed Markov maintenance models could be applied to the large-scale whole plant level and provides a proper transition from the classical PSA methods to new techniques. This approach integrates the effects of maintenance strategies and components degradation. Also, it provides a practical and a more accurate tool to determine the technical specification of a real nuclear power plant from the risk point of view.

scholarly journals Use of Reliability Block Diagram and Fault Tree Techniques in Reliability Analysis of Emergency Diesel Generators of Nuclear Power Plants

2019 ◽  
Vol 4 (4) ◽  
pp. 814-823 ◽  
Author(s):  
Vanderley de Vasconcelos ◽  
Wellington Antonio Soares ◽  
Antônio Carlos Lopes da Costa ◽  
Amanda Laureano Raso
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Block Diagram ◽  
Electrical Power ◽  
Fault Tree ◽  
Risk Level ◽  
Reliability Block Diagram ◽  
Advantages And Disadvantages ◽  
Diesel Generators

Nuclear power plants (NPPs) are subjected to events such as equipment failures, human errors and common-cause failures, in an environment of complex maintenance, inspection and testing managements. These events will affect the reliability of safety-related systems, as well as the risk level of the plant. Reliability block diagram (RBD) is often used to analyze the effect of item failures on system availability, taking into account their physical arrangement in the system. Fault tree (FT) is a commonly used technique for analyzing risk and reliability in nuclear, aeronautical and chemical industries. It represents graphically the basic events that will cause an undesired top event. Loss of electrical power is one of the main events that influences safe operation of NPPs, as well as accident prevention and mitigation. In case of unavailability of offsite power, emergency diesel generators (EDGs) supply onsite electrical power. This paper carries out reliability analyses of EDGs of NPPs using both RBD and FT techniques. Each technique has its own advantages and disadvantages, allowing a variety of qualitative and quantitative analyses. Outcomes using these two techniques are compared for a typical NPP EDG system.

Sign in / Sign up

Export Citation Format

Share Document