PSA Application in the Diverse Actuation System Design

2017 ◽  
Author(s):  
Zhan Wenhui ◽  
Zhang Binbin
Keyword(s):  
Nuclear Power ◽  
Safety Assessment ◽  
Power Plants ◽  
The Core ◽  
Backup System ◽  
Core Damage ◽  
Actuation System ◽  
Safety Monitoring System ◽  
Internal Event ◽  
Safety Features

Diverse Actuation System (DAS) is designed as a diverse backup system for Protection and Safety Monitoring System (PMS) to perform the functions of reactor trip and engineered safety features actuation in AP1000 type nuclear power plants. However, not all of the PMS functions should be included in the DAS design. In this paper, the Probabilistic Safety Assessment (PSA) technique was used to identify the DAS functions by comparing the core damage frequency caused by initiating events in at-power internal event PSA. Furthermore, protection parameter signals of DAS to actuate mitigating systems are identified by accident progress analysis.

Study on Shutdown Fire PRA for Nuclear Power Plant

2018 ◽  
Author(s):  
Meiru Liu ◽  
Qingnan Zhao ◽  
Wei Deng ◽  
Jinyan Du ◽  
Lin Sun
Keyword(s):  
Risk Assessment ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
East Coast ◽  
Fire Risk ◽  
The Core ◽  
Core Damage ◽  
Risk Result

Fire Probabilistic Risk Assessment (PRA) is one of the main methods of fire safety analysis for nuclear power plants (NPPs). At present, the fire PRA under the at-power condition has been widely studied, while the research on the low power and shutdown condition (LPSD) is quite limited. Therefore, in this paper, a second generation NPP on the east coast of China is taken as the research target, and the analysis methods are based on the latest LPSD fire PRA theory in report NUREG/CR-7114. This paper studies the initiating events and ignition frequencies of fire PRA considering the real conditions in LPSD, and established LPSD Fire PRA model, finally obtained the quantitative risk result of the core damage caused by the fire According to the results of this LPSD fire PRA, the fire risk-significant sources and fire risk weakness are found out and the improvement suggestions have been promoted.

Design Characteristics of Ulchin Nuclear Power Units 5&6

2002 ◽  
Author(s):  
Deucksoo Lee ◽  
Dong-Su Kim ◽  
Young-Taik Lee ◽  
O-Keol Kwon ◽  
Jung-Cha Kim
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
System Design ◽  
Nuclear Power ◽  
Safety Analysis ◽  
Reactor Coolant ◽  
The Core ◽  
Core Damage ◽  
Actuation System ◽  
Design Characteristics

Ulchin nuclear power plant units 5&6 (UCN 5&6), which started excavation on January 1999, are two loop pressurized water reactors (PWR) with the capacity of 1000 MWe, and planned to start commercial operation on June, 2004 and June, 2005, respectively. The reactor coolant system of the UCN 5&6 consist of a reactor vessel, internals, and two steam generators, four reactor coolant pumps, a pressurizer and primary piping. Based on the system design of the first Korean Standard Nuclear Power Plant (KSNP), UCN 5&6 is designed to provide improvements in safety, reliability and cost by applying both advanced proven technology and experiences gained from the construction and operation of the previous KSNP plants. The result of the preliminary probabilistic safety assessment study for UCN 5&6 shows that the core damage frequency is lowered significantly. Several design improvement items have been adopted to the system design and contributed to lower the core damage frequency value. Among the design improvements, digital PPS and digital ESFAS are the key to the UCN 5&6 design. Furthermore, digitization of the Plant Protection System (PPS) and Engineered Safety Feature Actuation System (ESFAS) for the PWR is the first case in the PWR construction history. The Korean regulatory body reviewed the design concept of the digital PPS and digital ESFAS, and evaluated to be acceptable for the plant safety. Also, in-depth review on the detail design of the digital PPS/ESFAS and the special evaluation/audit for the software design process are underway to secure the software quality. The safety of the UCN 5&6 design has been evaluated through a two-year review on the preliminary safety analysis report. As a result, the construction permit was issued on May 17, 1999 by the government. In this paper, design characteristics of UCN 5&6 are discussed focussed on design improvements comparing with KSNR. And, some of the safety analysis results are presented as well as licensing status.

Hazard Analysis of Digital Engineered Safety Features Actuation System in Advanced Nuclear Power Plants Using a Redundancy-Guided Approach

2020 ◽  
Author(s):  
Han Bao ◽  
Tate Shorthill ◽  
Hongbin Zhang
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Hazard Analysis ◽  
Potential Threat ◽  
Safety Control ◽  
Application Systems ◽  
Actuation System ◽  
Safety Features ◽  
Safety Systems

Abstract Replacing the existing aging analog instrumentation and control (I&C) systems with modern safety control and protection digital technology offers one of the foremost means of performance improvements and cost reductions for the existing nuclear power plants (NPPs). However, the qualification of digital I&C systems remains a challenge, especially considering the issue of software common-cause failures (CCFs), which are difficult to address. With the application and upgrades of advanced digital I&C systems, software CCFs have become a potential threat to plant safety because most redundant designs use similar digital platforms or software in the operating and application systems. With complex designs of multilayer redundancy to meet the single-failure criterion, digital I&C safety systems (e.g., engineered safety-features actuation system [ESFAS]) are of a particular concern in the U.S. Nuclear Regulatory Commission (NRC) licensing procedures. This paper applies a modularized approach to conduct redundancy-guided systems-theoretic hazard analysis for an advanced digital ESFAS with multilevel redundancy designs. Systematic methods and risk-informed tools are incorporated to address both hardware and software CCFs, which provide guidance to eliminate the triggers of potential single points of failure in the design of digital safety systems in advanced plant designs.

Research on Start-Up Design of Nuclear Safety Level Parallel Redundant Control Station

2021 ◽  
Author(s):  
Guilian Shi ◽  
Yunxu Shou ◽  
Gang Li
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Safety ◽  
Operating Life ◽  
Safety Level ◽  
Actuation System ◽  
Start Up ◽  
Redundant Control ◽  
Safety Features ◽  
Start Ups

Abstract Compared with the operating life cycle of the digital I&C system in nuclear power plants, the start-up process of the control station is minimal and easily overlooked. A design that is too simple is not suitable for nuclear power applications. The complexity of the start-up design comes from three aspects: One is the diversity of start-up scenarios. In addition to the start of the normal plan, there are unexpected start-ups that cannot be ignored; the second is the complexity of data synchronization in the redundant system; the third is the consideration of human factors. The start-up process involves a lot of human-computer interaction, and how to reduce human risk is also an important design requirement. If the factors are not considered properly, the control station will easily cause disturbance of the controlled equipment when starting, and may even cause the malfunction of the engineered safety features actuation system. This article focuses on the nuclear safety-level parallel redundant control station, analyzes various scenarios of the control station start-up, and synthesizes the design requirements for the start-up phase. According to the requirements, the overall design plan of “initialization-synchronization-comparison-commissioning” is proposed, and the human operation risks involved in each stage are analyzed, and corresponding prevention plans are proposed. The FirmSys parallel redundant control station implemented according to this scheme has been successfully applied in ten commercial nuclear power units including Unit 5 and Unit 6 of Yangjiang Nuclear Power Plant.

scholarly journals Mathematical Formulation and Analytic Solutions for Uncertainty Analysis in Probabilistic Safety Assessment of Nuclear Power Plants

Energies ◽  
2021 ◽  
Vol 14 (4) ◽  
pp. 929
Author(s):  
Gyun Seob Song ◽  
Man Cheol Kim
Keyword(s):  
Monte Carlo ◽  
Uncertainty Analysis ◽  
Probability Density ◽  
Nuclear Power ◽  
Safety Assessment ◽  
Power Plants ◽  
Mathematical Formulation ◽  
Analytic Solutions ◽  
Probabilistic Safety Assessment ◽  
Probabilistic Safety

Monte Carlo simulations are widely used for uncertainty analysis in the probabilistic safety assessment of nuclear power plants. Despite many advantages, such as its general applicability, a Monte Carlo simulation has inherent limitations as a simulation-based approach. This study provides a mathematical formulation and analytic solutions for the uncertainty analysis in a probabilistic safety assessment (PSA). Starting from the definitions of variables, mathematical equations are derived for synthesizing probability density functions for logical AND, logical OR, and logical OR with rare event approximation of two independent events. The equations can be applied consecutively when there exist more than two events. For fail-to-run failures, the probability density function for the unavailability has the same probability distribution as the probability density function (PDF) for the failure rate under specified conditions. The effectiveness of the analytic solutions is demonstrated by applying them to an example system. The resultant probability density functions are in good agreement with the Monte Carlo simulation results, which are in fact approximations for those from the analytic solutions, with errors less than 12.6%. Important theoretical aspects are examined with the analytic solutions such as the validity of the use of a right-unbounded distribution to describe the uncertainty in the unavailability/probability. The analytic solutions for uncertainty analysis can serve as a basis for all other methods, providing deeper insights into uncertainty analyses in probabilistic safety assessment.

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.

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