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.

A Nuclear Safety Basis for Assigning Priorities to Human Engineering Concerns at Nuclear Power Plants

1983 ◽  
Vol 27 (2) ◽  
pp. 180-183
Author(s):  
Edward J. Sheehy
Keyword(s):  
Human Factors ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Safety ◽  
Human Engineering ◽  
Control Room ◽  
Current State ◽  
Operator Interface ◽  
Safety Features ◽  
Actual Control

At this point in the integration of human factors into the nuclear power industry an important issue is how are the various discrepancies between the post Three Mile Island operator interface guidelines and the actual control room configuration to be assigned priorities for resolution? While it is generally agreed that safety-related discrepancies should receive first priority, it is not at all clear just what “safety-related” means when considered in an operational sense. This paper describes the current state of evolving standard definitions for hardware oriented safety features. It discusses how these definitions could be expanded to include operationally related considerations and explores some benefits which would ensue. These include (1) establishing a nuclear safety basis which can be used in establishing priorities for resolution of human engineering concerns, and (2) obtaining legal recognition of operational concerns in general in the design and licensing arenas.

Preliminary Study on the Identification of DECs Without Significant Fuel Degradation in New NPP Designs

2017 ◽  
Author(s):  
Zhiyi Yang ◽  
Yimin Chong ◽  
Chun Li ◽  
Jiajia Zhang
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Safety ◽  
Severe Accident ◽  
Concept Design ◽  
Energy Agency ◽  
Safety Level ◽  
Safety Requirements ◽  
Specific Safety

New nuclear safety objectives and principles are being studied in main nuclear power countries and organizations after Fukushima Dai-ichi nuclear accident, to further improve the safety level of nuclear power plants (NPPs). Based on International Atomic Energy Agency (IAEA) Specific Safety Requirements (No.SSR-2/1), “Safety of Nuclear Power Plants: Design” (HAF102-2016) is issued in China. The concept “design extension condition (DEC)” is put forward, which is intend to enhance the plant’s capability to withstand accidents that are more severe than Design Basis Accidents (DBA). DEC could include conditions without significant fuel degradation (DEC-A in this paper) and conditions with core melting (DEC-B in this paper), e.g. severe accident. In this paper, the DEC-A and its application was discussed preliminarily, firstly, the development and connotation was introduced, then the identification of DEC-A, and the safety analysis principles of DEC-A were mainly described. This study may play a valuable role for implementation of new nuclear safety requirements in China.

Results of Regulatory Investigations as a Basis for Safety Improvements of Nuclear Power Plants

1996 ◽  
Vol 210 (5) ◽  
pp. 345-350
Author(s):  
H P Berg ◽  
U Volland ◽  
L Weil
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Safety ◽  
Technical Development ◽  
Regulatory State ◽  
Safety Level ◽  
Public Issue ◽  
State Authority ◽  
The Subject

Nuclear safety is not only a matter of technical development but is also an important public issue. Therefore it is mandatory that the responsible regulatory state authority performs its own investigation programme in order to enhance the safety level. The purpose and main objectives of this German regulatory investigation programme as well as the subject of some investigations are presented.

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.

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.

Design and Analysis of a Reliable Communication System in Nuclear Safety Instrument & Control System

2021 ◽  
Author(s):  
Le Li ◽  
Zhihui Zhang ◽  
Chao Gao ◽  
Fei Zhou ◽  
Guangqiang Ma
Keyword(s):  
Communication Networks ◽  
Communication System ◽  
Communication Systems ◽  
Nuclear Power ◽  
Data Exchange ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Safety ◽  
Safety Communication ◽  
Digital Control Systems

Abstract With the development of digital instrument and control technology for nuclear power plants in recent decades, communication networks have become an important part of safety digital control systems, which takes charge in data exchange between the various sub-systems, and extremely impact on the reliability and safety of the entire I&C system. Traditional communication systems where some special features, such as reliability, safety, real-time, certainty, and independence are not strictly required are various illustrated. However, how to implement a communication system in a safety I&C system is rarely stated in current research. In this research, a reliable safety communication system applied in nuclear power plants is designed and analyzed. The five key characteristics of nuclear safety communication networks are explained, followed by explanation of how to achieve these characteristics. The analysis and verification of the designed system are also stated in this paper, which contributes to proving that the designed nuclear safety communication system could applied in the nuclear power plants.

scholarly journals Study on Probabilistic Safety Goals for Multimodule High-Temperature Gas-Cooled Reactor Based on Chinese Societal Risks

2021 ◽  
Vol 2021 ◽  
pp. 1-10
Author(s):  
Jinghan Zhang ◽  
Jun Zhao ◽  
Jiejuan Tong
Keyword(s):  
High Temperature ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Regulatory Commission ◽  
Nuclear Safety ◽  
Energy Agency ◽  
Water Reactor ◽  
Regulatory Commission ◽  
High Temperature Gas ◽  
Probabilistic Safety

Nuclear safety goal is the basic standard for limiting the operational risks of nuclear power plants. The statistics of societal risks are the basis for nuclear safety goals. Core damage frequency (CDF) and large early release frequency (LERF) are typical probabilistic safety goals that are used in the regulation of water-cooled reactors currently. In fact, Chinese current probabilistic safety goals refer to the Nuclear Regulatory Commission (NRC) and the International Atomic Energy Agency (IAEA), and they are not based on Chinese societal risks. And the CDF and LERF proposed for water reactor are not suitable for high-temperature gas-cooled reactors (HTGR), because the design of HTGR is very different from that of water reactor. And current nuclear safety goals are established for single reactor rather than unit or site. Therefore, in this paper, the development of the safety goal of NRC was investigated firstly; then, the societal risks in China were investigated in order to establish the correlation between the probabilistic safety goal of multimodule HTGR and Chinese societal risks. In the end, some other matters about multireactor site were discussed in detail.

Explore the Self-Governance on Nuclear Safety Emergency - A Case Study of the 3rd Nuclear Power Plant

2012 ◽  
Vol 260-261 ◽  
pp. 103-106
Author(s):  
Yi Chun Lin ◽  
Yung Nane Yang
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Alternative Energy ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Safety ◽  
Security Issue ◽  
Safety Issues ◽  
Plant Safety

The ripples of the tsunami crisis in Japan triggered introspections of nuclear plant safety issues in the worldwide. Many countries have claimed the suspension of nuclear power plants. However, some countries such as Taiwan, under nearly 99% energy is exported, the disasters force government and citizen to face the importance of nuclear safety, especially the neighborhoods nearby the nuclear power plants. We have to face the nuclear safety since there is no other alternative energy presently. The 3rd nuclear power plant located in the south of Taiwan, which has the same geographic features with Fukushima, Japan. Presently, there is no precedent in Taiwan of precaution and rescue team and civil supervised mechanic on nuclear security issue. This paper will review according to transparent information, public participation and cross-organization cooperation to propose the execution and work division principles, including information monitor, educational propagation, hide and evacuation, emergence aid and care, rear and refuge service. The ultimate target is to establish self-governance inside nearby neighborhood to confront nuclear disaster at the critical moment.

Studies on Safety Assessment Method for Operating License Extension of Nuclear Power Plants in China

2017 ◽  
Author(s):  
Zhilin Chen ◽  
Ping Huang ◽  
Chunhui Wang ◽  
Zhiyuan Chi ◽  
Fangjie Shi ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Safety Assessment ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Assessment Method ◽  
Nuclear Safety ◽  
Management Systems ◽  
Effect Evaluation ◽  
Environment Effect ◽  
Assessment Environment

It’s the trend to extend the operating license time, called Operating License Extension (OLE) in China, of nuclear power plants (NPPs) in the future. It needs to be adequately demonstrated by licensees and approved by the regulator to gain an extended license time, such as 20 years. The demonstration methods for OLE are different among countries due to the different management systems for NPPs. Safety assessment, environment effect evaluation and update of the final safety analysis report (FSAR) will be the main aspects during OLE demonstration of NPPs in China according to the technical policy issued by National Nuclear Safety Administration (NNSA). Technical methods for scoping and screening, aging management review and time-limited aging analyses, which are the main contents of safety assessment are established based on the technical policy drafted by NNSA and international experiences in order to assist the operators to implement the safety assessment for OLE of NPP.

Upgrading the Operational Safety of Nuclear Power Plants Through the TACIS Nuclear Safety Assistance Programme

2004 ◽  
Author(s):  
M. Bie`th ◽  
R. Ahlstrand ◽  
C. Rieg ◽  
P. Trampus
Keyword(s):  
European Commission ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Safety ◽  
Reactor Pressure Vessel ◽  
Nuclear Safeguards ◽  
Integrity Assessment ◽  
Reactor Pressure

The European Union’ TACIS programme was established for the New Independent States since 1991. One priority for TACIS funding is nuclear safety. The European Commission has made available a total of € 944 million for nuclear safety programmes covering the period 1991–2003. The TACIS nuclear safety programme is devoted to the improvement of the safety of Soviet designed nuclear installations in providing technology and safety culture transfer. The Joint Research Center (JRC) of the European Commission is carrying out works in the following areas: • On-Site Assistance for TACIS Nuclear Power Plants; • Design Safety and Dissemination of TACIS results; • Reactor Pressure Vessel Embrittlement for VVER in Russia and Ukraine; • Regulatory Assistance; • Industrial Waste Management and Nuclear Safeguards. This paper gives an overview of the Scientific and Technical support that JRC is providing for the programming and the implementation of the TACIS nuclear safety programmes. In particular, two new projects are being implemented to get an extensive understanding of the VVER reactor pressure vessel embritttlement and integrity assessment.

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