scholarly journals Integrated Security System (ISS) Design and Evaluation for Commercial Nuclear Power Plant

2021 ◽  
Vol 2021 ◽  
pp. 1-15
Author(s):  
Arafat. H. Hamadah ◽  
Mohamed S. Nagy ◽  
Hanaa Abou-Gabal ◽  
Sayed. A. El Mongy
Keyword(s):  
Nuclear Power ◽  
Threshold Value ◽  
Security System ◽  
Nuclear Industry ◽  
Nuclear Safeguards ◽  
Nuclear Materials ◽  
Physical Security ◽  
Integrated Network ◽  
Testing Procedures ◽  
System Effectiveness

Physical security system, which is also called physical protection system, is very crucial in the nuclear industry for protecting staff, visitors, buildings, assets, and nuclear materials against theft, sabotage, and harmful activities. Theft of nuclear materials has a major impact on the essence of nuclear safeguards. Sabotage of a nuclear facility could endanger the public at large. Reviewing the published literature, it is found that there are no complete physical security system designs based on an integrated network of electronic devices that are devoted to commercial NPPs. And there is no definite evaluation factor that was set to approve such a system. This paper is an evolving solution to this deficiency by proposing an unpreceded integrated security system design applicable to a commonly structured physical layout of any commercial NPP. This proposal provides comprehensive security coverage for the NPP boundaries employing a high level of integration for all subsystems communicated via an IP data network controlled by central management software. This paper is proposing also testing procedures to be followed to evaluate the proposed design. The security system effectiveness will be calculated using mathematical codes by assuming external intrusion attack scenarios. Attributes of each attack scenario will be numerically introduced to the evaluation software EASI and ASSESS codes developed by Sandia Labs, USA. This paper also proposes a threshold value of such security system effectiveness which should be achieved by the commercial NPP security system to achieve the so-called security license.

Nuclear icebreaker fleet and power supply on the basis of floating power units

2018 ◽  
Vol 35 (4) ◽  
pp. 110-113
Author(s):  
V. A. Tupchienko ◽  
H. G. Imanova
Keyword(s):  
Foreign Policy ◽  
Power Supply ◽  
Nuclear Power ◽  
Arctic Region ◽  
Nuclear Industry ◽  
The Arctic ◽  
Far North ◽  
The Arctic Region

The article deals with the problem of the development of the domestic nuclear icebreaker fleet in the context of the implementation of nuclear logistics in the Arctic. The paper analyzes the key achievements of the Russian nuclear industry, highlights the key areas of development of the nuclear sector in the Far North, and identifies aspects of the development of mechanisms to ensure access to energy on the basis of floating nuclear power units. It is found that Russia is currently a leader in the implementation of the nuclear aspect of foreign policy and in providing energy to the Arctic region.

scholarly journals An Approach to Analyze Diagnosis Errors in Advanced Main Control Room Operations Using the Cause-Based Decision Tree Method

Energies ◽  
2021 ◽  
Vol 14 (13) ◽  
pp. 3832
Author(s):  
Awwal Mohammed Arigi ◽  
Gayoung Park ◽  
Jonghyun Kim
Keyword(s):  
Decision Tree ◽  
Nuclear Power ◽  
Power Plants ◽  
Human Error ◽  
Cognitive Activity ◽  
Emergency Operation ◽  
Nuclear Industry ◽  
Working Environment ◽  
Control Rooms ◽  
Diagnosis Errors

Advancements in the nuclear industry have led to the development of fully digitized main control rooms (MCRs)—often termed advanced MCRs—for newly built nuclear power plants (NPPs). Diagnosis is a major part of the cognitive activity in NPP MCRs. Advanced MCRs are expected to improve the working environment and reduce human error, especially during the diagnosis of unexpected scenarios. However, with the introduction of new types of tasks and errors by digital MCRs, a new method to analyze the diagnosis errors in these new types of MCRs is required. Task analysis for operator diagnosis in an advanced MCR based on emergency operation was performed to determine the error modes. The cause-based decision tree (CBDT) method—originally developed for analog control rooms—was then revised to a modified CBDT (MCBDT) based on the error mode categorizations. This work examines the possible adoption of the MCBDT method for the evaluation of diagnosis errors in advanced MCRs. We have also provided examples of the application of the proposed method to some common human failure events in emergency operations. The results show that with some modifications of the CBDT method, the human reliability in advanced MCRs can be reasonably estimated.

Inservice Testing Program Improvements for New Reactors Small Modular Reactors (SMRs)

2014 ◽  
Author(s):  
Ronald C. Lippy
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Nuclear Industry ◽  
Small Modular Reactors ◽  
American Society ◽  
Construction Permit ◽  
Regulatory Commission

The nuclear industry is preparing for the licensing and construction of new nuclear power plants in the United States. Several new designs have been developed and approved, including the “traditional” reactor designs, the passive safe shutdown designs and the small modular reactors (SMRs). The American Society of Mechanical Engineers (ASME) provides specific Codes used to perform preservice inspection/testing and inservice inspection/testing for many of the components used in the new reactor designs. The U.S. Nuclear Regulatory Commission (NRC) reviews information provided by applicants related to inservice testing (IST) programs for Design Certifications and Combined Licenses (COLs) under Part 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants,” in Title 10 of the Code of Federal Regulations (10 CFR Part 52) (Reference 1). The 2012 Edition of the ASME OM Code defines a post-2000 plant as a nuclear power plant that was issued (or will be issued) its construction permit, or combined license for construction and operation, by the applicable regulatory authority on or following January 1, 2000. The New Reactors OM Code (NROMC) Task Group (TG) of the ASME Code for Operation and Maintenance of Nuclear Power Plants (NROMC TG) is assigned the task of ensuring that the preservice testing (PST) and IST provisions in the ASME OM Code to address pumps, valves, and dynamic restraints (snubbers) in post-2000 nuclear power plants are adequate to provide reasonable assurance that the components will operate as needed when called upon. Currently, the NROMC TG is preparing proposed guidance for the treatment of active pumps, valves, and dynamic restraints with high safety significance in non-safety systems in passive post-2000 reactors including SMRs.

The Human Reliability Analysis in Fire PSA Using Spar-H Method

2013 ◽  
Author(s):  
Yao Wang
Keyword(s):  
Reliability Analysis ◽  
Nuclear Power ◽  
Significant Contribution ◽  
Fire Risk ◽  
Nuclear Industry ◽  
Total Risk ◽  
Human Reliability ◽  
Human Reliability Analysis ◽  
Study Results ◽  
In Fire

According to existing research results, fire risk makes a significant contribution to the total risk of a nuclear power plant (NPP). So fire probabilistic safety analysis (PSA) for NPPs is becoming more and more important in recent years. How to perform human reliability analysis (HRA) which is an essential part of PSA is therefore being paid more and more attention in fire PSA. This paper describes the characteristics and special considerations of HRA in fire PSA, and demonstrates in fire PSA how to use SPAR-H method which is so-called an advanced second-generation HRA method and is being widely used in PSA for Chinese NPPs. The study results can be a reference for other HRA analysts to use SPAR-H method in fire PSA models or other PSA models in Chinese NPPs or the world-wide nuclear industry.

Architecting “Maintenance Program” for Piping and Pipe Support

2013 ◽  
Author(s):  
Koichi Tsumori ◽  
Yoshizumi Fukuhara ◽  
Hiroyuki Terunuma ◽  
Koji Yamamoto ◽  
Satoshi Momiyama
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Industry ◽  
Maintenance Management ◽  
3D Cad ◽  
Cad System ◽  
Nuclear Disaster ◽  
Safety Regulations ◽  
Maintenance Program

A new inspection standard that enhanced quality of operating /maintenance management of the nuclear power plant was introduced in 2009. After the Fukushima Daiichi nuclear disaster (Mar. 11th 2011), the situation surrounding the nuclear industry has dramatically changed, and the requirement for maintenance management of nuclear power plants is pushed for more stringent nuclear safety regulations. The new inspection standard requires enhancing equipment maintenance. It is necessary to enhance maintenance of not only equipment but also piping and pipe support. In this paper, we built the methodology for enhancing maintenance plan by rationalizing and visualizing of piping and pipe support based on the “Maintenance Program” in cooperating with 3D-CAD system.

The Development and Utilization of Nuclear Power from the Perspective of National Economic Security

2011 ◽  
Vol 328-330 ◽  
pp. 704-707
Author(s):  
Ze Hao Min
Keyword(s):  
Nuclear Power ◽  
Economic Security ◽  
Security System ◽  
National Economic ◽  
Development And Utilization ◽  
Nuclear Crisis ◽  
National Economic Security

The Japan Fukushima nuclear crisis upgrades global discussion about the development and utilization of the nuclear power while provides experiences and lessons for the development and utilization of China's nuclear power. From the perspective of national economic security, the essay deals with the development and utilization of nuclear power for the purposes and giving some suggestions on developing a strategy to protect national economic security, building national economic security system.

Kerntechnik als Sicherheitsversprechen

2021 ◽  
Vol 29 (84) ◽  
pp. 99-112
Author(s):  
Sascha Brünig
Keyword(s):  
Public Relations ◽  
Nuclear Power ◽  
Nuclear Industry ◽  
Industrial Society ◽  
Electricity Production ◽  
Nuclear Technology ◽  
Reactor Accident ◽  
The West ◽  
The Face ◽  
Public Relations Campaign

Abstract In the mid-1970s, the dangers associated with nuclear power moved to the center of risk debates in Germany. Following the reactor accident at Three Mile Island (1979) and the Chernobyl disaster (1986), the West German nuclear industry’s business prospects severely deteriorated. How did the nuclear industry perceive and confront the challenge of nuclear skepticism? And how did this emerging challenge alter the perceived future of nuclear technology in the Federal Republic and beyond? The article argues that the nuclear industry did not passively accept the »depletion of utopian energies« (J. Habermas) to which the peaceful use of the atom was subjected. Instead, the industry worked to create new (utopian) prospects for nuclear power. The industry’s public relations campaign positioned nuclear power in two interrelated fields of insecurity: the decline of industrial society and environmental crises. Both threats, ran the argument put forth by nuclear proponents, could only be combatted by relying on nuclear power for electricity production. In this way, nuclear power was translated into a comprehensive promise of security that was intended to salvage the future of nuclear power as well as that of its investors in the face of growing anti-nuclear sentiment.

Research Progress on Radiation Shielding Concrete

2012 ◽  
Vol 253-255 ◽  
pp. 303-307 ◽  
Author(s):  
Jing Yang ◽  
Zhen Fu Chen ◽  
Yuan Chu Gan ◽  
Qiu Wang Tao
Keyword(s):  
Nuclear Waste ◽  
Material Properties ◽  
Nuclear Power ◽  
Power Plants ◽  
Radiation Shielding ◽  
Nuclear Industry ◽  
Research Progress ◽  
Concrete Durability ◽  
Technology Research ◽  
Shielding Material

Radiation shielding concrete is widely used in nuclear power plants, accelerators, hospitals, etc. With the development of nuclear industry technology, research on radiation shielding material properties is of great importance. Research on properties of radiation shielding concrete with different aggregates or admixtures and the effect of high temperature on the performance of shielding concrete are introduced. Along with the nuclear waste increase, shielding concrete durability and nuclear waste disposal are getting paramount.

Investigation of In-Flow Fluidelastic Instability of Square Tube Arrays Subjected to Air Cross Flow

2015 ◽  
Author(s):  
Tomomichi Nakamura ◽  
Shinichiro Hagiwara ◽  
Joji Yamada ◽  
Kenji Usuki
Keyword(s):  
Nuclear Power ◽  
Flow Instability ◽  
Flow Direction ◽  
Cross Flow ◽  
Diameter Ratio ◽  
Nuclear Industry ◽  
Flexible Cylinder ◽  
Triangular Arrays ◽  
Tube Arrays ◽  
Fluidelastic Instability

In-flow instability of tube arrays is a recent major issue in heat exchanger design since the event at a nuclear power plant in California [1]. In our previous tests [2], the effect of the pitch-to-diameter ratio on fluidelastic instability in triangular arrays is reported. This is one of the present major issues in the nuclear industry. However, tube arrays in some heat exchangers are arranged as a square array configuration. Then, it is important to study the in-flow instability on the case of square arrays. The in-flow fluidelastic instability of square arrays is investigated in this report. It was easy to observe the in-flow instability of triangular arrays, but not for square arrays. The pitch-to-diameter ratio, P/D, is changed from 1.2 to 1.5. In-flow fluidelastic instability was not observed in the in-flow direction. Contrarily, the transverse instability is observed in all cases including the case of a single flexible cylinder. The test results are finally reported including the comparison with the triangular arrays.

Slow Crack Growth Resistance of Parent and Joint Materials From PE4710 Piping for Safety-Related Nuclear Power Plant Piping

2011 ◽  
Author(s):  
S. Kalyanam ◽  
D.-J. Shim ◽  
P. Krishnaswamy ◽  
Y. Hioe
Keyword(s):  
Crack Growth ◽  
Nuclear Power ◽  
Power Plants ◽  
Slow Crack Growth ◽  
Nuclear Industry ◽  
Pipe Material ◽  
Service Water ◽  
Hdpe Pipe ◽  
Pipe Materials ◽  
Hdpe Pipes

HDPE pipes are considered by the nuclear industry as a potential replacement option to currently employed metallic piping for service-water applications. The pipes operate under high temperatures and pressures. Hence HDPE pipes are being evaluated from perspective of design, operation, and service life requirements before routine installation in nuclear power plants. Various articles of the ASME Code Case N-755 consider the different aspects related to material performance, design, fabrication, and examination of HDPE materials. Amongst them, the material resistance (part of Article 2000) to the slow crack growth (SCG) from flaws/cracks present in HDPE pipe materials is an important concern. Experimental investigations have revealed that there is a marked difference (almost three orders less) in the time to failure when the notch/flaw is in the butt-fusion joint, as opposed to when the notch/flaw is located in the parent HDPE material. As part of ongoing studies, the material resistance to SCG was investigated earlier for unimodal materials. The current study investigated the SCG in parent and butt-fusion joint materials of bimodal HDPE (PE4710) pipe materials acquired from two different manufacturers. The various stages of the specimen deformation and failure during the creep test are characterized. Detailed photographs of the specimen side-surface were used to monitor the specimen damage accumulation and SCG. The SCG was tested using a large specimen (large creep frame) as well as using a smaller size specimen (PENT frame) and the results were compared. Further, the effect of polymer orientation or microstructure in the bimodal HDPE pipe on the SCG was studied using specimens with axial and circumferential notch orientations in the parent pipe material.

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