2025 Nuclear Code: The Vision for the Future of ASME Nuclear Codes and Standards

2018 ◽  
Author(s):  
Dale E. Matthews ◽  
Ralph S. Hill ◽  
Charles W. Bruny
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
New Technologies ◽  
Fusion Energy ◽  
Life Cycle Management ◽  
Economic Benefits ◽  
Plant Design ◽  
Element Analysis ◽  
Energy Devices ◽  
The Future

ASME Nuclear Codes and Standards are used worldwide in the construction, inspection, and repair of commercial nuclear power plants. As the industry looks to the future of nuclear power and some of the new plant designs under development, there will be some significant departures from the current light water reactor (LWR) technology. Some examples are gas-cooled and liquid metal-cooled high temperature reactors (HTRs), small modular reactors (SMRs), and fusion energy devices that are currently under development. Many of these designs will have different safety challenges from the current LWR fleet. Variations of the current LWR technology are also expected to remain in use for the foreseeable future. Worldwide, many LWRs are planned or are already under construction. However, technology for construction of these plants has advanced considerably since most of the current construction codes were written. As a result, many modern design and fabrication methods available today, which provide both safety and economic benefits, cannot be fully utilized since they are not addressed by Code rules. For ASME Nuclear Codes and Standards to maintain and enhance their position as the worldwide leader in the nuclear power industry, they will need to be modernized to address these items. Accordingly, the ASME Nuclear Codes and Standards organizations have initiated the “2025 Nuclear Code” initiative. The purpose of this initiative is to modernize all aspects of ASME’s Nuclear Codes and Standards to adopt new technologies in plant design, construction, and life cycle management. Examples include modernized finite element analysis and fatigue rules, and incorporation of probabilistic and risk-informed methodology. This paper will present the vision for the 2025 ASME Nuclear Codes and Standards and will discuss some of the key elements that are being considered.

Research on Weight-Reduced Optimization Design of Large Nuclear Power Low Pressure Cylinder

2017 ◽  
Author(s):  
Yifeng Hu ◽  
Xingzhu Ye ◽  
Gang Chen
Keyword(s):  
Nuclear Power ◽  
Optimization Design ◽  
Power Plants ◽  
Low Pressure ◽  
Economic Benefits ◽  
Steam Turbines ◽  
Design Standards ◽  
Element Analysis ◽  
And Performance ◽  
Stress And Deformation

Steam turbines need to be safer and more reliable when used in nuclear power plants. In order to ensure long-term reliability of nuclear power equipment, a high safety factor is usually adopted in the design of low-pressure (LP) inner casing of steam turbines. It not only leads to larger self-weight of LP outer casings and fundamental load, but also causes higher manufacturing and transportation costs. In this paper, the stress and deformation behaviors of the LP outer casings of steam turbines are first evaluated using the numerical finite element analysis. Then, two optimization design methods, size optimization and topology optimization are used to conduct the weight reduced optimization design of inner casing, in combination with the design standards, so that the structural efficiency and performance of LP inner casings are achieved. At the same time, the self-weight and related costs are also greatly reduced. This study proposes a more optimized structural design of LP inner casings of steam turbines, and it offers considerable economic benefits.

Use of HFE Guidance for the Review of Nuclear Power Plants

2020 ◽  
Vol 64 (1) ◽  
pp. 1-5
Author(s):  
John O’Hara ◽  
Stephen Fleger
Keyword(s):  
Interface Design ◽  
Nuclear Power ◽  
Power Plants ◽  
Health And Safety ◽  
Nuclear Regulatory Commission ◽  
Program Review ◽  
Human Factors Engineering ◽  
Plant Design ◽  
Review Model ◽  
Regulatory Commission

The U.S. Nuclear Regulatory Commission (NRC) evaluates the human factors engineering (HFE) of nuclear power plant design and operations to protect public health and safety. The HFE safety reviews encompass both the design process and its products. The NRC staff performs the reviews using the detailed guidance contained in two key documents: the HFE Program Review Model (NUREG-0711) and the Human-System Interface Design Review Guidelines (NUREG-0700). This paper will describe these two documents and the method used to develop them. As the NRC is committed to the periodic update and improvement of the guidance to ensure that they remain state-of-the-art design evaluation tools, we will discuss the topics being addressed in support of future updates as well.

Operating Reliability Assessment of FPGA-Based NPP I&C Systems: Approach, Technique and Implementation

2017 ◽  
Author(s):  
Eugene Babeshko ◽  
Ievgenii Bakhmach ◽  
Vyacheslav Kharchenko ◽  
Eugene Ruchkov ◽  
Oleksandr Siora
Keyword(s):  
Integrated Circuits ◽  
Nuclear Power ◽  
Power Plants ◽  
Failure Modes ◽  
New Technologies ◽  
Systems Approach ◽  
Reliability Assessment ◽  
Design Stage ◽  
Fpga Design ◽  
Operating Reliability

Operating reliability assessment of instrumentation and control systems (I&Cs) is always one of the most important activities, especially for critical domains like nuclear power plants (NPPs). Intensive use of relatively new technologies like field programmable gate arrays (FPGAs) in I&C which appear in upgrades and in newly built NPPs makes task to develop and validate advanced operating reliability assessment methods that consider specific technology features very topical. Increased integration densities make the reliability of integrated circuits the most crucial point in modern NPP I&Cs. Moreover, FPGAs differ in some significant ways from other integrated circuits: they are shipped as blanks and are very dependent on design configured into them. Furthermore, FPGA design could be changed during planned NPP outage for different reasons. Considering all possible failure modes of FPGA-based NPP I&C at design stage is a quite challenging task. Therefore, operating reliability assessment is one of the most preferable ways to perform comprehensive analysis of FPGA-based NPP I&Cs. This paper summarizes our experience on operating reliability analysis of FPGA based NPP I&Cs.

Development of 1500-r/min 75-Inch Ultra-Long Last Stage Blades for Nuclear Steam Turbine

2017 ◽  
Author(s):  
Deqi Yu ◽  
Jiandao Yang ◽  
Wei Lu ◽  
Daiwei Zhou ◽  
Kai Cheng ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Large Scale ◽  
Vibration Monitoring ◽  
Steam Turbines ◽  
Element Analysis ◽  
Rotating Blade ◽  
Lifetime Analysis ◽  
Computational Fluid Dynamics Cfd ◽  
Last Stage

The 1500-r/min 1905mm (75inch) ultra-long last three stage blades for half-speed large-scale nuclear steam turbines of 3rd generation nuclear power plants have been developed with the application of new design features and Computer-Aided-Engineering (CAE) technologies. The last stage rotating blade was designed with an integral shroud, snubber and fir-tree root. During operation, the adjacent blades are continuously coupled by the centrifugal force. It is designed that the adjacent shrouds and snubbers of each blade can provide additional structural damping to minimize the dynamic stress of the blade. In order to meet the blade development requirements, the quasi-3D aerodynamic method was used to obtain the preliminary flow path design for the last three stages in LP (Low-pressure) casing and the airfoil of last stage rotating blade was optimized as well to minimize its centrifugal stress. The latest CAE technologies and approaches of Computational Fluid Dynamics (CFD), Finite Element Analysis (FEA) and Fatigue Lifetime Analysis (FLA) were applied to analyze and optimize the aerodynamic performance and reliability behavior of the blade structure. The blade was well tuned to avoid any possible excitation and resonant vibration. The blades and test rotor have been manufactured and the rotating vibration test with the vibration monitoring had been carried out in the verification tests.

scholarly journals CS Measures for Nuclear Power Plant Protection: A Systematic Literature Review

Signals ◽  
2021 ◽  
Vol 2 (4) ◽  
pp. 803-819
Author(s):  
Nabin Chowdhury
Keyword(s):  
Literature Review ◽  
Systematic Literature Review ◽  
Cyber Security ◽  
Nuclear Power ◽  
Power Plants ◽  
New Technologies ◽  
Plant Protection ◽  
Security Requirements ◽  
Protection Measures ◽  
Defense In Depth

As digital instrumentation in Nuclear Power Plants (NPPs) is becoming increasingly complex, both attack vectors and defensive strategies are evolving based on new technologies and vulnerabilities. Continued efforts have been made to develop a variety of measures for the cyber defense of these infrastructures, which often consist in adapting security measures previously developed for other critical infrastructure sectors according to the requirements of NPPs. That being said, due to the very recent development of these solutions, there is a lack of agreement or standardization when it comes to their adoption at an industrial level. To better understand the state of the art in NPP Cyber-Security (CS) measures, in this work, we conduct a Systematic Literature Review (SLR) to identify scientific papers discussing CS frameworks, standards, guidelines, best practices, and any additional CS protection measures for NPPs. From our literature analysis, it was evidenced that protecting the digital space in NPPs involves three main steps: (i) identification of critical digital assets; (ii) risk assessment and threat analysis; (iii) establishment of measures for NPP protection based on the defense-in-depth model. To ensure the CS protection of these infrastructures, a holistic defense-in-depth approach is suggested in order to avoid excessive granularity and lack of compatibility between different layers of protection. Additional research is needed to ensure that such a model is developed effectively and that it is based on the interdependencies of all security requirements of NPPs.

The Climate Vulnerabilities of Global Nuclear Power

2019 ◽  
Vol 19 (4) ◽  
pp. 3-13 ◽  
Author(s):  
Sarah M. Jordaan ◽  
Afreen Siddiqi ◽  
William Kakenmaster ◽  
Alice C. Hill
Keyword(s):  
Climate Change ◽  
Sea Level Rise ◽  
Sea Level ◽  
Nuclear Power ◽  
Power Plants ◽  
Extreme Event ◽  
International Standards ◽  
Plant Design ◽  
Low Carbon ◽  
And Sea Level

Nuclear power—a source of low-carbon electricity—is exposed to increasing risks from climate change. Intensifying storms, droughts, extreme precipitation, wildfires, higher temperatures, and sea-level rise threaten supply disruptions and facility damage. Approximately 64 percent of installed capacity commenced operation between thirty and forty-eight years ago, before climate change was considered in plant design or construction. Globally, 516 million people reside within a fifty mile (80 km) radius of at least one operating nuclear power plant, and 20 million reside within a ten mile (16 km) radius, and could face health and safety risks resulting from an extreme event induced by climate change. Roughly 41 percent of nuclear power plants operate near seacoasts, making them vulnerable to increasing storm intensity and sea-level rise. Inland plants face exposure to other climate risks, such as increasingly severe wildfires and warmer water temperatures. No entity has responsibility for conducting risk assessments that adequately evaluate the climate vulnerabilities of nuclear power and the subsequent threats to international energy security, the environment, and human health. A comprehensive risk assessment by international agencies and the development of national and international standards is necessary to mitigate risks for new and existing plants.

Electrical Power System Modeling of Atucha II Nuclear Power Plant Using etap

2019 ◽  
Vol 5 (2) ◽  
Author(s):  
Nicolás Alejandro Malinovsky
Keyword(s):  
Power System ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Electrical Power ◽  
Analysis Tool ◽  
Plant Design ◽  
Electrical Power System ◽  
Power Circuit ◽  
Electrical Systems

This work shows the introduction of the Electrical Power System Analysis (etap) software as a calculation and analysis tool for power electrical systems of the nuclear power plants (NPP) under the orbit of Nucleoeléctrica Argentina S.A (NASA). Through the use of the software, the model of the electrical power system of the Atucha II NPP was developed. To test the functionality of the modeled electrical power circuit, studies of load flow and short-circuit analysis were conducted, yielding satisfactory results, which were contrasted with the plant design values. Once the model has been validated, this will be the basis for carrying out different studies in the plant through simulation. Furthermore, with the incorporation of etap as a fundamental calculation and analysis tool for power electrical systems at the company's engineering departments, it is expected to improve the safety, operation, quality, reliability, and maintenance of both the Atucha II NPP electrical power system and the other nuclear power plants operated by Nucleoeléctrica Argentina S.A.

Canadian Nuclear Safety Commission: Readiness to Regulate SMRs in Canada

2011 ◽  
Author(s):  
S. Herstead ◽  
M. de Vos ◽  
S. Cook
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Safety Analysis ◽  
Reactor Power ◽  
Nuclear Safety ◽  
Regulatory Framework ◽  
Nuclear Industry ◽  
Plant Design ◽  
Regulatory Documents

The success of any new build project is reliant upon all stakeholders — applicants, vendors, contractors and regulatory agencies — being ready to do their part. Over the past several years, the Canadian Nuclear Safety Commission (CNSC) has been working to ensure that it has the appropriate regulatory framework and internal processes in place for the timely and efficient licensing of all types of reactor, regardless of size. This effort has resulted in several new regulatory documents and internal processes including pre-project vendor design reviews. The CNSC’s general nuclear safety objective requires that nuclear facilities be designed and operated in a manner that will protect the health, safety and security of persons and the environment from unreasonable risk, and to implement Canada’s international commitments on the peaceful use of nuclear energy. To achieve this objective, the regulatory approach strikes a balance between pure performance-based regulation and prescriptive-based regulation. By utilizing this approach, CNSC seeks to ensure a regulatory environment exists that encourages innovation within the nuclear industry without compromising the high standards necessary for safety. The CNSC is applying a technology neutral approach as part of its continuing work to update its regulatory framework and achieve clarity of its requirements. A reactor power threshold of approximately 200 MW(th) has been chosen to distinguish between large and small reactors. It is recognized that some Small Modular Reactors (SMRs) will be larger than 200 MW(th), so a graded approach to achieving safety is still possible even though Nuclear Power Plant design and safety requirements will apply. Design requirements for large reactors are established through two main regulatory documents. These are RD-337 Design for New Nuclear Power Plants, and RD-310 Safety Analysis for Nuclear Power Plants. For reactors below 200 MW(th), the CNSC allows additional flexibility in the use of a graded approach to achieving safety in two new regulatory documents: RD-367 Design of Small Reactors and RD-308 Deterministic Safety Analysis for Small Reactors. The CNSC offers a pre-licensing vendor design review as an optional service for reactor facility designs. This review process is intended to provide early identification and resolution of potential regulatory or technical issues in the design process, particularly those that could result in significant changes to the design or analysis. The process aims to increase regulatory certainty and ultimately contribute to public safety. This paper outlines the CNSC’s expectations for applicant and vendor readiness and discusses the process for pre-licensing reviews which allows vendors and applicants to understand their readiness for licensing.

Investigation of Blast Wave Effects on Containment Wall and Steam Generator

2018 ◽  
Author(s):  
Tae Jin Kim ◽  
Yoon-Suk Chang
Keyword(s):  
Blast Wave ◽  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Failure Criteria ◽  
High Energy ◽  
Element Analysis ◽  
Secondary Damage ◽  
Main Steam Line ◽  
Main Steam

When a sudden rupture occurs in high energy lines such as MSL (Main Steam Line) and safety injection line of nuclear power plants, ejection of inner fluid with high temperature and pressure causes blast wave, and may lead to secondary damage of adjacent major components and/or structures. The objective of this study is to assess integrity of containment wall and steam generator due to the blast wave under a postulated high energy line break condition at the MSL piping. In this context, a preliminary analysis was conducted to examine the blast wave simulation using coupled Eulerian-Lagrangian technique. Subsequently, a finite element analysis was carried out to assess integrity of the structures. As typical results, strain and stress values were calculated at the containment wall and steam generator, which did not exceed their failure criteria.

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