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.

scholarly journals Review on the Change Law of the Properties and Performance of Polymer-Matrix Nuclear Radiation Shielding Materials Under the Coupling of Nuclear Radiation and Thermal Effects

2021 ◽  
Vol 9 ◽  
Author(s):  
Guang Hu ◽  
Weiqiang Sun ◽  
Yihong Yan ◽  
Rongjun Wu ◽  
Hu Xu
Keyword(s):  
Mechanical Properties ◽  
Polymer Matrix ◽  
Thermal Effects ◽  
Nuclear Power ◽  
Power Plants ◽  
Radiation Shielding ◽  
Nuclear Radiation ◽  
And Performance ◽  
Shielding Material

The polymer-matrix nuclear radiation shielding material is an important component of nuclear power plants. However, its mechanical properties and shielding performance gradually deteriorate due to the long-term synergy of nuclear radiation and thermal effects, which brings hidden dangers to the safe operation of the device. Based on this problem, this article makes a comprehensive review. First, the degradation of mechanical properties and shielding performance of polymer-matrix nuclear radiation materials in service is briefly described. Then, the research methods adopted by scholars to study the change law of properties and performance are introduced, and the main existing difficulties encountered by the study are summarized. Finally, the physical mechanism of the change of material properties is explained in detail, and a reference approach to solving the problem is proposed.

Radio Active Waste Management: Underground Repository Method

2002 ◽  
Author(s):  
Rudrapati Sandesh Kumar ◽  
Payal Shrivastava
Keyword(s):  
Nuclear Waste ◽  
Nuclear Power ◽  
Power Plants ◽  
Research Effort ◽  
Nuclear Industry ◽  
Eastern Bloc ◽  
Spent Fuel ◽  
Acceptable Solution ◽  
Water Reactors ◽  
High Level

Finding a solution for nuclear waste is a key issue, not only for the protection of the environment but also for the future of the nuclear industry. Ten years from now, when the first decisions for the replacement of existing nuclear power plants will have to be made, The general public will require to know the solution for nuclear waste before accepting new nuclear plants. In other words, an acceptable solution for the management of nuclear waste is a prerequisite for a renewal of nuclear power. Most existing wastes are being stored in safe conditions waiting for permanent solution, with some exceptions in the former Eastern Bloc. Temporary surface or shallow storage is a well known technique widely used all over the world. A significant research effort has been made by the author of this paper in the direction of underground repository. The underground repository appears to be a good solution. Trying to transform dangerous long lived radionuclides into less harmful short lived or stable elements is a logical idea. It is indeed possible to incinerate or transmute heavy atoms of long lived elements in fast breeder reactors or even in pressurised or boiling water reactors. There are also new types of reactors which could be used, namely accelerator driven systems. High level and long lived wastes (spent fuel and vitrified waste) contain a mixture of high activity (heat producing) short lived nuclides and low activity long lived alpha emitting nuclides. To avoid any alteration due to temperature of the engineered or geological barrier surrounding the waste underground, it is necessary to store the packages on the surface for several decades (50 years or more) to allow a sufficient temperature decrease before disposing of them underground. In all cases, surface (or shallow) storage is needed as a temporary solution. This paper gives a detailed and comprehensive view of the Deep Geological Repository, providing a pragmatic picture of the means to make this method, a universally acceptable one.

scholarly journals Preliminary Analysis of an Aged RPV Subjected to Station Blackout

Energies ◽  
2021 ◽  
Vol 14 (15) ◽  
pp. 4394
Author(s):  
Rosa Lo Frano ◽  
Salvatore Angelo Cancemi ◽  
Piotr Darnowski ◽  
Riccardo Ciolini ◽  
Sandro Paci
Keyword(s):  
Finite Element ◽  
Material Properties ◽  
Nuclear Power ◽  
Power Plants ◽  
Elevated Temperatures ◽  
Thermomechanical Properties ◽  
Nuclear Industry ◽  
Fe Model ◽  
Term Operation ◽  
Station Blackout

Today, 46% of operating Nuclear Power Plants (NPP) have a lifetime between 31 and 40 years, while 19% have been in operation for more than 40 years. Long Term Operation (LTO) is an urgent requirement for all of the nuclear industry. The aim of this study is to assess the performance of a reactor pressure vessel (RPV) subjected to a station blackout (SBO) event. Alterations suffered by the material properties and creep at elevated temperatures are considered. In this study, coupling between MELCOR and Finite Element Method (FEM) codes is carried out. In the Finite Element (FE) model, the combined effects of ageing and creep are implemented through degraded material properties and a viscoplastic model. The reliability of the model is validated by comparing the FOREVER/C1 experimental results. The results show that the RPV lower head bends downwards with a maximum radial expansion of about 260 mm and RPV thermomechanical properties are reduced by more than 50% at high temperatures. The effects of ageing, creep and long heat-up strongly affect the resistance of the RPV system until the point of compromising it in the absence of/delayed emergency intervention. Aged RPV at end-of-life may collapse earlier, and in less time, with the same accidental conditions.

Recent Patents on Valve Mechanism Device

2020 ◽  
Vol 13 (3) ◽  
pp. 230-241
Author(s):  
Ye Dai ◽  
Hui-Bing Zhang ◽  
Yun-Shan Qi
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Flow Direction ◽  
Research Progress ◽  
Valve Mechanism ◽  
Advantages And Disadvantages ◽  
Valve Structure ◽  
Valve Leakage ◽  
Safety And Reliability

Background: Valves are an important part of nuclear power plants and are the control equipment used in nuclear power plants. It can change the cross-section of the passage and the flow direction of the medium and has the functions of diversion, cutoff, overflow, and the like. Due to the earthquake, the valve leaks, which will cause a major nuclear accident, endangering people's lives and safety. Objective: The purpose of this study is to synthesize the existing valve devices, summarize and analyze the advantages and disadvantages of various devices from many literatures and patents, and solve some problems of existing valves. Methods: This article summarizes various patents of nuclear-grade valve devices and recent research progress. From the valve structure device, transmission device, a detection device, and finally to the valve test, the advantages and disadvantages of the valve are comprehensively analyzed. Results: By summarizing the characteristics of a large number of valve devices, and analyzing some problems existing in the valves, the outlook for the research and design of nuclear power valves was made, and the planning of the national nuclear power strategic goals and energy security were planned. Conclusion: Valve damage can cause serious safety accidents. The most common is valve leakage. Therefore, the safety and reliability of valves must be taken seriously. By improving the transmission of the valve, the problems of complicated valve structure and high cost are solved.

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.

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.

NUCLEAR WASTE MANAGEMENT IN SPAIN: ANALYSIS OF THE CURRENT SITUATION AND ALTERNATIVE STRATEGIES

10.6036/10156 ◽  
2021 ◽  
Vol 96 (4) ◽  
pp. 355-358
Author(s):  
Pablo Fernández Arias ◽  
DIEGO VERGARA RODRIGUEZ
Keyword(s):  
Nuclear Waste ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Spent Fuel ◽  
Nuclear Waste Storage ◽  
Alternative Strategies ◽  
Temporary Storage ◽  
Long Term Storage ◽  
High Level

Centralized Temporary Storage Facility (CTS) is an industrial facility designed to store spent fuel (SF) and high level radioactive waste (HLW) generated at Spanish nuclear power plants (NPP) in a single location. At the end of 2011, the Spanish Government approved the installation of the CTS in the municipality of Villar de Cañas in Cuenca. This approval was the outcome of a long process of technical studies and political decisions that were always surrounded by great social rejection. After years of confrontations between the different political levels, with hardly any progress in its construction, this infrastructure of national importance seems to have been definitively postponed. The present research analyzes the management strategy of SF and HLW in Spain, as well as the alternative strategies proposed, taking into account the current schedule foreseen for the closure of the Spanish NPPs. In view of the results obtained, it is difficult to affirm that the CTS will be available in 2028, with the possibility that its implementation may be delayed to 2032, or even that it may never happen, making it necessary to adopt an alternative strategy for the management of GC and ARAR in Spain. Among the different alternatives, the permanence of the current Individualized Temporary Stores (ITS) as a long-term storage strategy stands out, and even the possibility of building several distributed temporary storage facilities (DTS) in which to store the SF and HLW from several Spanish NPP. Keywords: nuclear waste, storage, nuclear power plants.

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.

scholarly journals Radiation shielding structures : Concepts, behaviour and the role of the heavy weight concrete as a shielding material - Rewiev

2020 ◽  
Vol 21 ◽  
pp. 24-30
Author(s):  
Suha Ismail Ahmed Ali ◽  
Éva Lublóy
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Building Materials ◽  
Physical And Mechanical Properties ◽  
Cost Effective ◽  
Radiation Shielding ◽  
Radiation Emission ◽  
Neutron Shielding ◽  
Heavy Weight ◽  
Industrial Buildings

The construction of radiation shielding buildings still developed. Application of ionizing radiations became necessary for different reasons, like electricity generation, industry, medical (therapy treatment), agriculture, and scientific research. Different countries all over the world moving toward energy saving, besides growing the demand for using radiation in several aspects. Nuclear power plants, healthcare buildings, industrial buildings, and aerospace are the main neutrons and gamma shielding buildings. Special design and building materials are required to enhance safety and reduce the risk of radiation emission. Radiation shielding, strength, fire resistance, and durability are the most important properties, cost-effective and environmentally friendly are coming next. Heavy-weight concrete (HWC) is used widely in neutron shielding materials due to its cost-effectiveness and worthy physical and mechanical properties. This paper aims to give an overview of nuclear buildings, their application, and behaviour under different radiations. Also to review the heavy-weight concrete and heavy aggregate and their important role in developing the neutrons shielding materials. Conclusions showed there are still some gaps in improving the heavy-weight concrete (HWC) properties.

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