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.

Accident Sequence Precursor Analysis of Daya Bay and Lingao Nuclear Power Plants

2010 ◽  
Author(s):  
Pingping Liu ◽  
Haiying Xi
Keyword(s):  
Nuclear Power ◽  
Safety Assessment ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Regulatory Commission ◽  
Daya Bay ◽  
Probabilistic Safety Assessment ◽  
Regulatory Commission ◽  
Probabilistic Safety

Firstly this article describes how to analyze accident sequence precursor in U.S. nuclear regulatory commission. Following this method, the licensee operational events, internal operational events and inspection reports of Daya Bay and Lingao Nuclear Power Plants were reviewed to identify the precursors with support of probabilistic safety assessment. All the identified precursors were calculated, documented and ranked. Then the trends on precursors can be obtained. Finally this article analyzes the trends available and gains many beneficial insights.

Technologies Used for D&D of the A-1 NPP in Slovakia and Their Comparison With Advanced Worldwide Approaches

2001 ◽  
Author(s):  
Kamil Kravárik ◽  
Vladimír Míchal ◽  
Peter Menyhardt
Keyword(s):  
Radioactive Waste ◽  
Nuclear Power ◽  
Power Plants ◽  
Slovak Republic ◽  
Liquid Radioactive Waste ◽  
Nuclear Regulatory Commission ◽  
Technical Level ◽  
Nuclear Facilities ◽  
Energy Agency ◽  
Regulatory Commission

Abstract This paper deals with technologies used for decommissioning and decontamination of the A-1 Nuclear Power Plant in Slovakia and their comparison with advanced worldwide approaches. Present status and main results in the field of D&D of this first Czechoslovak NPP A-1 at Jaslovské Bohunice are described. NPP A-1 has one unit with reactor cooled by CO2 and moderated by heavy water. Plant was in operation from 1972 to 1977 and its final shutdown and closure were done due to relatively serious accident. The A-1 NPP Decommissioning Project – I. phase is performed at the present time and represents the most important project of NPP decommissioning in Central Europe. The main goal of the project is to achieve radiologically safe status of the NPP. It includes following activities: • conditioning, storage and disposal of liquid radioactive waste, solid and metallic radioactive waste, sludge and sorbents, • development, manufacture and verification of advanced methodologies and technologies for D&D of nuclear facilities, • decontamination of specified equipment and structures to reduce free activity, • technical support and preparation of following phases within the A-1 NPP overall decommissioning process. The project should give the complex solution of problems related to decommissioning and decontamination of NPPs in Slovakia. Verified methodology and technology should be used as a generic approach for decommissioning of the V-1, V-2 (Jaslovské Bohunice) and Mochovce Nuclear Power Plants as well as the other European NPPs with WWER reactors. Significant part of paper deals with following issues within D&D of the A-1 NPP: • computer aided technologies, • decontamination, • dismantling, demolishing and remote handling manipulators, • dosimetry measurements within D&D, • radioactive waste management. This paper also includes basic comparison with advanced worldwide approaches to decommissioning and decontamination mainly in USA, Japan and West Europe and the recommendations are done when it is possible. The comparison shows that trends in the field of D&D in the Slovak Republic are compatible and comparable with the most significant world trends. It is noted that some sorts of D&D technologies like for example telerobotic systems developed in the world are at the relatively higher technical level. Decommissioning technologies in Slovakia should be permanently improved on the base of experiences from home and abroad industry and from the real operation. It is supposed that after short time could be achieved technical level comparable with the best D&D robots and manipulators. A basic strategy of NPP decommissioning in the Slovak Republic is regulated by standards, which are in accordance with recommendations of international bodies like the International Atomic Energy Agency, European Commission, U.S. Nuclear Regulatory Commission, OECD Nuclear Energy Agency etc. In the field of NPP D&D the Slovak Republic co-operates with many international organizations and also with main active countries in D&D like Germany, France, Belgium, Great Britain, USA, Japan, Russian Federation, Hungary, Poland and Czech Republic. Intensive international co-operation at all levels has already been established at the present time.

scholarly journals Remarks on Consistent Development of Plant Nodalizations: An Example of Application to the ROSA Integral Test Facility

2012 ◽  
Vol 2012 ◽  
pp. 1-17 ◽  
Author(s):  
J. Freixa ◽  
A. Manera
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Large Scale ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Test Facility ◽  
Integral Test ◽  
Energy Agency ◽  
Water Reactors ◽  
Regulatory Commission

Experimental results obtained at integral test facilities (ITFs) are used in the validation process of system codes for the transient analyses of light water reactors (LWRs). The expertise and guidelines derived from this work are later applied to transient analyses of nuclear power plants (NPPs). However, the boundary conditions at the NPPs will always differ from those at the ITF, and hence, the soundness of the ITF model needs to be maximized. An unaltered ITF nodalization should prove to be able to simulate as many tests as possible, before any conclusion is derived to NPP analyses. The STARS group at the Paul Scherrer Institut (PSI) actively participates in several international programs, where ITFs are being used (e.g., ROSA, PKL). Several tests carried out at the ROSA large-scale test facility operated by the Japan Atomic Energy Agency (JAEA) have been simulated in recent years by using the United States Nuclear Regulatory Commission (US-NRC) system code TRACE. In this paper, 5 different posttest analyses are presented, along with the evolution of the employed TRACE nodalization and the process followed to track the consistency of the nodalization modifications. The ROSA TRACE nodalization provided results in a reasonable agreement with all 5 experiments.

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.

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.

Expectations for Inservice Testing Programs at New Nuclear Power Plants

2017 ◽  
Author(s):  
Thomas G. Scarbrough
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Reactor Core ◽  
Nuclear Regulatory Commission ◽  
Lessons Learned ◽  
Mechanical Equipment ◽  
American Society ◽  
Regulatory Commission ◽  
Safety Systems

In a series of Commission papers, the U.S. Nuclear Regulatory Commission (NRC) described its policy for inservice testing (IST) programs to be developed and implemented at nuclear power plants licensed under 10 CFR Part 52. This paper discusses the expectations for IST programs based on those Commission policy papers as applied in the NRC staff review of combined license (COL) applications for new reactors. For example, the design and qualification of pumps, valves, and dynamic restraints through implementation of American Society of Mechanical Engineers (ASME) Standard QME-1-2007, “Qualification of Active Mechanical Equipment Used in Nuclear Power Plants,” as accepted in NRC Regulatory Guide (RG) 1.100 (Revision 3), “Seismic Qualification of Electrical and Active Mechanical Equipment and Functional Qualification of Active Mechanical Equipment for Nuclear Power Plants,” will enable IST activities to assess the operational readiness of those components to perform their intended functions. ASME has updated the Operation and Maintenance of Nuclear Power Plants (OM Code) to improve the IST provisions for pumps, valves, and dynamic restraints that are incorporated by reference in the NRC regulations with applicable conditions. In addition, lessons learned from performance experience and testing of motor-operated valves (MOVs) will be implemented as part of the IST programs together with application of those lessons learned to other power-operated valves (POVs). Licensee programs for the Regulatory Treatment of Non-Safety Systems (RTNSS) will be implemented for components in active nonsafety-related systems that are the first line of defense in new reactors that rely on passive systems to provide reactor core and containment cooling in the event of a plant transient. This paper also discusses the overlapping testing provisions specified in ASME Standard QME-1-2007; plant-specific inspections, tests, analyses, and acceptance criteria; the applicable ASME OM Code as incorporated by reference in the NRC regulations; specific license conditions; and Initial Test Programs as described in the final safety analysis report and applicable RGs. Paper published with permission.

Probabilistic Fracture Mechanics Benchmarking Study Involving the xLPR and PASCAL-SP Codes: Analysis by PASCAL-SP

2020 ◽  
Author(s):  
Akihiro Mano ◽  
Jinya Katsuyama ◽  
Yinsheng Li
Keyword(s):  
Fracture Mechanics ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Nuclear Power ◽  
Power Plants ◽  
Corrosion Cracking ◽  
Energy Agency ◽  
Pressurized Water ◽  
Water Reactor ◽  
Probabilistic Fracture Mechanics

Abstract A probabilistic fracture mechanics (PFM) analysis code, PASCAL-SP, has been developed by Japan Atomic Energy Agency (JAEA) to evaluate the failure probability of piping within nuclear power plants considering aged-related degradations such as stress corrosion cracking and fatigue for both pressurized water reactor and boiling water reactor environments. To strengthen the applicability of PASCAL-SP, a benchmarking study is being performed with a PFM analysis code, xLPR, which has been developed by U.S.NRC in collaboration with EPRI. In this benchmarking study, deterministic and probabilistic analyses are undertaken on primary water stress corrosion cracking using the common analysis conditions. A deterministic analysis on the weld residual stress distributions is also considered. These analyses are carried out by U.S.NRC and JAEA independently using their own codes. Currently, the deterministic analyses by both xLPR and PASCAL-SP codes have been finished and probabilistic analyses are underway. This paper presents the details of conditions and comparisons of the results between the two aforementioned codes for the deterministic analyses. Both codes were found to provide almost the same results including the values of stress intensity factor. The conditions and results of the probabilistic analysis obtained from PASCAL-SP are also discussed.

Incorporation of Risk Insights in the Regulatory Treatment of Nuclear Power Plant Structures, Systems, and Components

2002 ◽  
Author(s):  
Eugene Imbro ◽  
Thomas G. Scarbrough
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Special Treatment ◽  
Regulatory Commission ◽  
High Level

The U.S. Nuclear Regulatory Commission (NRC) has established an initiative to risk-inform the requirements in Title 10 of the Code of Federal Regulations (10 CFR) for the regulatory treatment of structures, systems, and components (SSCs) used in commercial nuclear power plants. As discussed in several Commission papers (e.g., SECY-99-256 and SECY-00-0194), Option 2 of this initiative involves categorizing plant SSCs based on their safety significance, and specifying treatment that would provide an appropriate level of confidence in the capability of those SSCs to perform their design functions in accordance with their risk categorization. The NRC has initiated a rulemaking effort to allow licensees of nuclear power plants in the United States to implement the Option 2 approach in lieu of the “special treatment requirements” of the NRC regulations. In a proof-of-concept effort, the NRC recently granted exemptions from the special treatment requirements for safety-related SSCs categorized as having low risk significance by the licensee of the South Texas Project (STP) Units 1 and 2 nuclear power plant, based on a review of the licensee’s high-level objectives of the planned treatment for safety-related and high-risk nonsafety-related SSCs. This paper discusses the NRC staff’s views regarding the treatment of SSCs at STP described by the licensee in its updated Final Safety Analysis Report (FSAR) in support of the exemption request, and provides the status of rulemaking that would incorporate risk insights into the treatment of SSCs at nuclear power plants.

scholarly journals Using Wireless Sensor Networks to Achieve Intelligent Monitoring for High-Temperature Gas-Cooled Reactor

2017 ◽  
Vol 2017 ◽  
pp. 1-8
Author(s):  
Jianghai Li ◽  
Jia Meng ◽  
Xiaojing Kang ◽  
Zhenhai Long ◽  
Xiaojin Huang
Keyword(s):  
High Temperature ◽  
Nuclear Power ◽  
Power Plants ◽  
Heterogeneous Data ◽  
Wireless Sensor ◽  
Support Vector ◽  
Intelligent Monitoring ◽  
Novel Algorithms ◽  
Non Gaussian ◽  
High Temperature Gas

High-temperature gas-cooled reactors (HTGR) can incorporate wireless sensor network (WSN) technology to improve safety and economic competitiveness. WSN has great potential in monitoring the equipment and processes within nuclear power plants (NPPs). This technology not only reduces the cost of regular monitoring but also enables intelligent monitoring. In intelligent monitoring, large sets of heterogeneous data collected by the WSN can be used to optimize the operation and maintenance of the HTGR. In this paper, WSN-based intelligent monitoring schemes that are specific for applications of HTGR are proposed. Three major concerns regarding wireless technology in HTGR are addressed: wireless devices interference, cybersecurity of wireless networks, and wireless standards selected for wireless platform. To process nonlinear and non-Gaussian data obtained by WSN for fault diagnosis, novel algorithms combining Kernel Entropy Component Analysis (KECA) and support vector machine (SVM) are developed.

US Nuclear Regulatory Commission Perspective on Buried and Underground Piping Issues at Nuclear Power Plants in the United States

2014 ◽  
Author(s):  
David Alley
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Action Plan ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Buried Pipe ◽  
Development Organizations ◽  
Regulatory Commission ◽  
License Renewal

This paper provides a historical perspective on the need for, and development of, buried and underground piping tanks programs at nuclear power plants. Nuclear power plant license renewal activities, Nuclear Regulatory Commission Buried Piping Action Plan, and the rationale for addressing the issue of buried pipe through an industry initiative as opposed to regulation are discussed. The paper also addresses current NRC activities including the results of Nuclear Regulatory Commission inspections of buried piping programs at nuclear power plants as well as Nuclear Regulatory Commission involvement in industry and standards development organizations. Finally, the paper outlines the Nuclear Regulatory Commission’s future plans concerning the issue of buried piping at US nuclear power plants.

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