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.

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.

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.

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.

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.

Overview of Boiling Water Reactor Steam Dryer Alternating Stress Assessment Procedures

2018 ◽  
Vol 4 (2) ◽  
Author(s):  
Stephen A. Hambric ◽  
Samir Ziada ◽  
Richard J. Morante
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
High Cycle Fatigue ◽  
Structural Integrity ◽  
Fatigue Cracking ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Boiling Water ◽  
Cycle Fatigue ◽  
Regulatory Commission

The United States Nuclear Regulatory Commission (USNRC) has approved several extended power uprates (EPU) for Boiling Water Reactors (BWRs). In some of the BWRs, operating at the higher EPU power levels and flow rates led to high-cycle fatigue damage of Steam Dryers, including the generation of loose parts. Since those failures occurred, all BWR owners proposing EPUs have been required by the USNRC to ensure that the steam dryers would not experience high-cycle fatigue cracking. This paper provides an overview of BWR steam dryer design; the fatigue failures that occurred at the Quad Cities (QC) nuclear power plants and their root causes; a brief history of BWR EPUs; and a discussion of steam dryer modifications/replacements, alternating stress mechanisms on steam dryers, and structural integrity evaluation methods (static and alternating stress).

scholarly journals Regulatory Treatment of Nonsafety Systems in Passive Advanced Light-Water Reactors

2014 ◽  
Author(s):  
Thomas Scarbrough
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Reactor Core ◽  
Nuclear Regulatory Commission ◽  
Active Systems ◽  
Light Water ◽  
Defense In Depth ◽  
Water Reactors ◽  
Regulatory Commission ◽  
High Level

Some new nuclear power plants have advanced light-water reactor (ALWR) designs with passive safety systems that rely on natural forces, such as density differences, gravity, and stored energy, to supply safety-injection water and to provide reactor-core and containment cooling. Active systems in such passive ALWR designs are categorized as nonsafety systems with limited exceptions. Active systems in passive ALWR designs provide the first line of defense to reduce challenges to the passive systems in the event of a transient at the nuclear power plant. Active systems that provide a defense-in-depth function in passive ALWR designs need not meet all of the acceptance criteria for safety-related systems. However, there should be a high level of confidence that these active systems will be available and reliable when challenged. Multiple activities will provide confidence in the capability of these active systems to perform their defense-in-depth functions; these are collectively referred to as the Regulatory Treatment of Nonsafety Systems (RTNSS) program. The U.S. Nuclear Regulatory Commission (NRC) addresses policy and technical issues associated with RTNSS equipment in passive ALWRs in several documents. This paper discusses the NRC staff’s review of pumps, valves, and dynamic restraints within the scope of the RTNSS program in passive ALWRs. Paper published with permission.

scholarly journals Inservice Testing Program Improvements for New Reactors: Small Modular Reactors

2014 ◽  
Author(s):  
Ronald 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, including more traditional evolutionary designs, passive reactor designs, and small modular reactors (SMRs). ASME (formerly the American Society of Mechanical Engineers) provides specific codes used to perform inspections and testing, both preservice and inservice, 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 certification (DC) and combined license (COL) applications under Part 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants,” of Title 10, “Energy,” of the Code of Federal Regulations (10 CFR Part 52) (Reference 1). The 2012 Edition of the ASME OM Code, Operation and Maintenance of Nuclear Power Plants, 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 after January 1, 2000. The ASME New Reactors OM Code (NROMC) Task Group (TG) is assigned the task of ensuring that the preservice testing (PST) and inservice testing (IST) provisions in the ASME OM Code are adequate to provide reasonable assurance that pumps, valves, and dynamic restraints (snubbers) for post-2000 plants will operate when needed. Currently, the NROMC TG is preparing proposed guidance for the treatment of active pumps, valves, and dynamic restraints with high safety significance in nonsafety systems for passive post-2000 plants, including SMRs. (Note: For purposes of this paper, “post-2000 plant” and “new reactor” are used interchangeably throughout.) Paper published with permission.

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.

Regulatory and Industry Guidance on Use of Risk Technology in Nuclear Power Plant Applications

2007 ◽  
Author(s):  
Mansoor H. Sanwarwalla
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Regulatory Body ◽  
Current Application ◽  
Technical Specifications ◽  
Regulatory Commission ◽  
The Impact

Since the United States Nuclear Regulatory Commission (USNRC) published its landmark “Reactor Safety Study — An Assessment of Accident Risks in U. S. Commercial Nuclear Power Plants” in late 1975, commercial nuclear power industry, encouraged by the USNRC, have since then been applying Probabilistic Risk Assessment (PRAs) in their nuclear power units in areas of in-service testing, in-service inspection, quality assurance, technical specifications, maintenance, etc. To guide and regulate the industry in use of PRAs, Regulatory Guides and Standards have been written and are being revised continuously by the USNRC, American Society of Mechanical Engineers (ASME) and American Nuclear Society (ANS). The current use of PRA takes credit for single failure criterion based on applicability of codes and standards. The proposed new USNRC regulation 10 CFR Part 53 applicable for all reactor technologies is silent on the applicability of current standards endorsed by the regulatory body. The impact of the proposed new rule to both new and the current application needs to be studied. This paper will review the application of the various guidance documents for their use in commercial nuclear power plants with emphasis on the new generation nuclear power plants.

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.

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