scholarly journals Combustible Performance of Floor Finishing Materials and Coating Applied to Nuclear Power Plants

2020 ◽  
Vol 20 (4) ◽  
pp. 139-144
Author(s):  
Inkyu Kwon
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Construction Materials ◽  
Nuclear Regulatory Commission ◽  
Plant Industry ◽  
Testing Methods ◽  
Regulatory Commission

Finishing construction materials applied to nuclear power plants and other attached structures are manufactured domestically; however, their fire-related performance has not yet been clarified, and data exist only for common materials with general purposes. Finishing construction materials must meet the requirement of Nuclear Regulatory Commission (NRC), which is regarded as a global standard in the nuclear power plant industry. In this study, to support data when a new guideline for evaluation of fire safety in nuclear power plant and the attached structures thereof are prepared, the finishing materials on the floor and the coating applied onto the floor and other portions were selected and tested using related standards of two nations. The results showed that there were differences in the manner of evaluation and testing. Moreover, certain criteria did not meet Korean standards. Nevertheless, most criteria were satisfied with testing methods suggested by the NRC.

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 Overview of NRC NUREG-1482, Revision 2, Guidelines for Inservice Testing at Nuclear Power Plants: Inservice Testing of Pumps and Valves, and Inservice Examination and Testing of Dynamic Restrains (Snubbers) at Nuclear Power Plants

2014 ◽  
Author(s):  
Gurjendra S. Bedi
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Regulatory Commission ◽  
Lessons Learned ◽  
Regulatory Guidance ◽  
American Society ◽  
Regulatory Commission

The U.S. Nuclear Regulatory Commission (NRC) staff issued Revision 2 to NUREG-1482, “Guidelines for Inservice Testing at Nuclear Power Plant,” to assist the nuclear power plant licensees in establishing a basic understanding of the regulatory basis for pump and valve inservice testing (IST) programs and dynamic restraints (snubbers) inservice examination and testing programs. Since the Revision 1 issuance of NUREG-1482, certain tests and measurements required by earlier editions and addenda of the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) have been clarified, updated, revised or eliminated. The revision to NUREG-1482 incorporates and addresses those changes, and includes the IST programs guidelines related to new reactors. The revised guidance incorporates lessons learned and experience gained since the last issue. This paper provides an overview of the contents of the NUREG-1482 and those changes and discusses how they affect NRC guidance on implementing pump and valve inservice testing (IST) programs. For the first time, this revision added dynamic restraint (snubber) inservice examination and testing program guidelines along with pump and valve IST programs. This paper highlights important changes to NUREG-1482, but is not intended to provide a complete record of all changes to the document. The NRC intends to continue to develop and improve its guidance on IST methods through active participation in the ASME OM Code consensus process, interactions with various technical organizations, user groups, and through periodic updates of NRC-published guidance and issuance of generic communications as the need arises. Revision 2 to NUREG-1482 incorporates regulatory guidance applicable to the 2004 Edition including 2005 and 2006 Addenda to the ASME OM Code. Revision 0 and Revision 1 to NUREG-1482 are still valid and may continue to be used by those licensees who have not been required to update their IST program to the 2004 Edition including the 2005 and 2006 Addenda (or later Edition) of the ASME OM Code. The guidance provided in many sections herein may be used for requesting relief from or alternatives to ASME OM Code requirements. However, licensees may also request relief or authorization of an alternative that is not in conformance with the guidance. In evaluating such requested relief or alternatives, the NRC uses the guidelines/recommendations of the NUREG, where applicable. The guidelines and recommendations provided in this NUREG and its Appendix A do not supersede the regulatory requirements specified in Title 10 of the Code of Federal Regulations (10 CFR) 10 CFR 50.55a, “Codes and standards”. Further, this NUREG does not authorize the use of alternatives to, grant relief from, the ASME OM Code requirements for inservice testing of pumps and valves, or inservice examination and testing of dynamic restraints (snubbers), incorporated by reference in 10 CFR 50.55a. Paper published with permission.

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.

Research on Earthquake Acceleration Alarm of Nuclear Power Plant

2021 ◽  
Author(s):  
Li Liang ◽  
Pan Rong ◽  
Ren Guopeng ◽  
Zhu Xiuyun
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
False Alarm ◽  
Ground Motion ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Time History ◽  
Seismic Acceleration ◽  
Trigger Signal

Abstract Almost all nuclear power plants in the world are equipped with seismic instrument system, especially the third generation nuclear power plants in China. When the ground motion measured by four time history accelerometers of containment foundation exceeds the preset threshold, the automatic shutdown trigger signal will be generated. However, from the seismic acceleration characteristics, isolated and prominent single high frequency will be generated the acceleration peak, which has no decisive effect on the seismic response, may cause false alarm, which has a certain impact on the smooth operation of nuclear power plant. According to the principle of three elements of ground motion, this paper puts forward a method that first selects the filtering frequency band which accords with the structural characteristics of nuclear power plants, then synthesizes the three axial acceleration time history, and finally selects the appropriate acceleration peak value for threshold alarm. The results show that the seismic acceleration results obtained by this method can well represent the actual magnitude of acceleration, and can solve the problem of false alarm due to the randomness of single seismic wave, and can be used for automatic reactor shutdown trigger signal of seismic acceleration.

scholarly journals Comparison and Evaluation of Domestic and Foreign Radiation Environmental Standards for Nuclear Power Plants

2021 ◽  
Vol 2083 (2) ◽  
pp. 022020
Author(s):  
Jiahuan Yu ◽  
Xiaofeng Zhang
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
The United States ◽  
Discharge System ◽  
Radioactive Effluent ◽  
Effluent Discharge ◽  
The Impact

Abstract With the development of the nuclear energy industry and the increasing demand for environmental protection, the impact of nuclear power plant radiation on the environment has gradually entered the public view. This article combs the nuclear power plant radiation environmental management systems of several countries, takes the domestic and foreign management of radioactive effluent discharge from nuclear power plants as a starting point, analyses and compares the laws and standards related to radioactive effluents from nuclear power plants in France, the United States, China, and South Korea. In this paper, the management improvement of radioactive effluent discharge system of Chinese nuclear power plants has been discussed.

scholarly journals Seismic safety evaluation during site selection for the nuclear power plants in Bangladesh

2018 ◽  
Vol 4 (4) ◽  
pp. 251-256 ◽  
Author(s):  
Sergey Shcheklein ◽  
Ismail Hossain ◽  
Mohammad Akbar ◽  
Vladimir Velkin
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Spent Nuclear Fuel ◽  
Heat Removal ◽  
Good Time ◽  
Diagnostics System ◽  
Seismic Impacts

Bangladesh lies in a tectonically active zone. Earlier geological studies show that Bangladesh and its adjoining areas are exposed to a threat of severe earthquakes. Earthquakes may have disastrous consequences for a densely populated country. This dictates the need for a detailed analysis of the situation prior to the construction of nuclear power plant as required by the IAEA standards. This study reveals the correlation between seismic acceleration and potential damage. Procedures are presented for investigating the seismic hazard within the future NPP construction area. It has been shown that the obtained values of the earthquake’s peak ground acceleration are at the level below the design basis earthquake (DBE) level and will not lead to nuclear power plant malfunctions. For the most severe among the recorded and closely located earthquake centers (Madhupur) the intensity of seismic impacts on the nuclear power plant site does not exceed eight points on the MSK-64 scale. The existing predictions as to the possibility of a super-earthquake with magnitude in excess of nine points on the Richter scale to take place on the territory of the country indicate the necessity to develop an additional efficient seismic diagnostics system and to switch nuclear power plants in good time to passive heat removal mode as stipulated by the WWER 3+ design. A conclusion is made that accounting for the predicted seismic impacts in excess of the historically recorded levels should be achieved by the establishment of an additional efficient seismic diagnostics system and by timely switching the nuclear power plants to passive heat removal mode with reliable isolation of the reactor core and spent nuclear fuel pools.

Comparison of Plant Response of Forsmark 3 Nuclear Power Plant to Forsmark 1 Type of Transient (25th July 2006), Using the HAMBO BWR Simulator: Thermal-Hydraulic Aspects

2008 ◽  
Author(s):  
Esko Pekkarinen
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Short Circuit ◽  
Complex Interplay ◽  
Operator Interface ◽  
Control Rooms ◽  
Hydraulic Process

Modernisation of control rooms of the nuclear power plants has been a major issue during the last few years. With this as a basis, the BWR plants in Sweden and Finland funded, in co-operation with the Halden Project, an experimental HAMBO BWR simulator project based on the Forsmark 3 plant in Sweden. VTT Energy in Finland developed the simulator models for HAMBO with the aid of their APROS tool, while the operator interface was developed by the Halden Project. The simulator and its performance have been described in other publications [1, 2]. On July 25th 2006 there was a short circuit at Forsmark 1 nuclear power plant when manoeuvring equipment in the 400kV-switch yard. Due to the short circuit, the plant suffered an electrical disturbance that led to scram and failure of two out of four diesel generators. The purpose of the study carried out at VTT in 2007 was to assess the capabilities of the HAMBO BWR simulator to handle Forsmark 1 type of events in different nuclear power plants (Forsmark 3 in this case). The Forsmark 1 incident showed (among other things) that the intention to protect certain components (in this case the UPS-system) can in certain situations affect negatively to the safety functions. It is concluded that most of this type of BWR transients may be simulated to a certain extent using the existing HAMBO- and APROS- models. A detailed modelling of the automation and electric systems is required sometimes if the complex interplay between these systems and the process is to be predicted reliably. The modelling should be plant specific and level of detail should be assessed case-by-case (i.e. what kind of transient is in question, what systems are available, what is the main purpose of the analyses etc.). The thermal-hydraulic models of the APROS-code seem to replicate well the real behaviour of thermal-hydraulic process provided that there is enough information about the transient in consideration.

scholarly journals “True Believers” or Numerical Terrorism at the Nuclear Power Plant

2011 ◽  
Vol 231 (5-6) ◽  
Author(s):  
Walter Krämer ◽  
Gerhard Arminger
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Childhood Cancer ◽  
Political Party ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
The Political ◽  
Heated Debate ◽  
Data Points

SummaryFor decades, there has been a heated debate about whether or not nuclear power plants contribute to childhood cancer in their respective neighbourhoods, with statisticians testifying on both sides. The present paper points to some flaws in the pro-arguments, taking a recent study prepared for the political party “Bündnis 90 /Grüne” as a specimen. Typical mistakes include an understatement of the size of tests of significance, disregard of important covariates and extreme reliance on very few selected data points.

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