Review of the Applicability of Nuclear Regulatory Commission Information Notice (IN) 2012-14, “Motor Operated Valve Inoperable Because of Stem-Disc Separation,” to Constellation Energy Nuclear Group’s Ginna Station

2014 ◽  
Author(s):  
David Garofoli ◽  
Gregg Joss
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Operating Experience ◽  
Nuclear Regulatory Commission ◽  
Corrective Action ◽  
Level Of Confidence ◽  
Valve Stem ◽  
The Subject ◽  
Regulatory Commission

U.S. Nuclear Regulatory Commission (NRC) Information Notice (IN) 2012-14, “Motor-Operated Valve Inoperable Because of Stem-Disc Separation”, was issued to inform nuclear power-plant licensees of recent operating experience involving a motor-operated valve (MOV) that failed at the connection between the valve stem and disc. The NRC expectation was that recipients would review the information for applicability to their facilities and consider actions, as appropriate, to avoid similar problems. Additional regulatory suggestions and insights contained in the IN are not NRC requirements. On closer examination of the events involved, it became apparent that the undetected stem-disc separation observed with the subject MOV was not necessarily limited to that type or style of valve. In fact, the vast majority of inservice testing (IST) valves, and the manner in which they are tested, could also be susceptible to loss of functionality going undetected. The intent of the compliance project performed at the R.E. Ginna Station nuclear power plant was to examine the current testing performed on each IST program valve and determine the level of confidence that stem-disc separation would be detected. If the level of confidence was deemed less than acceptable for a subject valve, one or more augmented actions, as deemed both practicable and viable, were recommended for implementation. The purpose of this presentation paper is to describe the systematic methodology that was employed to validate the effectiveness of the current periodic IST valve testing conducted at the R.E. Ginna Station and the corrective-action recommendations that were made as deemed appropriate. The corrective action(s) were designed to preclude the occurrence of future stem-disc separation issues going undetected, which could result in the loss of valve and potentially the loss of the associated accident-mitigation system’s operational readiness condition. Paper published with permission.

scholarly journals Small Modular Reactor Control Room Workstation Demonstration

2019 ◽  
Vol 63 (1) ◽  
pp. 479-480
Author(s):  
Jessica Stevens ◽  
Kevin LaFerriere ◽  
Ryan Flamand NuScale
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Operating Experience ◽  
Nuclear Regulatory Commission ◽  
Lessons Learned ◽  
Agile Development ◽  
Control Room ◽  
Small Modular Reactor ◽  
Regulatory Commission

A control room simulator was designed to model the operation of a NuScale small modular reactor (SMR) nuclear power plant and provide enough fidelity to perform staffing validation studies for Nuscale’s Nuclear Regulatory Commission Design Certification Application. The simulator serves as a simulated control room with work stations to mimic the operation of an SMR module, turbine generator, and support systems using a proprietary human system interface (HSI) software package. The simulator, which includes all HSI screens, was designed by a team of Human Factors and Plant Operations staff to capitalize on best practices, lessons learned, and operating experience using the Agile development process. Finally, the design process included the development of plant operating procedures and training material as well as a training platform for future plant operators at an SMR nuclear power plant.

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.

Managing nuclear power plant induced disasters

2015 ◽  
Vol 13 (5) ◽  
pp. 417 ◽  
Author(s):  
Dean Kyne, PhD, MPA, MPS
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Disaster Management ◽  
Nuclear Power ◽  
Source Term ◽  
Computer Software ◽  
Nuclear Regulatory Commission ◽  
Assessment System ◽  
Management Process ◽  
Regulatory Commission

Objective: To understand the management process of nuclear power plant (NPP) induced disasters. The study shields light on phases and issues associated with the NPP induced disaster management. Setting: This study uses Palo Verde Nuclear Generation Station as study subject and Arizona State as study area.Design: This study uses the Radiological Assessment System for Consequence Analysis (RASCAL) Source Term to Dose (STDose) of the Nuclear Regulatory Commission, a computer software to project and assess the source term dose and release pathway. This study also uses ArcGIS, a geographic information system to analyze geospatial data. A detailed case study of Palo Verde Nuclear Power Generation (PVNPG) Plant was conducted.Results: The findings reveal that the NPP induced disaster management process is conducted by various stakeholders. To save lives and to minimize the impacts, it is vital to relate planning and process of the disaster management.Conclusions: Number of people who expose to the radioactive plume pathway and level of radioactivity could vary depending on the speed and direction of wind on the day the event takes place. This study findings show that there is a need to address the burning issue of different racial and ethnic groups’ unequal exposure and unequal protection to potential risks associated with the NPPs.

Various MTC Studies of TRACE with RCS Pressure Predictions under ATWS for Maanshan

2013 ◽  
Vol 284-287 ◽  
pp. 1151-1155
Author(s):  
Che Hao Chen ◽  
Jong Rong Wang ◽  
Hao Tzu Lin ◽  
Chun Kuan Shih
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Thermal Power ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Graphic User Interface ◽  
Pressurized Water ◽  
Plant Safety ◽  
Regulatory Commission

The objective of this study is to utilize TRACE (TRAC/RELAP Advanced Computational Engine) code to analyze the reactor coolant system (RCS) pressure transients under ATWS (Anticipated Transient Without Scram) for Maanshan PWR (Pressurized Water Reactor) in various MTC (Moderator Temperature Coefficient) conditions. TRACE is an advanced thermal hydraulic code for nuclear power plant safety analysis, which is currently under development by the United States Nuclear Regulatory Commission (USNRC). A graphic user interface program named SNAP (Symbolic Nuclear Analysis Package), which processes inputs and outputs for TRACE is also under development. Maanshan nuclear power plant (NPP) is the only Westinghouse PWR in Taiwan. The rated core thermal power of Maanshan with MUR (Measurement Uncertainty Recapture) is 2822 MWt. In document SECY-83-293, all initializing events were classified as either turbine trip or non-turbine trip events and their ATWS risks were also evaluated according to these two events. Loss of condenser vacuum (LOCV) and Loss of normal feedwater (LONF) ATWS were identified as limiting transients of turbine trip and non-turbine trip events in this study. According to ASME Code Level C service limit criteria, the RCS pressure for Maanshan NPP must be under 22.06 MPa. Furthermore, we select the LOCV transient to analyze various MTC effects on RCS pressure variations.

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.

scholarly journals Demonstrating Structural Adequacy of Nuclear Power Plant Containment Structures for Beyond Design-Basis Pressure Loadings

2010 ◽  
Author(s):  
Joseph Braverman ◽  
Richard Morante ◽  
Charles Hofmayer ◽  
Robert Roche-Rivera ◽  
Jose Pires
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Nuclear Regulatory Commission ◽  
Gas Generation ◽  
Design Basis ◽  
And Performance ◽  
Regulatory Commission ◽  
Technical Basis

Demonstrating the structural integrity of U.S. nuclear power plant (NPP) containment structures, for beyond design-basis internal pressure loadings, is necessary to satisfy Nuclear Regulatory Commission (NRC) requirements and performance goals. This paper discusses methods for demonstrating the structural adequacy of the containment for beyond design-basis pressure loadings. Three distinct evaluations are addressed: (1) estimating the ultimate pressure capacity of the containment structure (10 CFR 50 [1] and US NRC Standard Review Plan, Section 3.8) [2]; (2) demonstrating the structural adequacy of the containment subjected to pressure loadings associated with combustible gas generation (10 CFR 52 [3] and 10 CFR 50 [1]); and (3) demonstrating the containment structural integrity for severe accidents (10 CFR 52 [3] as well as SECY 90–016 [4], SECY 93–087 [5], and related NRC staff requirements memoranda (SRMs)). The paper describes the technical basis for specific aspects of the methods presented. It also presents examples of past issues identified in licensing activities related to these evaluations.

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