Comprehensive Guidance for the Evaluation of Human-Systems Interfaces in Complex Systems

1995 ◽  
Vol 39 (18) ◽  
pp. 1160-1164 ◽  
Author(s):  
John O'Hara ◽  
William Stubler ◽  
William Brown ◽  
Jerry Wachtel ◽  
J. Persensky
Keyword(s):  
Nuclear Power ◽  
Nuclear Regulatory Commission ◽  
Evaluation Methodology ◽  
Plant Safety ◽  
Nuclear Plants ◽  
Machine Systems ◽  
Regulatory Commission ◽  
Plant Evaluation ◽  
The U.S ◽  
Guidance Development

Advanced human-system interface (HSI) technologies are being developed in the commercial nuclear power industry. These HSIs may have significant implications for plant safety in that they will affect the ways in which the operator interacts with and supervises an increasingly complex system. The U.S. Nuclear Regulatory Commission (NRC) reviews the HSI aspects of nuclear plants to ensure that operator performance and reliability are supported. The NRC is developing guidance to support its review of these advanced designs. The guidance consists of an evaluation methodology and an extensive set of human factors guidelines which are used in one aspect of the evaluation. The paper describes the guidance development of the evaluation methodology and the guidelines. While originally developed for nuclear plant evaluation, the methodology is applicable to other types of complex human-machine systems as well.

Evaluation of Complex Human-Machine Systems Using HFE Guidelines

1994 ◽  
Vol 38 (16) ◽  
pp. 1008-1012 ◽  
Author(s):  
John M. O'Hara
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Health And Safety ◽  
Nuclear Regulatory Commission ◽  
Human Factors Engineering ◽  
Plant Safety ◽  
Plant Personnel ◽  
Machine Systems ◽  
Regulatory Commission ◽  
Development And Validation

The purpose of this paper is to discuss the role of human factors engineering (HFE) guidelines in the evaluation of complex human-machine systems, such as advanced nuclear power plants. Advanced control rooms will utilize human-system interface (HSI) technologies that can have significant implications for plant safety in that they will affect the ways in which plant personnel interact with the system. In order to protect public health and safety, the U.S. Nuclear Regulatory Commission reviews the HFE aspects of plant HSIs to ensure that they are designed to HFE principles and that operator performance and reliability are appropriately supported. Evaluations using HFE guidelines are an important part of the overall review methodology. The Advanced HSI Design Review Guideline (DRG) was developed to provide these review criteria. This paper will address (1) the issues associated with guideline-based evaluations, (2) DRG development and validation, and (3) the DRG review procedures.

Foundation Engineering for Nuclear Plants: U.S. vs. European Code and Practice

2009 ◽  
Author(s):  
Nasser Massoudi
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Regulatory Commission ◽  
Foundation Engineering ◽  
Regulatory Requirements ◽  
Nuclear Plants ◽  
Geotechnical Investigations ◽  
European Code ◽  
Regulatory Commission ◽  
The U.S

This paper reviews the current regulatory and industry practices in geotechnical investigations for nuclear power plants in the U.S. and Europe, with the intent to highlight the common features and underscore the differences. Specifically, applicable sections of regulatory and industry-established codes and practices are reviewed as relate to geotechnical practices and foundation engineering. Similarly, regulatory requirements such as those established by the U.S. Nuclear Regulatory Commission and the European equivalents will be reviewed. The paper serves as a vehicle to highlight industry and regulatory common grounds, as well as variations in the two practices, in the spirit of disseminating knowledge on codes and standards and facilitating international cooperation between the foundation engineering community in the U.S. and Europe.

Human Factor Data in Nuclear Power Plant Applications

1980 ◽  
Vol 24 (1) ◽  
pp. 123-123
Author(s):  
Linda O. Hecht
Keyword(s):  
Nuclear Power ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Reliability Modeling ◽  
Plant Safety ◽  
Modeling Techniques ◽  
Nuclear Systems ◽  
Factor Data ◽  
Regulatory Commission ◽  
The Impact

Due to the concern for safety the nuclear power industry in the United States has fostered the use of reliability analysis to assess system performance and the impact of system failure on overall plant safety. The need for system and component failure rate data has been recognized and has spurred such efforts as NPRDS (Nuclear Power Research Data System) and IEEE's Std 500 (The Reliability Data Manual). Reliability modeling techniques have been developed for application to nuclear systems and are presently being considered by the Nuclear Regulatory Commission for licensing purposes.

TITRAM Fast Transient Licensing Analysis Methodology for Kuosheng BWR/6

2010 ◽  
Author(s):  
Jan-Ru Tang ◽  
Hon-Chin Jien ◽  
Yang-Kai Chiu ◽  
Cheng-Der Wang ◽  
Julian S. C. Chian
Keyword(s):  
Nuclear Power ◽  
Transient Analysis ◽  
Nuclear Regulatory Commission ◽  
Nuclear Industry ◽  
Evaluation Methodology ◽  
Uncertainty Evaluation ◽  
Power Company ◽  
Fast Transients ◽  
Fast Transient ◽  
Regulatory Commission

This paper presents the TITRAM (TPC/INER Transient Analysis Method) methodology for the fast transient analysis of Kuosheng Nuclear Power Station (KSNPS) with two units of General Electric (GE) designed BWR/6 (Boiling Water Reactor). The purpose of this work is to provide a technical basis of Taiwan Power Company (TPC)/Institute of Nuclear Energy Research (INER)’s qualification to perform plant specific licensing safety analyses for the Final Safety Analysis Report (FSAR) basis system fast transients, and related plant operational transient analyses for the Kuosheng plant. The major task of qualifying TITRAM as a licensing method for BWR transient analysis is to adequately quantify its analysis uncertainty. A similar approach as the CSAU (Code Scaling, Applicability, and Uncertainty Evaluation) methodology developed by the USNRC (United States Nuclear Regulatory Commission) was adopted. The CSAU methodology could be characterized as three significant processes, namely code applicability, transient scenario specification and uncertainty evaluation based on Phenomena Identification and Ranking. The applicability of the TITRAM code package primarily using the SIMULATE-3 and RETRAN-3D codes are demonstrated with analyses of integral plant tests such as Peach Bottom Turbine Trip Test and plant startup tests of KSNPS. A Phenomena Identification and Ranking Table (PIRT) with uncertainty values for each identified parameter to cover 95% of possible values are established for the selected KSNPS fast transients. The experience from BWR organizations in the nuclear industry is used as a guide in construction of the PIRT. Sensitivity studies and associated statistical analyses are performed to determine the overall uncertainty of fast transient analysis with TITRAM based on the KSNPS Analysis Nominal Model. Finally, the Licensing Model is established for future licensing applications.

scholarly journals Quality assurance program for Angra 1 licence renewal and long-term operation

2021 ◽  
Vol 9 (2B) ◽  
Author(s):  
YOUSSEF MORGHI ◽  
Amir Zacarias Mesquita ◽  
Ana Rosa BALIZA MAIA
Keyword(s):  
Quality Assurance ◽  
Nuclear Power ◽  
Nuclear Regulatory Commission ◽  
Quality Assurance Program ◽  
Term Operation ◽  
Safety Requirements ◽  
Plant Safety ◽  
Regulatory Commission ◽  
License Renewal

In Brazil, according to Cnen standard, a nuclear power plant has authorization to operate for 40 years. Angra 1 commercial operation started in 1985 and it has license to operate until 2024. Eletronuclear aims to extend the operation of the Angra 1 plant from 40 to 60 years. To obtain the license renewal by more than 20 years (long-term operation), Eletronuclear will need to meet the requirements of 10 CFR Part 54, Cnen NT-CGRC-007/18 and NT-CGRC-008/18 (Cnen technical notes). To obtain a license renewal to a long-term operation it is necessary to demonstrate that the plants will operate according to safety requirements, through analysis, testing, aging management, system upgrades, as well as additional inspections. Plant operators and regulators must always ensure that plant safety is maintained and, when it is possible, strengthened during the long-term operation of the plant. One of the documents to obtain a license renewal to a long-term operation is the Quality Assurance Program (QAP). Angra 1 has a QAP according to 10CFR 50 App B and Cnen NN 1.16 for safety related items. However, according to 10 CFR50.34, Nureg-1800 Appendix A.2, Nureg-1801 Appendix A-1 of Nuclear Regulatory Commission (NRC) and NT-CGRC-007/18 and NT-CGRC-008/18 of Cnen, the QAP needs to include the items that are not safety related but are included in the Aging Management. This article will discuss the Angra 1 QAP for the license renewal to a long-term operation according the standards approved by Cnen.

Lessons Learned From the Application and Refinement of Risk-Informed Inservice Inspection Programs

2008 ◽  
Author(s):  
Alan D. Chockie ◽  
M. Robin Graybeal ◽  
Scott D. Kulat
Keyword(s):  
Cost Effective ◽  
Nuclear Regulatory Commission ◽  
Lessons Learned ◽  
Nuclear Industry ◽  
Plant Safety ◽  
Nuclear Plants ◽  
New Information ◽  
Effective Manner ◽  
Regulatory Commission ◽  
Available Information

The risk-informed inservice inspection (RI-ISI) process provides a structured and systematic framework for allocating inspection resources in a cost-effective manner while improving plant safety. It helps focus inspections where failure mechanisms are likely to be and where enhanced inspections are warranted. To date, over eighty-five percent of US nuclear plants and a number of non-US plants have implemented, or are in the process of implementing, RI-ISI programs. Many are already involved in the periodic update of their RI-ISI program. The development of RI-ISI methodologies in the US has been a long and involved process. The risk-informed procedures and rules were developed to take full advantage of PRA data, industry and plant experiences, information on specific damage mechanisms, and other available information. An important feature of the risk-informed methodologies is the requirement to make modifications and improvements to the plant’s RI-ISI application as new information and insights become available. The nuclear industry, ASME Section XI, and the Nuclear Regulatory Commission have all worked together to take advantage of the lessons learned over the years to refine and expand the use of risk-informed methodologies. This paper examines the lessons learned and the benefits received from the application and refinement of risk-informed inservice inspection programs. Also included in the paper is a review of how the information and insights have been used to improve the risk-informed methodologies.

Long Term Operation of Nuclear Power Plants: Status in the U.S.

2013 ◽  
Author(s):  
Garry G. Young
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Regulatory Commission ◽  
Safe Operation ◽  
Term Operation ◽  
The Status ◽  
Regulatory Commission ◽  
License Renewal ◽  
The U.S

As of January 2013, the U.S. Nuclear Regulatory Commission (NRC) has renewed the operating licenses of 73 nuclear units out of a total of 104 licensed units, allowing for up to 60 years of safe operation. In addition, the NRC has license renewal applications under review for 15 units and more than 13 additional units have announced plans to submit applications over the next few years [1]. This brings the total of renewed licenses and plans for renewal to over 97% of the 104 operating nuclear units in the U.S. This paper presents the status of the U.S. license renewal process and issues being raised for possible applications for subsequent renewals for up to 80 years of operation. By the end of 2013 there will be 26 nuclear plants in the U.S. (or 25% of the 104 units) that will be eligible to seek a second license renewal and by the end of 2016 this number will increase to about 50% of the 104 licensed units. Although some nuclear plant owners have announced plans to shutdown before reaching 60 years, the majority are keeping the option open for long term operation beyond 60 years. The factors that impact decisions for both the first license renewals and subsequent renewals for 80 years of safe operation are presented and discussed in this paper.

Human Reliability Data Bank: Pilot Implementation

1983 ◽  
Vol 27 (2) ◽  
pp. 175-179
Author(s):  
Kay Comer ◽  
Dwight P. Miller
Keyword(s):  
Nuclear Power ◽  
Data Bank ◽  
Nuclear Regulatory Commission ◽  
Human Reliability ◽  
Reliability Data ◽  
Three Phase ◽  
Pilot Implementation ◽  
Feasibility Test ◽  
Regulatory Commission ◽  
The U.S

The U.S. Nuclear Regulatory Commission and Sandia National Laboratories have initiated a three-phase research program to develop a plan for a human reliability data bank. This research is in response to the data needs of the nuclear power industry's probabilistic risk assessment community. The three phases are: A - Develop the data bank concept, B - Develop an implementation plan and conduct a feasibility test, and C - Assist sponsor in implementing the data bank. This paper describes the results of work performed during Phase A and the program tasks scheduled for Phase B.

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.

Potential Ways to Modify the NRC Safety Goal Policy

2009 ◽  
Author(s):  
Steven A. Arndt ◽  
Richard Denning
Keyword(s):  
Nuclear Power ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Policy Statement ◽  
The Past ◽  
Core Damage ◽  
Comparative Risk ◽  
Regulatory Commission ◽  
Health Objectives ◽  
The U.S

There have been significant discussions over the past few years by the U.S. Nuclear Regulatory Commission (NRC) staff and the Advisory Committee on Reactor Safeguards (ACRS), as to the adequacy of the safety goals for use with the next generation of nuclear power reactors to be built in the United States. The NRC, in its safety goals policy statement, has provided general qualitative safety goals and basic quantitative health objectives (QHOs) for nuclear reactors in the U.S. Risk metrics such as core damage frequency (CDF) and large early release frequency (LERF) have also been used as surrogates for the QHOs. This study will examine a potential approach to update the safety goals that includes the establishment of new qualitative goals associated with the comparative risk of generating electricity by viable competing technologies, and the development of preliminary tests in support of a new qualitative goal.

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