scholarly journals Human Factors Guidance for Control Room Evaluation

2000 ◽  
Vol 44 (22) ◽  
pp. 519-522 ◽  
Author(s):  
John O'Hara ◽  
William Brown ◽  
William Stubler ◽  
James Higgins ◽  
Jerry Wachtel ◽  
...  
Keyword(s):  
Human Factors ◽  
Interface Design ◽  
Nuclear Regulatory Commission ◽  
Nuclear Industry ◽  
Human Factors Engineering ◽  
Evaluation Tool ◽  
Design Review ◽  
Human System ◽  
Software Application ◽  
Regulatory Commission

The Human-System Interface Design Review Guideline (NUREG-0700, Revision 1) was developed by the U.S. Nuclear Regulatory Commission (NRC) to provide human factors guidance as a basis for the review of advanced human-system interface technologies. The guidance consists of three components: design review procedures, human factors engineering guidelines, and a software application to provide design review support called the “Design Review Guideline.” Since it was published in June 1996, Rev. 1 to NUREG-0700 has been used successfully by NRC staff, contractors and nuclear industry organizations, as well as by interested organizations outside the nuclear industry. The NRC has committed to the periodic update and improvement of the guidance to ensure that it remains a state-of-the-art design evaluation tool in the face of emerging and rapidly changing technology. This paper addresses the current research to update of NUREG-0700 based on the substantial work that has taken place since the publication of Revision 1.

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.

Putting Gonuke Into Practice: Considerations for Human Factors Evaluations

2021 ◽  
Vol 65 (1) ◽  
pp. 618-622
Author(s):  
Ronald L. Boring ◽  
Thomas A. Ulrich ◽  
Roger Lew
Keyword(s):  
Human Factors ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Regulatory Commission ◽  
Human Factors Engineering ◽  
Phases Of Development ◽  
Review Model ◽  
Regulatory Commission ◽  
System Interfaces

The Guideline for Operational Nuclear Usability and Knowledge Elicitation (GONUKE) framework was introduced in 2015 to support human factors evaluations needed for control room upgrades at nuclear power plants. NUREG-0711, the Human Factors Engineering Program Review Model, is used by the U.S. Nuclear Regulatory Commission to review human factors activities associated with human-system interfaces at nuclear power plants, and GONUKE is anchored to the phases of development and design in NUREG-0711. This paper addresses five considerations to help users of GONUKE better apply the framework to evaluations for NUREG-0711 and beyond. These five considerations are: (1) GONUKE only specifies evaluation, not design; (2) GONUKE is a framework, not a method or process; (3) GONUKE goes beyond NUREG-0711 requirements; (4) GONUKE application shouldfollow a graded approach; (5) different evaluations are required fo r formative vs. summative phases.

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.

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.

Derivation of the New Pressurized Thermal Shock Screening Criteria

2011 ◽  
Author(s):  
Terry Dickson ◽  
Shengjun Yin ◽  
Mark Kirk ◽  
Hsuing-Wei Chou
Keyword(s):  
United States ◽  
Fracture Mechanics ◽  
Thermal Shock ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Nuclear Industry ◽  
Screening Criteria ◽  
Pressurized Thermal Shock ◽  
Regulatory Commission ◽  
Technical Basis

As a result of a multi-year, multi-disciplinary effort on the part of the United States Nuclear Regulatory Commission (USNRC), its contractors, and the nuclear industry, a technical basis has been established to support a risk-informed revision to pressurized thermal shock (PTS) regulations originally promulgated in the mid-1980s. The revised regulations provide alternative (optional) reference-temperature (RT)-based screening criteria, which is codified in 10 CFR 50.61(a). How the revised screening criteria were determined from the results of the probabilistic fracture mechanics (PFM) analyses will be discussed in this paper.

Sign in / Sign up

Export Citation Format

Share Document