Human engineering research to support regulatory guidelines for nuclear power plant application

1982 ◽  
Author(s):  
James P. Jenkins
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Human Engineering ◽  
Engineering Research ◽  
Regulatory Guidelines

Survey and Assessment of Nuclear Power Plant Control Room Operability

1980 ◽  
Vol 24 (1) ◽  
pp. 267-270
Author(s):  
Clifford C. Baker ◽  
Robin West ◽  
Kenneth M. Mallory
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Data Collection ◽  
Nuclear Power ◽  
Human Engineering ◽  
Control Room ◽  
Design Problems ◽  
Power Plant Control ◽  
Control Rooms ◽  
Plant Control

As part of an effort to develop human engineerging guidelines and a methodology for the evaluation of nuclear power plant control room operability, the Essex Corporation conducted T & E (test and evaluation) reviews of a wide sample of nuclear power plant control rooms. The objectives of these design reviews were: 1) selection, application, and development of human engineering evaluation guidelines applicable to the nuclear power industry; 2) selection and development of data collection and analysis procedures; and 3) identification of recurrent human engineering design problems in the control rooms of currently operating nuclear power plants. The present paper discusses the approach taken and the findings in item three above. Thirteen control rooms were visited, and guidelines and data collection methods under various degrees of development were applied. Following control room visits, data were analyzed according to usability, number of incidences of similar or identical operability design problems, criticality of problems with respect to both public and plant safety, and subjective assessment of operational affects due to human engineering problems in design. Results to date show that the following areas have recurrent operability design problems: layout of controls and displays according to either operational or functional use; coding of information for visual and auditory presentation; job performance aid and procedures design; communications; environmental factors such as ambient noise; violations in control and display conventions employed; use of conventions which violate population stereotypes; and failure to design within anthropometric constraints. Further work is being conducted by Essex Corporation to identify critical human engineering deficiencies in control room design and to select adequate yet cost-effective and corrective backfits.

Procedures for Human Engineering Evaluation of Nuclear Power Plant Control Rooms

1980 ◽  
Vol 24 (1) ◽  
pp. 271-275
Author(s):  
Kenneth M. Mailory ◽  
Clifford C. Baker ◽  
Robin K. West
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Human Engineering ◽  
Control Room ◽  
Room Design ◽  
Engineering Problems ◽  
Power Plant Control ◽  
Control Rooms ◽  
Plant Control

Few human engineering standards or criteria for the design of nuclear power plant control rooms existed prior to the accident at Three Mile Island — Unit 2. For the most part control room design was dictated by electrical criteria, costs, and, most importantly, by precedent evolved from fossil fuel plant experience. Since the TMI-2 accident, the Nuclear Regulatory Commission has undertaken an ambitious program to develop control room design and operational guidelines to be used by utilities in evaluating the human engineering fitness of control rooms and in identifying human engineering problems requiring backfit. The following paper reviews the method used to develop control room guidelines, the process suggested to the utilities for performing control room evaluations, and sources for and the content of guidelines. As reported in the paper, evaluation guidelines evolved from a basic set of military standards and checklists through a series of on-site control room reviews. The methods used in these reviews involve surveys, checklists, and videotaped walk-throughs of emergency procedures. The final product is a Guidebook containing: (a) procedures for scheduling, planning, administration, and staffing of human engineering reviews; (b) the evaluation procedures to be used, including guidelines, human engineering data, references, and methods; (c) a trade-off process for sorting out problems needing immediate vs. more remote attention; and (d) suggestions for backfits for the human engineering problems most widespread in the industry.

Physicochemical Properties of Vitrified Forms for LILW Generated From Korean Nuclear Power Plant

2015 ◽  
Vol 1744 ◽  
pp. 211-216 ◽  
Author(s):  
Cheon-Woo Kim ◽  
Hyehyun Lee ◽  
In-Sun Jang ◽  
Hyun-Jun Jo ◽  
Hyun-Je Cho
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Physicochemical Properties ◽  
Nuclear Power ◽  
Ion Exchange Resin ◽  
Immersion Test ◽  
Borosilicate Glasses ◽  
Irradiation Test ◽  
Regulatory Guidelines ◽  
Ph Buffer

ABSTRACTSince 1994, the KHNP has developed a vitrification technology to treat the LILW generated from Korean nuclear power plant. To vitrify the LILW including combustible Dry Active Waste (DAW) and Ion Exchange Resin (IER) containing Zeolite, two borosilicate glasses are formulated. One of the formulated glass, DG2, is for the DAW vitrification solely and the other one, AG8W1, is for the blended wastes (DAW & IER) vitrification in a commercial vitrification facility in HanUl (former Ulchin) nuclear power plant. The physicochemical properties of the two glasses have been evaluated. To evaluate the processability of the glasses, the viscosities and electrical conductivities of the glass melts were measured in the laboratory within a temperature range between 950 and 1,350 degrees C, respectively. The liquidus temperatures of the glasses were evaluated using a gradient furnace for DG2 and data from heat treatment for AG8W1. The Mössbauer spectroscopy for AG8W1 was employed to evaluate the relations between the redox equilibria of iron. In addition, to verify the waste acceptance criteria for the final disposal of the vitrified forms, the compressive strengths of the vitrified forms were tested after an immersion test, a thermal cycling test, and an irradiation test. To verify the chemical durability of the glasses, several tests such as PCT, ISO, VHT, Soxhlet, MCC-1, and ANS16.1 were carried out. The PCT showed leach rates of B, Na, Li and Si were much less than those of the benchmark glass. The ISO test was performed at 90 degrees C for 1,022 days and Cumulative Fraction Leached of all elements in the glasses were analyzed. According to the VHT, the glasses had an outstanding chemical resistance under humid environment at 200 degrees C for 7 days. The Soxhlet leaching was performed on rectangular glass samples at 98 degrees C for 30 days. To analyze the forward dissolution rates of major glass elements, the MCC-1 was conducted at temperatures of 40, 70, and 90 degrees C for three weeks in pH buffer solutions ranging from pH 4 to 11. The processability of the glasses was in the desired ranges. And the product quality of the glasses met all regulatory guidelines. Using two glasses, the CCIM commissioning tests in the UVF were successfully performed and they showed good workability.

scholarly journals Recommendations to the NRC on human engineering guidelines for nuclear power plant maintainability

1986 ◽  
Author(s):  
R.V. Badalamente ◽  
B.A. Fecht ◽  
D.E. Blahnik ◽  
J.D. Eklund ◽  
C.S. Hartley
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Human Engineering

The Application of Piping Material Code in the CAP1400 Nuclear Plant Project

2017 ◽  
Author(s):  
Shen Jie ◽  
Zhang Zhen-ning ◽  
Liu Yu
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Nuclear Industry ◽  
Nuclear Engineering ◽  
Code System ◽  
Design Institute ◽  
Engineering Research ◽  
Long Period ◽  
Information Coverage

Nuclear industry differs from most other industries in the characters of large scaling, long period and multiple collaborating institutions [1]. Traditionally, the material codes are compiled respectively by collaborating institutions of nuclear power plant projects, so the whole code system lacks unified management and planning, causing many defects such as the incompleteness in information coverage, the inconformity in classification and description of materials and the confusion in commodity codes. So, it is of great significance in setting up a standard nuclear power material code system to effectively enhance the efficiency of design and management, and to ensure the schedules of the projects. This article introduces in detail the entire process of the setting up and application of the material code system of State Nuclear Power Technology Company (SNPTC) by Shanghai Nuclear Engineering Research & Design Institute (SNERDI) for the CAP1400 nuclear power plant project. The application of the material code system in the CAP1400 project remarkably simplifies the work of design, material take-off and purchase, improving the project’s quality.

The Development and Application of the Engineering and Design Analyzer of CAP1400 Nuclear Power Plant

2017 ◽  
Author(s):  
He Yuanlei ◽  
Zhang Qijiang ◽  
Li Xiaoyan
Keyword(s):  
Control System ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Research And Development ◽  
Nuclear Power ◽  
Human Factor ◽  
Typical Application ◽  
Human Factor Engineering ◽  
Engineering Research ◽  
Acceptance Condition

In the process of research and development for a new nuclear power plant, it is very necessary to develop a dynamic platform and tools to analyze and verify the plant control & protect system and human factor engineering. Therefor, Shanghai Nuclear Engineering Research and Development Institute (SNERDI) developed the Engineering & Design Analyzer of CAP1400 Nuclear Power Plant (CAP1400 EDA) which provides a dynamic platform environment for analyzing and verifying the control system and human factor engineering of the CAP1400 nuclear power plant, a new Gen III passive nuclear power plant. In this paper, the mechanism and implement approach of the CAP1400 EDA will be mainly introduced, for example, the platform architecture of the EDA, analysis tools integrated in the EDA and CAP1400 nuclear power plant modes based on the EDA. In the meantime, a typical application case based on the CAP1400 EDA will be demonstrated in this paper, for example the capability of the NSSS control system will be verified in a ramp load down & raise operate transient. In this transient process, the NSSS control system of the plant is assessed whether it has the capability to keep the key parameter and state of the plant in an acceptance condition or range. And also other transients such as step load transient, large load transient can be simulated on CAP1400 EDA to verify whether or not the NSSS control systems are properly designed.

Seismic Margin Assessment and Improvements of Qinshan Nuclear Power Plant

2017 ◽  
Author(s):  
Leilei Xu ◽  
Xiaoling Zhao ◽  
Zhaohua Li
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Engineering ◽  
Margin Assessment ◽  
Seismic Safety ◽  
Engineering Research ◽  
Research And Design

Seismic Margin Assessment (SMA) is one of the methods for seismic safety assessment of nuclear power plants. The United States in the 1990s requires all running nuclear power plants to carry out an Individual Plant Examination of External Events (IPEEE), in the completion of the IPEEE of 110 units, 65 units using the seismic margin assessment method. The commercial nuclear power industry of China started late, although the Shanghai Nuclear Engineering Research and Design Institute in the 1980s to carry out seismic safety evaluation of relevant research. After the Fukushima nuclear accident, the National Nuclear Safety Administration requires all nuclear power plants in operation to carry out Seismic Margin Assessment. In view of the above background, the Shanghai Nuclear Engineering Research and Design Institute and Qinshan Nuclear Power Co., Ltd. on the Qinshan nuclear power plant to carry out the seismic margin assessment. Through the assessment, some weak links were found in the Qinshan nuclear power plant in some aspects. The Qinshan nuclear power plant in the implementation of the targeted improvements, the power plant’s seismic capacity has been effectively improved.

A Human Engineering Approach to the Enhancement of Operational Systems

1982 ◽  
Vol 26 (8) ◽  
pp. 727-730
Author(s):  
Kenneth A. Schulz ◽  
Steven M. Pine
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Human Factors ◽  
Nuclear Power ◽  
Systematic Approach ◽  
Human Engineering ◽  
Power Plant Control ◽  
Operational Systems ◽  
Control Rooms ◽  
Plant Control

The practicing human factors engineer is often called upon to help in improving a system which has already been designed, built and placed into operation. The authors' organization has recently completed a major program to develop guidelines for enhancing nuclear power plant control rooms. On the basis of this and other experiences, an outline for the process of enhancing a system has been developed. This paper will discuss elements of this process which lead to a comprehensive and systematic approach to improvement.

Sign in / Sign up

Export Citation Format

Share Document