Study on NPP Reactivity Accident Operating Strategy Design Based on Function Analysis and Task Analysis Technology

2018 ◽  
Author(s):  
Yu Aimin ◽  
Xu Zhao ◽  
Du Yu ◽  
Sun Qian
Keyword(s):  
Task Analysis ◽  
Nuclear Power ◽  
Power Plants ◽  
Function Analysis ◽  
Operation Strategy ◽  
Operating Condition ◽  
Accident Condition ◽  
Strategy Design ◽  
Reactivity Accident ◽  
And Task

Nuclear Power Plants (NPP) have multiple levels of defense in depth hierarchy. The NPP accident condition operation strategy belongs to the 3rd level. It is used to supervise the operator to handle the NPP under accident operating condition. NPP accident condition operation strategy is an essential and difficult work in NPP design field, hence only few organizations are able to develop the accident operating strategies independently all over the world. In this paper, a systematic NPP accident operating condition strategy design methodology is raised based on function analysis and task analysis technology. Based on the methodology, a reactivity accident operation strategy is designed and proved to be reasonable through preliminary verification and validation work.

Advanced Gen III PWR NPP General Operation Strategy Design Based on Function Analysis and Task Analysis Technology

2017 ◽  
Author(s):  
Xu Zhao ◽  
Miao Zhuang
Keyword(s):  
Task Analysis ◽  
Function Analysis ◽  
Operation Strategy ◽  
Design Features ◽  
Design Work ◽  
Innovative Design ◽  
New Type ◽  
Strategy Design ◽  
General Operation ◽  
And Task

General operation means NPP start-up and shut down operation, and it is one of the most important NPP operation manners. It is necessary to design a general operation strategy and related procedures for those advanced Gen III PWR reactor because of their innovative design features. By applying function analysis and task analysis technology, and considering the innovative design features, this paper carries out a hypothetical advanced PWR (HAP) general operation strategy design work with a systematical method. It is indicated that this method is able to be used in the general operation strategy design of other new type advanced reactors efficiently and accurately as a common method.

Research on Electric and I&C Equipment Safety Function Classification of Nuclear Power Plant

2021 ◽  
Author(s):  
Wang Yuqi ◽  
Sun Qian
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Function Analysis ◽  
System Component ◽  
Design Work ◽  
Equipment Safety ◽  
Safety Function

Abstract Classification of System, Component and Structure (SSC) is the base as well as high level demand of nuclear power plant. Equipment classification including electric and Instrument and Control (I&C) equipment is the precondition of correct design regulation and standard. Safety function classification is key pass of electric and I&C equipment classification. This paper researches the method of nuclear power plant electric and I&C equipment safety function classification. Firstly from view of function, it explains the importance of function classification. Then function analysis and classification of equipment is implemented by design order. Lastly from view of accident analysis, function classification is validated, and a complete approach of function classification is formed. The purpose of this paper is the NPP electric and I&C equipment safety function classification as an example, to study and summarize the method of the electric and I&C equipment safety function classification, and to provide the basis for specific items design work according to design requirements. At the same time, a practical method is provided for other similar NPP electric and I&C equipment classification work. The electric and I&C equipment function classification of nuclear power plant satisfy the basic principles requirement of relative nuclear power rules and codes. It provides an important basis of equipment classification for next nuclear power plants.

TRACE and RELAP5 Codes for Beyond Design Accident Condition Simulation in the SPES3 Facility

2012 ◽  
Author(s):  
Roberta Ferri ◽  
Fulvio Mascari ◽  
Paride Meloni ◽  
Giuseppe Vella
Keyword(s):  
Experimental Data ◽  
Nuclear Power ◽  
Power Plants ◽  
Three Dimensional ◽  
Heat Removal ◽  
Nuclear Fission ◽  
Dynamic Coupling ◽  
Hydraulic Conditions ◽  
Loss Of Coolant ◽  
Accident Condition

Code validation on qualified experimental data is a fundamental issue in the design and safety analyses of nuclear power plants. The SPES3 facility is being built at the SIET laboratories for an integral type SMR simulation, in the frame of an R&D program on nuclear fission, funded by the Italian Ministry of Economic Development and led by ENEA. The facility, based on the IRIS reactor design, reproduces the primary, secondary and containment systems with 1:100 volume scale, full elevation and prototypical fluid and thermal-hydraulic conditions. It is suitable to test the plant response to design and beyond design accidents in order to verify the effectiveness of the primary and containment system dynamic coupling to cope with loss of coolant accidents. Full and complete nodalizations of SPES3 were developed for TRACE and RELAP5 codes in order to investigate the code response to the simulation of the same accidental transient. The DVI line DEG break was simulated in beyond design conditions, assuming the failure of all emergency heat removal systems and relying on PCC intervention for containment depressurization and decay heat removal. The comparison of the code simulation results, other than providing information on the system behavior, allowed to investigate specific phenomena evidenced by the codes, according to the related modeling approach of components with one and three-dimensional volumes. The TRACE and RELAP5 codes will be applied for further transient analyses and will be validated on SPES3 experimental data, once the facility will be available.

Application of the Saint Code in Modelling and Analyzing Some Human Tasks in Nuclear Power Plants

1981 ◽  
Vol 25 (1) ◽  
pp. 105-109
Author(s):  
A. Mohsen M. Metwally ◽  
Zeinab A. Sabri ◽  
S. Keith Adams ◽  
Abdo A. Husseiny
Keyword(s):  
Task Analysis ◽  
Nuclear Power ◽  
Power Plants ◽  
Human Error ◽  
Nuclear Power Plants ◽  
Human Performance ◽  
Human Engineering ◽  
Performance Shaping Factors ◽  
Power Plant Operation ◽  
Error Probabilities

Two techniques useful for the simulation and analysis of human performance in tasks involving nuclear power plant operation, maintenance and testing are evaluated. The SAINT and THERP techniques are compared with respect to their relevance to conducting task analysis, estimation of human error probabilities and accounting for performance shaping factors in nuclear power plants. The results show that the SAINT is more flexible and has promising features for human engineering studies of complex systems when compared to the static THERP technique currently used in nuclear safety analysis.

Task Analysis in Designing the Control Room of Nuclear Power Plant and Safety Reviewing

2017 ◽  
Author(s):  
Yan Feng ◽  
Liang Li ◽  
Jingbin Liu ◽  
Ning Qiao ◽  
Yinhui Guo
Keyword(s):  
Sequence Analysis ◽  
Interface Design ◽  
Task Analysis ◽  
Nuclear Power ◽  
Human Performance ◽  
Hazard Identification ◽  
Information Control ◽  
Control Room ◽  
Task Support ◽  
And Task

Task analysis identifies the specific tasks needed to accomplish human actions, and the information, control and task support required to complete those tasks. In the control room design, the functions allocated to personnel shall be performed analysis, and it is identified that the specific task can be accomplished by the personnel. Task analysis can provide one of the bases for the human system interface design decisions; match human performance requirements with human capabilities; provide input to procedure development and provide input to staffing, training, and communications requirements of the plant. The method about task analysis and the screening methodology used to select the tasks for analysis are very important. Usually, the scope of task analysis is divided into two complementary activities: function-based task analysis and traditional task analysis, or operational sequence analysis. Function-based task analysis is the functional decomposition (goal-means analysis) for normal power operations. The traditional task analysis, or operational sequence analysis is developed for a representative set of operational and maintenance tasks. The operational sequence analysis include OSA-1 and OSA-2.The task analysis considers alerts, such as alarms and warnings, decision-making, response, teamwork and communication, workload task support, workplace factors, situational and performance shaping factors and hazard identification. The paper also describes the selected basis of selecting the operational sequence to be analyzed in the operational sequence analysis. The method about task analysis and the screening methodology used to select the tasks for analysis are very important. The objective of task analysis review is to verify that the applicant undertook analyses identifying the specific tasks needed to accomplish personnel functions, and also the alarms, information, control- and task-support required to complete those duties. The paper mainly describes not only the above content, but also the reviewing work about task analysis and questions found in reviewing. In this article, there are some examples about the function-based task analysis. At last, the verification for task analysis is addressed simply.

scholarly journals Radionuclide Disperse To Environtment From Pwr Reactor On Severe Accident Condition Using Maccs Program

KnE Energy ◽  
2016 ◽  
Vol 1 (1) ◽  
Author(s):  
Sri Kuntjoro
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Dispersion Coefficient ◽  
Computer Code ◽  
Dispersion Analysis ◽  
Severe Accident ◽  
Plant Site ◽  
Accident Condition

<p>The atmosphere is an important pathway in the transfer of radionuclides from nuclear power plants into the environment and population. Acceptance of radiation dose to the environment and population affected by the radionuclides release and site conditions surrounding of the nuclear power plant. The radionuclides release in the atmosphere is determined by the dispersion coefficient parameter. The aim of this paper is to obtain dispersion coefficient and radionuclide released in Sebagin (West Bangka district) caused by severe accident condition from the PWR Nuclear Power Plant. Dispersion analysis of radionuclides into the environment from nuclear power PWR on severe accident conditions have been done using MACCS program. Reference for the calculation of source term fraction is selected from calculation results of the MELCOR computer code and it is implemented to PWR reactors Westinghouse 3411 MWth subject. The calculation of radionuclides release performed using MACCS program for aspiring nuclear power plant site in West Bangka. Simulation calculations for the area radius from 0.80 kmup to 20 km from the nuclear power plant site are performed. Meteorological datas used in calculation are the meteorology data from Sebagin meteorological stations for the years of 2012 period. The result is the dispersion coefficient decreases as a function of time and distance. The concentration of radionuclides through soil pathway decreases as a function of the distance, and the dominant contributor of radionuclide radiation Xe-133 and   I-131. Radionuclide concentrations obtained through the air pathway decreases as a function of distance, and dominant contributors of radionuclide radiation is contributed also from I-131 and Xe-133. The presence of I-131 radionuclides are giving dangerous to humans, it is necessary to further treatment for prevent its impacts. </p>

Task Evaluation: Comparing Existing Curricula to Job Analysis Results

1985 ◽  
Vol 14 (2) ◽  
pp. 157-163 ◽  
Author(s):  
Jeffrey A. Cantor
Keyword(s):  
Systematic Review ◽  
Detailed Analysis ◽  
Task Analysis ◽  
Nuclear Power ◽  
Job Analysis ◽  
Analysis Data ◽  
Lesson Plans ◽  
Audit Trail ◽  
And Task

The process of modifying existing curricula to reflect current job analysis results is often approached in a disorganized and unplanned manner. This is not the case in the nuclear power generating industry. The purpose of this article is to describe one such process for comparing existing curricula to new job and task analysis data. This model systematically allows for coding of all tasks, followed by a grouping of tasks by specific departmental requirements. A review by task is then conducted to determine if lesson plans exist which are applicable to the task, and to determine if further task analysis is indicated. The model allows for systematic review of curricula against specific needs of the job. It provides a basis for detailed analysis in any area needing work. Further, the model establishes a documented audit trail between lesson plans and specific tasks required on the job.

scholarly journals An Integral Numerical Analysis of Impact of a Commercial Aircraft on Nuclear Containment

2019 ◽  
Vol 2019 ◽  
pp. 1-15
Author(s):  
Xiaoxin Wang ◽  
Qin Zhou ◽  
Li Shi ◽  
Haitao Wang ◽  
Xiaotian Li
Keyword(s):  
Reinforced Concrete ◽  
Nuclear Power ◽  
Power Plants ◽  
Function Analysis ◽  
Time History ◽  
Element Analysis ◽  
Fuel Distribution ◽  
Integral Analysis ◽  
Force Time ◽  
The Impact

After the September 11 attack, the resistant capability of containments against aircraft impacts is required to be assessed for newly constructed nuclear power plants (NPPs). In this paper, the crash of a commercial airplane Boeing 767-200ER on the reinforced concrete containment building of an NPP is analyzed using the missile-target interaction method. Two plane models with the same total weight but different fuel distribution are analyzed. The force-time history obtained by FEA (finite element analysis) is compared with the one calculated by the Riera function. In the integral analysis, the mesh sensitivity of the reinforced concrete containment model is studied, and recommendations are provided on the modelling of containment. The impact phenomenon and damage on the containment are investigated through the validated model. The fuel distribution in the aircraft is found to have strong influence on the damage of the containment, which indicates that the load distribution in the transverse direction is critical in the analysis of aircraft impact. The classic load-time function analysis is unable to incorporate this factor and may not be adequate to provide satisfactory results. For this reason, the application of an integral analysis is advantageous in the safety assessment of aircraft impact.

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