Improving Human-System Interface Design Through Human Behavior Assessment in the Control Room of Nuclear Power Plants

Author(s):  
Xiaolu Dong ◽  
Dan Pan ◽  
Zhizhong Li ◽  
Yunbo Zhang ◽  
Yan Feng ◽  
...  
Work ◽  
2012 ◽  
Vol 41 ◽  
pp. 714-721 ◽  
Author(s):  
Yong Zhou ◽  
HaiYing Mu ◽  
Jianjun Jiang ◽  
Li Zhang

Author(s):  
Ronald L. Boring ◽  
David I. Gertman ◽  
Tuan Q. Tran ◽  
Brian F. Gore

Author(s):  
Eric Davey

This paper summarizes the findings from several observational studies to characterize the basis for a process monitoring strategy used by operators in ‘normal’ operations at CANDU nuclear power plants. These studies were undertaken in support of projects to develop improved control room displays and information systems to better support operators in both normal and abnormal operating situations. With the assistance of operators from several plants, an underlying basis for process monitoring was defined and a ‘generic’ strategy for monitoring process conditions in ‘normal’ operations has been established.


Author(s):  
Jingxi Li ◽  
Gaofeng Huang ◽  
Lili Tong

The major threat that nuclear power plants (NPPs) pose to the safety of the public comes from the large amount radioactive material released during design-basis accidents (DBAs). Additionally, many aspects of Control Room Habitability, Environmental Reports, Facility Siting and Operation derive from the design analyses that incorporated the earlier accident source term and radiological consequence of NPPs. Depending on current applications, majority of Chinese NPPs adopt the method of TID-14844, which uses the whole body and thyroid dose criteria. However, alternative Source Term (AST) are commonly used in AP1000 and some LWRs (such as Beaver Valley Power Station, Units No. 1 and No. 2, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 And 2, Kewaunee Power Station and so on), so it is attempted to adopt AST in radiological consequence analysis of other nuclear power plants. By introducing and implementing the method of AST defined in RG 1.183 and using integral safety analysis code, a pressurized water reactor (PWR) of 900 MW nuclear power plant analysis model is constructed and the radiological consequence induced by Main Steam Line Break (MSLB) accident is evaluated. For DBA MSLB, the fractions of core inventory are assumed to be in the gap for various radionuclides and then the release from the fuel gap is assumed to occur instantaneously with the onset of assumed damage. According to the assumptions for evaluating the radiological consequences of PWR MSLB, dose calculation methodology is performed with total effective dose equivalent (TEDE) which is the criteria of dose evaluation. Compared with dose criteria of RG 1.183, the dose of control room, exclusion area boundary and outer boundary of low population zone are acceptable.


Author(s):  
Kevin LaFerriere ◽  
Jessica Stevens ◽  
Ryan Flamand NuScale

The NuScale Small Modular Reactor (SMR) is premised on well-established nuclear technology principles with a focus on integration of components, simplification or elimination of systems, automation, and use of passive safety features. Traditional nuclear power plants have in some cases operated up to four modules from a single control room. Due to the unique nontraditional operating characteristics of this technology a state-of-the art control room design was needed to ensure proper staffing totals for monitoring and control of multiple modules (twelve) from a single control room. To accomplish this, the human system interface and control room layout must translate the functional and task requirements needed for safe operation of the plant into the detailed design of workstations, alarms, controls, navigation, and other needs of the control room operations staff.


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