Identify the Broken Loop and Break Sizes of Recirculation Line LOCA for a BWR6/MarkIII Plant

2020 ◽  
Author(s):  
Sheng-Dih Hwang ◽  
Te-Chuan Wang
Keyword(s):  
Steady State ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Temperature Difference ◽  
Nuclear Power ◽  
Preliminary Estimate ◽  
Control Room

Abstract In this study, the MAAP5 code was used as the tool to simulate a reactor operated at steady-state and occurred an LOCA with different break sizes at the recirculation loops. A BWR6/MarkIII nuclear power plant (NPP) was selected as the sample deck. By detecting the temperature difference between the loops, and then the break size could be determined including which loop and how is the break size. Moreover, this study had also analyzed the effect of break sizes affecting the characteristic of an NPP. These results were consistent with the previous experiments and codes. This study also suggested to the operators in the main control room (MCR), when they observed some syndrome of the acquired data and they could preliminary estimate which loop broke and how was the size of the LOCA. Those could help the operators to make a strategy to avoid the probability of core melted.

Verification of Alarm Displays for the Nuclear Power Plant With Two Modular High-Temperature Gas-Cooled Reactors

2018 ◽  
Author(s):  
Jia Qianqian ◽  
Guo Chao ◽  
Li Jianghai ◽  
Qu Ronghong
Keyword(s):  
High Temperature ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Modular Design ◽  
Early Stage ◽  
Integrated System ◽  
Alarm System ◽  
Control Room ◽  
High Temperature Gas

The nuclear power plant with two modular high-temperature gas-cooled reactors (HTR-PM) is under construction now. The control room of HTR-PM is designed. This paper introduces the alarm displays in the control room, and describes some verification and validation (V&V) activities of the alarm system, especially verification for some new human factor issues of the alarm system in the two modular design. In HTR-PM, besides the regular V&V similar to other NPPs, the interference effect of the alarm rings of the two reactor modules at the same time, and the potential discomfort of the two reactor operators after shift between them are focused. Verifications at early stage of the two issues are carried on the verification platform of the control room before the integrated system validation (ISV), and all the human machine interfaces (HMIs) in the control room, including the alarm system are validated in ISV. The test results on the verification platform show that the alarm displays and rings can support the operators understand the alarm information without confusion of the two reactors, and the shift between the two reactor operators have no adverse impact on operation. The results in ISV also show that the alarm system can support the operators well.

Application control chart concepts of designing a pre-alarm system in the nuclear power plant control room

2008 ◽  
Vol 238 (12) ◽  
pp. 3522-3527 ◽  
Author(s):  
Sheue-Ling Hwang ◽  
Jhih-Tsong Lin ◽  
Guo-Feng Liang ◽  
Yi-Jan Yau ◽  
Tzu-Chung Yenn ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Control Chart ◽  
Nuclear Power ◽  
Alarm System ◽  
Control Room ◽  
Power Plant Control ◽  
Plant Control

scholarly journals Baseline Study Methodology for Future Phases of Research on Nuclear Power Plant Control Room Technologies

2016 ◽  
Author(s):  
Katya Lee Le Blanc ◽  
Gordon Ross Bower ◽  
Rachael Ann Hill ◽  
Zachary Alexander Spielman ◽  
Brandon Charles Rice
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Control Room ◽  
Baseline Study ◽  
Power Plant Control ◽  
Study Methodology ◽  
Plant Control

Evaluation on Radiological Consequence Induced by MSLB Accident With Alternative Source Term

2010 ◽  
Author(s):  
Jingxi Li ◽  
Gaofeng Huang ◽  
Lili Tong
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Source Term ◽  
Control Room ◽  
Power Station ◽  
Alternative Source ◽  
Radiological Consequence

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.

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