Comparative analysis of auxiliary feedwater system and passive safety system under typical accident scenarios for integrated pressurized water reactor (IPWR)

Abstract An integral Pressurized Water Reactor (iPWR) type SMR was studied featuring Passive Safety Systems (PSSs). Different from active systems, PSSs are easily influenced by system parameters referred to as phenomenological factors such as heat loss, flow friction, oxidation, non-condensable gases and void fraction due to the low driving force of natural circulation. The system parameters also contribute to the uncertainty and dependency of PSS leading to the system unreliability. Thus, efforts are made to improve the reliability of PSS. A classical Probabilistic Risk Assessment (PRA) model describing active systems does not consider time evolution nor event ordering for PSS that dynamic PRA can accommodate. Here we developed and realized coupling between LabVIEW and CAFTA. Isolation Condenser System (ICS) is taken as the benchmark system due to the simple design in single phase without phase change phenomena in order to mainly remove decay heat and secondarily depressurize the reactor pressure vessel (RPV). A classical PRA model of ICS using CAFTA is coupled with real-time simulation of primary loop and ICS in LabVIEW, leading to a dynamic simulation result. The difference in failure probability using dynamic versus classical PRA revealed that for one there are more component demands with different event ordering, such that improved PSS reliability in the iPWR-type SMR designs is possible.

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Preliminary LOCA Analysis of Heating-Reactor of Advanced Low-Pressurized and Passive Safety System (HAPPY)

Volume 1: Operations and Maintenance, Engineering, Modifications, Life Extension, Life Cycle, and Balance of Plant; Instrumentation and Control (I&C) and Influence of Human Factors; Innovative Nuclear Power Plant Design and SMRs ◽

10.1115/icone26-81271 ◽

2018 ◽

Author(s):

Mian Xing ◽

Zhaocan Meng ◽

Xiaotao Liao ◽

Canhui Sun ◽

Shuming Zhang ◽

...

Keyword(s):

Conceptual Design ◽

District Heating ◽

Central Research Institute ◽

Safety System ◽

Passive Safety ◽

Central Research ◽

Pressurized Water ◽

Design And Optimization ◽

Passive Safety System ◽

Safety Features

SPICRI (State Power Investment Central Research Institute) is developing a new conceptual design of heating-reactor, named Heating-reactor of Advanced low-Pressurized and Passive safetY system (HAPPY), which is targeted for the district heating, desalination of seawater, and other heat applications. It is a 200MWth two-loop low-pressurized water reactor with low thermal parameters. The whole reactor vessel is deployed inside a shielding and cooling pool with thermal insulation measure. The conceptual design of HAPPY is described in this paper, including the design criteria, safety features, main parameters and main components. A preliminary safety analysis is carried out to provide a reference for the design and optimization of HAPPY. In this paper, four different LOCA analyses are described and compared. The results show that the current design can deal well with all the selected LOCA scenarios and the effectiveness of the safety systems is proved.

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Test Analysis of CAP1400 5cm Break Accident Based on the ACME Test Facility

Volume 3: Thermal-Hydraulics ◽

10.1115/icone24-60231 ◽

2016 ◽

Author(s):

Sheng Zhu

Keyword(s):

System Design ◽

Nuclear Power ◽

Reactor Core ◽

Safety System ◽

Test Facility ◽

Passive Safety ◽

Pressurized Water ◽

Power Simulation ◽

Passive Safety System ◽

Core Cooling

CAP1400 is a large pressurized water reactor based on the passive safety conception. An ACME (Advanced Core-cooling Mechanism Experiment) facility has been designed and constructed in order to validate that the CAP1400 system design is acceptable to mitigate the loss of coolant accident (LOCA). The ACME test facility is an isotonic pressure, 1/3-scale height and 1/54.32-scale power simulation of the prototype CAP1400 nuclear power plant. It contains the main-loop system, passive safety system, secondary steam system and auxiliary system etc. The all of ACME test matrix including 5 kinds 21 cases .In this paper, the test results and the Realp5 prediction of the cold leg 5cm break accident of CAP1400 are compared and analyzed to briefly evaluate the ACME capability. Furthermore, 3 different types of 5cm cold leg break test cases are presented, and the transient process, system responses and key parameters tendency are analyzed based on the test. The results indicate that the passive safety system design can successfully combine to provide a continuous removal of core decay heat and the reactor core remains to be covered with considerable margin for the 3 different 5cm cold leg break accidents.

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Tightness Analysis of Pressured Equipment in Experimental Test on Steam Condensation on the Cold Surface

Volume 4: Thermal Hydraulics ◽

10.1115/icone21-15888 ◽

2013 ◽

Author(s):

Lin Yang ◽

Liyong Han

Keyword(s):

Experimental Test ◽

Cooling System ◽

Passive Safety ◽

Natural Phenomena ◽

Pressurized Water Reactor ◽

Steam Condensation ◽

Cold Surface ◽

Pressurized Water ◽

University Of Wisconsin ◽

Passive Safety System

To maintain the containment within the design limits of pressure and temperature, the advanced pressurized water reactor (APWR) designed by Westinghouse uses a passive safety system to transfer the heat from inner containment to outside. The passive containment cooling system (PCCS) includes many natural phenomena mechanisms. Steam condensation is one of the most important phenomena. Most heat is removed by steam condensation on inside surface of the containment during the postulated design basic accidents (DBA). It is very significant for engineering designing and code developing to study the mechanism of steam condensation on cold surface. There was an experiment made by University of Wisconsin on it. In this paper, the structure pressure of the pressured equipment is calculated and the tightness is also analyzed.

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