A Concept of Passive Safety, Pressurized Water Reactor System with Inherent Matching Nature of Core Heat Generation and Heat Removal

1995 ◽  
Vol 32 (9) ◽  
pp. 855-867 ◽  
Author(s):  
Yoshio MURAO ◽  
Fumimasa ARAYA ◽  
Takamichi IWAMURA ◽  
Keisuke OKUMURA
Keyword(s):  
Heat Generation ◽  
Heat Removal ◽  
Reactor System ◽  
Passive Safety ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor

Dynamic PRA Based on System Codes Coupling for Passive Safety System in Integral Pressurized Water Reactor

2020 ◽  
Author(s):  
Yi Mi ◽  
Akira Tokuhiro
Keyword(s):  
Natural Circulation ◽  
Passive Safety ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Active Systems ◽  
Water Reactor ◽  
System Parameters ◽  
Event Ordering ◽  
Passive Safety Systems ◽  
The Difference

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.

An Experimental Investigation of High Level Vibration on the Residual Heat Removal Line of a Pressurized Water Reactor

2009 ◽  
Author(s):  
Jeffrey A. Brown ◽  
Robert D. Blevins ◽  
H. Joseph Fernando
Keyword(s):  
Heat Removal ◽  
Coupled System ◽  
Acoustic Resonance ◽  
Acoustic Mode ◽  
Side Branch ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
High Level ◽  
Residual Heat

This paper presents the results of a scaled aero acoustic test that modeled a side branch resonance observed in the residual heat removal suction line of a large pressurized water reactor. Resolution of the acoustic resonance was sought by detuning the eddy shedding frequency from the fundamental side branch acoustic mode. The specific physical modifications and their ability to detune the coupled system are presented.

Analysis of PRHRS for an Integral Pressurized Water Reactor Using RELAP5/MOD3.4

2010 ◽  
Author(s):  
Xuhua Ye ◽  
Minjun Peng ◽  
Jiange Liu
Keyword(s):  
Natural Circulation ◽  
Heat Removal ◽  
Parameter Study ◽  
Pressurized Water Reactor ◽  
Steam Generators ◽  
Primary Coolant ◽  
Pressurized Water ◽  
Water Reactor ◽  
Pressure Peak ◽  
Residual Heat

An investigation on the thermal hydraulic characteristics of the passive residual heat removal system (PRHRS) which is used in an integral pressurized water reactor (INSURE-100) is presented in this paper. The main components of primary coolant system are enclosed in reactor vessel. Primary fluid flow circle is natural circulation. The PRHRS can remove the energy from the primary side as long as the residual heat exchanger (RHE) is submerged in the emergency cooldown tank (ECT). The parameter study is performed by considering the effects of an effective height between the steam generators and the RHE and a valve actuation time, which are useful for the design of the PRHRS. The mass flow in the PRHRS has been affected by the height difference between the steam generators and the RHE. The pressure peak of the primary side and PRHRS has been affected by the valve action time.

Loss of Feed Water Accident in an Integral Pressurized Water Reactor (IPWR)

2010 ◽  
Vol 171-172 ◽  
pp. 379-384
Author(s):  
Khan Salah Ud Din ◽  
Min Jun Peng ◽  
Muhammad Zubair
Keyword(s):  
Heat Removal ◽  
Feed Water ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Heat Removal System ◽  
Water Accident ◽  
Relap5 Code ◽  
Removal System ◽  
Residual Heat

In this paper research has been carried out on Loss of Feed Water Accident (LOFW) scenario of the Integral Pressurized Water Reactor ( IPWR) under two circumstances by the use of thermal hydraulic system code i.e Relap5/Mod3.4. In the first one, Passive Residual Heat Removal System (PRHRS) which is designed to absorb core residual heat in case of transient conditions is included which has the function of operating under the accident vulnerabilities. Concerning with the second case i.e without the use of PRHRS rather a tank of water which has the capacity of about 8% of the total feed water supply and is operated under accident scenario is considered. Taken into account these conditions,first the nodalization diagram of the two cases have been figured out then according to the LOFW accident time event scenario use the Relap5 code to simulate the accident. Finally the graphical explanation (separately) of the two cases with graphical approach as well as the conclusion is given at the end.

Natural Circulation Characteristics of an Integral Pressurized Water Reactor

2006 ◽  
Author(s):  
Junli Gou ◽  
Suizheng Qiu ◽  
Guanghui Su ◽  
Dounan Jia
Keyword(s):  
Steady State ◽  
Steam Generator ◽  
Natural Circulation ◽  
Heat Removal ◽  
Primary Circuit ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Secondary Side ◽  
Circulation Characteristics

Natural circulation potential is of great importance to the inherent safety of a nuclear reactor. This paper presents a theoretical investigation on the natural circulation characteristics of an integrated pressurized water reactor. Through numerically solved the one-dimensional model, the steady-state single phase conservative equations for the primary circuit and the steady-state two-phase drift-flux conservative equations for the secondary side of the once-through steam generator, the natural circulation characteristics are studied. Based on the preliminary calculation analysis, it is found that natural circulation mass flow rate is proportional to the exponential function of the power, and the value of the exponent is related to working conditions of the steam generator secondary side. The higher height difference between the core center and the steam generator center is favorable to the heat removal capacity of the natural circulation.

Heat Removal Tests for Pressurized Water Reactor Containment Spray by Large-Scale Facility

1983 ◽  
Vol 63 (2) ◽  
pp. 316-329 ◽  
Author(s):  
Yasuo Motoki ◽  
Mitsuo Naritomi ◽  
Mitsugu Tanaka ◽  
Gunji Nishio ◽  
Kazuichiro Hashimoto ◽  
...  
Keyword(s):  
Large Scale ◽  
Heat Removal ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor
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