scholarly journals Experimental Research on Passive Residual Heat Removal System of Chinese Advanced PWR

2014 ◽  
Vol 2014 ◽  
pp. 1-8 ◽  
Author(s):  
Zhuo Wenbin ◽  
Huang Yanping ◽  
Xiao Zejun ◽  
Peng Chuanxin ◽  
Lu Sansan
Keyword(s):  
Steady State ◽  
Scaling Laws ◽  
Heat Removal ◽  
Experimental Results ◽  
Test Facility ◽  
Semiempirical Model ◽  
Pressurized Water ◽  
Heat Removal System ◽  
Removal System ◽  
Residual Heat

Passive residual heat removal system (PRHRS) for the secondary loop is one of the important features for Chinese advance pressurized water reactor (CAPWR). To prove the safety characteristics of CAPWR, serials of experiments have been done on special designed PRHRS test facility in the former stage. The test facility was built up following the scaling laws to preserve the similarity to CAPWR. A total of more than 300 tests have been performed on the test facility, including 90% steady state cases and 10% transient cases. A semiempirical model was generated for passive heat removal functions based on the experimental results of steady state cases. The dynamic capability characteristics and reliability of passive safety system for CAPWR were evidently proved by transient cases. A new simulation code, MISAP2.0, has been developed and calibrated by experimental results. It will be applied in future design evaluation and optimization works.

Experimental Research on Non-Condensable Gases Effects in Passive Decay Heat Removal System

2016 ◽  
Author(s):  
Yang Liu ◽  
Haijun Jia ◽  
Li Weihua
Keyword(s):  
High Pressure ◽  
Heat Removal ◽  
Volume Ratio ◽  
Test Facility ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Decay Heat ◽  
Optimizing Design ◽  
Heat Removal System ◽  
Removal System

Passive decay heat removal (PDHR) system is important to the safety of integral pressurized water reactor (IPWR). In small break LOCA sequence, the depressurization of the reactor pressure vessel (RPV) is achieved by the PDHR that remove the decay heat by condensing steam directly through the SGs inside the RPV at high pressure. The non-condensable gases in the RPV significantly weaken the heat transfer capability of PDHR. This paper focus on the non-condensable gas effects in passive decay heat removal system at high pressure. A series of experiments are conducted in the Institute of Nuclear and New Energy Technology test facility with various heating power and non-condensable gas volume ratio. The results are significant to the optimizing design of the PDHR and the safety operation of the IPWR.

Modeling and Experiments of a Passive Decay Heat Removal System for Advanced Nuclear Reactors

2017 ◽  
Author(s):  
Andrea Bersano ◽  
Mario De Salve ◽  
Cristina Bertani ◽  
Nicolò Falcone ◽  
Bruno Panella
Keyword(s):  
Scaling Laws ◽  
Nuclear Reactors ◽  
Natural Circulation ◽  
Heat Removal ◽  
Experimental Results ◽  
Passive Systems ◽  
Decay Heat ◽  
Safety Systems ◽  
Heat Removal System ◽  
Removal System

Within the field of research and development of innovative nuclear reactors, in particular Generation IV reactors and Small Modular Reactors (SMR), the design and the improvement of safety systems play a crucial role. Among all the safety systems high attention is dedicated to passive systems that do not need external energy to operate, with a very high reliability also in the case of station blackout, and which are largely used in evolutionary technology reactors. The aim of this work is the experimental and numerical analysis of a passive system that operates in natural circulation in order to study the mechanism and the efficiency of heat removal. The final goal is the development of a methodology that can be used to study this class of systems and to assess the thermal-hydraulic code RELAP5 for these specific applications. Starting from a commercial size system, which is the decay heat removal system of the experimental lead cooled reactor ALFRED, an experimental facility has been designed, built and tested with the aim of studying natural circulation in passive systems for nuclear applications. The facility has been simulated and optimized using the thermal-hydraulic code RELAP5-3D. During the experimental tests, temperatures and pressures are measured and the experimental results are compared with the ones predicted by the code. The results show that the system operates effectively, removing the given thermal power. The code can predict well the experimental results but high attention must be dedicated to the modeling of components where non-condensable gases are present (condenser pool and surrounding ambient). This facility will be also used to validate the scaling laws among systems that operate in natural circulation.

Thermal Hydraulic Characteristics of the Integral Type Reactor, SMART-P for Validation of the Heat Removal Capability

2006 ◽  
Author(s):  
K. Y. Choi ◽  
S. Cho ◽  
S. J. Yi ◽  
H. S. Park ◽  
N. H. Choi ◽  
...  
Keyword(s):  
System Design ◽  
Heat Removal ◽  
Test Facility ◽  
Integral Type ◽  
Control Rod ◽  
Type Reactor ◽  
Loss Of Coolant ◽  
Heat Removal System ◽  
Removal System ◽  
Residual Heat

The SMART is an integral type reactor with new innovative design features aimed at achieving a highly enhanced safety and improved economics. This paper focuses on the thermal hydraulic experimental program for the development of SMART. Thermal hydraulic responses for the transient operations of the SMART-P are experimentally investigated by using an integral effect test facility. The test facility (VISTA) has been constructed to simulate the SMART-P, which is a pilot plant of the SMART. The VISTA facility is a full height and 1/96 volume scaled test facility with respect to the SMART-P with a power of 65MWt. In the present study, the VISTA facility was subjected to various transient conditions in order to understand the thermal-hydraulic responses following transients and finally to verify the system design of the SMART-P. Several experiments, including a heatup, a main coolant pump (MCP) speed change, and a power change, have been performed to investigate the heat transfer characteristics and the natural circulation performance of the primary system and the Passive Residual Heat Removal System (PRHRS) of the SMART-P by using the VISTA facility. Performance tests of a passive residual heat removal system (PRHRS) have also been carried out for its design optimization. Besides, several design basis accidents, such as an increase or a decrease of the feedwater flow, a loss of coolant flow, a control rod withdrawal, and a limited case of a loss of coolant accident (LOCA) on the line to the gas cylinder are under investigation in order to understand the thermal-hydraulic responses and finally to verify the system design of the SMART-P. Especially, the details of the experimental results for a loss of feedwater accident and a power increase accident due to a control rod withdrawal are explored in the present study.

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.

scholarly journals Thermal Hydraulic Analysis of a Passive Residual Heat Removal System for an Integral Pressurized Water Reactor

2009 ◽  
Vol 2009 ◽  
pp. 1-12 ◽  
Author(s):  
Junli Gou ◽  
Suizheng Qiu ◽  
Guanghui Su ◽  
Douna Jia
Keyword(s):  
Heat Exchanger ◽  
Steam Generator ◽  
Heat Removal ◽  
Heat Transfer Area ◽  
Pressurized Water Reactor ◽  
Height Difference ◽  
Pressurized Water ◽  
Heat Removal System ◽  
Removal System ◽  
Residual Heat

A theoretical investigation on the thermal hydraulic characteristics of a new type of passive residual heat removal system (PRHRS), which is connected to the reactor coolant system via the secondary side of the steam generator, for an integral pressurized water reactor is presented in this paper. Three-interknited natural circulation loops are adopted by this PRHRS to remove the residual heat of the reactor core after a reactor trip. Based on the one-dimensional model and a simulation code (SCPRHRS), the transient behaviors of the PRHRS as well as the effects of the height difference between the steam generator and the heat exchanger and the heat transfer area of the heat exchanger are studied in detail. Through the calculation analysis, it is found that the calculated parameter variation trends are reasonable. The higher height difference between the steam generator and the residual heat exchanger and the larger heat transfer area of the residual heat exchanger are favorable to the passive residual heat removal system.

Nuclear Power Plant Passive Residual Heat Removal System Design and Simulation

2013 ◽  
Author(s):  
Xiao Yuan ◽  
Minjun Peng ◽  
Genglei Xia
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Heat Removal ◽  
Primary Circuit ◽  
Two Phase ◽  
Pressurized Water ◽  
Heat Removal System ◽  
Removal System ◽  
Residual Heat

The passive safety systems employed in the design of pressurized water reactor (PWR) can accomplish the inherent safety functions and mitigate the consequences of the postulated accidents. In this paper, a passive residual heat removal system (PRHRs) is designed for a certain nuclear power plant. The RELAP5/MOD3.4 code was used to analyze the operation characteristics of the PRHRs. It shows the PRHRs could remove the decay heat from the primary loop effectively, and the single-phase and two-phase natural circulations could respectively establish in the primary circuit and the PRHRs circuit.

Evaluation of a SGTR Accident in the Multi-Application Integrated Pressurized Water Reactor

2010 ◽  
Author(s):  
Liguo Jiang ◽  
Minjun Peng ◽  
Jiange Liu
Keyword(s):  
Steam Generator ◽  
Flow Model ◽  
Heat Removal ◽  
Injection System ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Heat Removal System ◽  
Removal System ◽  
Residual Heat

One of more frequent events in the Pressurized Water Reactor (PWR) is Steam Generator Tube Rupture (SGTR) accident, which is among the main accidents in the field of nuclear safety. This paper studies the SGTR event in the Multi-application Integrated Pressurized Water Reactor (IPWR) using the best-estimate thermal-hydraulic code RELAP5/MOD3.4. In the reactor of IPWR, several Once-Through Steam Generator (OTSG) cassettes are used and located between the core support and the pressure vessel. The tube rupture location is on the top of the tube sheet of a steam generator. Three different tube rupture modeling methods and several different subcooled discharge coefficients in the critical flow model are considered and compared. In the safety analysis, high pressure safety injection system, core makeup system and Passive Residual Heat Removal System (PRHRS) that would affect the accident consequences are considered.

Optimization and Analysis for Passive Residual Heat Removal System of an Integral Pressurized Water Reactor

2009 ◽  
Author(s):  
Shoubao Dai ◽  
Minjun Peng ◽  
Jiange Liu
Keyword(s):  
Natural Circulation ◽  
Heat Removal ◽  
Design Parameters ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Secondary Loop ◽  
Heat Removal System ◽  
Removal System ◽  
Peak Value ◽  
Residual Heat

The characteristics of passive safety systems for an integral pressurized water reactor (IPWR) are quite different from the general reactor because of special configuration and dangerous run environment. Passive residual heat removal system (PRHRS) for the IPWR with three-interknited natural coolant circulation loops, safely remove the core decay heat to the ultimate heat sink. Using RELAP5/MOD3.4 code to simulate this system during the reactor trip, analyses the steady-state and transient-state thermohydraulic behaviors for the IPWR and its PRHRS, and the effects of design parameters on the system. it is found that on the initial period of reactor trip, due to the establishment of the natural circulation in three loops uncompleted, the secondary loop pressure have the peak value. Through analyzing the effects of design parameters on the system, the PRHRS are optimized. The results show that the larger the residual heat exchanger (RHE) heat transfer area and the higher the height difference between the steam generator and the residual heat exchanger, the easier the establishment of the natural circulation in the third loop, but which make the peak value of the secondary loop pressure higher. According to set the compensating water tank, which is parallel connected to the RHE, can lighten the higher the peak value of the secondary loop pressure, and optimize the design of PRHRS.

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