scholarly journals Calculation Method of Passive Residual Heat Removal Heat Exchanger and Numerical Simulation

2014 ◽  
Vol 02 (09) ◽  
pp. 8-14 ◽  
Author(s):  
Qiming Men ◽  
Xuesheng Wang ◽  
Xiang Zhou ◽  
Xiangyu Meng
Keyword(s):  
Numerical Simulation ◽  
Heat Exchanger ◽  
Calculation Method ◽  
Heat Removal ◽  
Residual Heat

scholarly journals Heat Transfer Analysis of Passive Residual Heat Removal Heat Exchanger under Natural Convection Condition in Tank

2014 ◽  
Vol 2014 ◽  
pp. 1-8 ◽  
Author(s):  
Qiming Men ◽  
Xuesheng Wang ◽  
Xiang Zhou ◽  
Xiangyu Meng
Keyword(s):  
Heat Transfer ◽  
Heat Exchanger ◽  
Thermal Stratification ◽  
Heat Removal ◽  
Heat Transfer Analysis ◽  
Average Error ◽  
Water Tank ◽  
Shaped Tube ◽  
Heat Transfer Calculation ◽  
Residual Heat

Aiming at the heat transfer calculation of the Passive Residual Heat Removal Heat Exchanger (PRHR HX), experiments on the heat transfer of C-shaped tube immerged in a water tank were performed. Comparisons of different correlation in literatures with the experimental data were carried out. It can be concluded that the Dittus-Boelter correlation provides a best-estimate fit with the experimental results. The average error is about 0.35%. For the tube outside, the McAdams correlations for both horizontal and vertical regions are best-estimated. The average errors are about 0.55% for horizontal region and about 3.28% for vertical region. The tank mixing characteristics were also investigated in present work. It can be concluded that the tank fluid rose gradually which leads to a thermal stratification phenomenon.

Experimental Study on Natural Convection Heat Transfer Outside Tube Bundle in Space Under Low Temperature Difference

2020 ◽  
Author(s):  
Junxiu Xu ◽  
Ming Ding ◽  
Changqi Yan ◽  
Guangming Fan
Keyword(s):  
Heat Transfer ◽  
Heat Exchanger ◽  
Low Temperature ◽  
Tube Bundle ◽  
Convection Heat Transfer ◽  
Heat Removal ◽  
Natural Convection Heat Transfer ◽  
Convection Heat ◽  
Reactor Pool ◽  
Residual Heat

Abstract The Passive Residual Heat Removal System (PRHRS) is very important for the safety of the heating reactor after shutdown. PRHRS is a natural circulation system driven by density difference, therefore, the heat transfer performance of the Passive Residual Heat Removal Heat Exchanger (PRHR HX) has a great impact to the heat transfer efficiency of PRHRS. However, the most research object of PRHR HX is the C-shape heat exchanger at present, which located in In-containment Refueling Water Storage Tank (IRWST). This heat exchanger is mainly used for the PRHRS of nuclear power plants. In the swimming pool-type low-temperature heating reactor (SPLTHR), the PRHR HX is placed in the reactor pool, which the pressure and temperature of the reactor pool are relatively low, and the outside heat transfer mode of tube bundle is mainly natural convection heat transfer. In this study, a miniaturized single-phase pool water cooling system was built to investigate the natural convective heat transfer coefficient of the heat exchanger under the large space and low temperature conditions. The experimental data had been compared with several correlations. The results show that the predicted value of Yang correlation is the closest to the experimental data, which the maximum deviation is about 11%.

Analysis of Heat Transfer and Flow Characteristics of AP1000 Passive Residual Heat Removal Heat Exchanger

2014 ◽  
Author(s):  
Xu Xie ◽  
Changhua Nie ◽  
Li Zhan ◽  
Hua Zheng ◽  
Pengzhou Li ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Heat Exchanger ◽  
Temperature Difference ◽  
Tube Bundle ◽  
Natural Circulation ◽  
Heat Removal ◽  
Large Temperature ◽  
Water Tank ◽  
Secondary Side ◽  
Residual Heat

In this paper, the computational fluid dynamics (CFD) method is applied to the thermal-hydraulic analysis, while the porous media model is used to simplify AP1000 passive residual heat removal heat exchanger tube. The temperature as well as flow distribution in the secondary side of the heat exchanger are obtained, aiming at analysis of natural circulation ability. It can be noted that the fluid in the secondary side of heat exchanger moves driven by the effect of thermal buoyancy, forming the natural cycle, which takes away heat in tube bundle region. The heat transfer in water tank is mainly enhanced by vortex and turbulent flow, caused by the large resistance of tube bundle region as well as large temperature difference. This phenomenon is obvious especially for the recirculation of flow near the tube bundle. The enduring change of flow rate and direction enhance the heat transfer. Besides, the big temperature difference helps to increase the driving effect of natural circulation. Consequently, the heat transfer of the tank is enhanced by above mechanism. The results of this study contribute to the capacity analysis of passive residual heat removal of natural circulation system, providing valuable information for safe operation of AP1000.

AP1000 Passive Residual Heat Removal Heat Exchanger Confirmatory Analysis

2006 ◽  
Author(s):  
Richard F. Wright ◽  
James R. Schwall ◽  
Creed Taylor ◽  
Naeem U. Karim ◽  
Jivan G. Thakkar ◽  
...  
Keyword(s):  
Heat Exchanger ◽  
Cooling System ◽  
Heat Removal ◽  
Passive Safety ◽  
Tube Heat Exchanger ◽  
Removal Capacity ◽  
Safety Margins ◽  
Final Design ◽  
Safety Features ◽  
Residual Heat

The AP1000 is an 1100 MWe advanced nuclear power plant that uses passive safety features to enhance plant safety and to provide significant and measurable improvements in plant simplification, reliability, investment protection and plant costs. The AP1000 received final design approval from the US-NRC in 2004. The AP1000 design is based on the AP600 design that received final design approval in 1999. Wherever possible, the AP1000 plant configuration and layout was kept the same as AP600 to take advantage of the maturity of the design and to minimize new design efforts. As a result, the two-loop configuration was maintained for AP1000, and the containment vessel diameter was kept the same. It was determined that this significant power uprate was well within the capability of the passive safety features, and that the safety margins for AP1000 were greater than those of operating PWRs. A key feature of the passive core cooling system is the passive residual heat removal heat exchanger (PRHR HX) that provides decay heat removal for postulated LOCA and non-LOCA events. The PRHR HX is a C-tube heat exchanger located in the in-containment refueling water storage tank (IRWST) above the core promoting natural circulation heat removal between the reactor cooling system and the tank. Component testing was performed for the AP600 PRHR HX to determine the heat transfer characteristics and to develop correlations to be used for the AP1000 safety analysis codes. The data from these tests were confirmed by subsequent integral tests at three separate facilities including the ROSA facility in Japan. Owing to the importance of this component, an independent analysis has been performed using the ATHOS-based computational fluid dynamics computer code PRHRCFD. Two separate models of the PRHR HX and IRWST have been developed representing the ROSA test geometry and the AP1000 plant geometry. Confirmation of the ROSA test results were used to validate PRHRCFD, and the AP1000 plant model was used to confirm the heat removal capacity for the full-sized heat exchanger. The results of these simulations show that the heat removal capacity of the PRHR HX is conservatively represented in the AP1000 safety analyses.

Normal Operational State Results of Integral Pressurized Water Reactor by Using Relap5/Mod3.4

2010 ◽  
Vol 171-172 ◽  
pp. 374-378
Author(s):  
Khan Salah Ud Din ◽  
Min Jun Peng ◽  
Muhammad Zubair
Keyword(s):  
Heat Exchanger ◽  
Nuclear Power ◽  
Transient Analysis ◽  
Heat Removal ◽  
Zirconium Hydride ◽  
Water Tank ◽  
Nuclear Power Reactor ◽  
Pressurized Water ◽  
Conceptual Study ◽  
Residual Heat

In this paper, a research has been carried out on the normal operational state of IPWR by using the thermal hydraulic system code Relap5/Mod3.4.In this study the conceptual study analysis of the reactor named as Inherent Safe Uranium Zirconium Hydride Nuclear Power Reactor INSURE-100 is considered but is only based on the conceptual study so the current research focuses on the normal operational of the reactor by using the Relap5/Mod3.4 code. For this purpose, two passive safety methods have been included for the safe operation as well as for transient analysis of the reactor. In the first one, Passive Residual Heat Removal System (PRHRS) has been modeled by taking the heat exchanger out of the Reactor Pressure Vessel (RPV) in the water tank so that it can absorb core decay heat in case of transient conditions and in the second one heat exchanger is placed both in inside and outside the RPV so that there can be another way to absorb the core residual heat. Considering these concepts figured out the normal operational state of the reactor by using Relap5/Mod3.4 in comparison with the conceptual design study of the reactor under consideration and the results extracted can be a good agreement for the transient analysis of the reactor.

Assessment of ASME Code Examinations on Regenerative, Letdown and Residual Heat Removal Heat Exchangers

2005 ◽  
Author(s):  
S. R. Gosselin ◽  
F. A. Simonen ◽  
S. E. Cumblidge ◽  
G. A. Tinsley ◽  
B. Lydell ◽  
...  
Keyword(s):  
Heat Exchanger ◽  
Heat Exchangers ◽  
Power Plants ◽  
National Laboratory ◽  
Operating Experience ◽  
Heat Removal ◽  
Pacific Northwest National Laboratory ◽  
Regenerative Heat ◽  
Asme Code ◽  
Residual Heat

Inservice inspection requirements for pressure retaining welds in the regenerative, letdown, and residual heat removal heat exchangers are prescribed in Section XI Articles IWB and IWC of the ASME Boiler and Pressure Vessel Code. Accordingly, volumetric and/or surface examinations are performed on heat exchanger shell, head, nozzle-to-head, and nozzle-to-shell welds. Inspection difficulties associated with the implementation of these Code-required examinations have forced operating nuclear power plants to seek relief from the U.S. Nuclear Regulatory Commission. The nature of these relief requests are generally concerned with metallurgical factors, geometry, accessibility, and radiation burden. Over 60% of licensee requests to the NRC identify significant radiation exposure burden as the principal reason for relief from the ASME Code examinations on regenerative heat exchangers. For the residual heat removal heat exchangers, 90% of the relief requests are associated with geometry and accessibility concerns. Pacific Northwest National Laboratory was funded by the NRC Office of Nuclear Regulatory Research to review current practice with regard to volumetric and/or surface examinations of shell welds of letdown heat exchangers, regenerative heat exchangers, and residual (decay) heat removal heat exchangers. Design, operating, common preventative maintenance practices, and potential degradation mechanisms were reviewed. A detailed survey of domestic and international PWR-specific operating experience was performed to identify pressure boundary failures (or lack of failures) in each heat exchanger type and NSSS design. The service data survey was based on the PIPExp® database and covers PWR plants worldwide for the period 1970–2004. Finally a risk assessment of the current ASME Code inspection requirements for residual heat removal, letdown, and regenerative heat exchangers was performed. The results were then reviewed to discuss the examinations relative to plant safety and occupational radiation exposures.

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