scholarly journals Two-Phase Flow in the Natural Circulation of a Pressurized Water Reactor

2020 ◽  
Author(s):  
Mohammad W. Abdulrahman ◽  
Mikdam M. Saleh
Keyword(s):  
Two Phase Flow ◽  
Natural Circulation ◽  
Phase Flow ◽  
Pressurized Water Reactor ◽  
Two Phase ◽  
Pressurized Water ◽  
Water Reactor

Interfacial behavior of the air-water counter-current two-phase flow in a 1/30 scale-down of pressurized water reactor (PWR) hot leg

2018 ◽  
Vol 116 ◽  
pp. 376-387
Author(s):  
Apip Badarudin ◽  
Andriyanto Setyawan ◽  
Okto Dinaryanto ◽  
Arif Widyatama ◽  
Indarto ◽  
...  
Keyword(s):  
Two Phase Flow ◽  
Phase Flow ◽  
Pressurized Water Reactor ◽  
Interfacial Behavior ◽  
Two Phase ◽  
Pressurized Water ◽  
Water Reactor ◽  
Scale Down ◽  
Counter Current

Development of Steam Generator Model Using RELAP5 to Simulate Two-Phase Natural Circulation for Pressurized Water Reactor

2014 ◽  
Author(s):  
Shinya Miyata ◽  
Satoru Kamohara ◽  
Wataru Sakuma ◽  
Hiroaki Nishi
Keyword(s):  
Steam Generator ◽  
Natural Circulation ◽  
Injection System ◽  
Pressurized Water Reactor ◽  
Two Phase ◽  
Pressurized Water ◽  
Water Reactor ◽  
Loss Of Coolant ◽  
Mass Inventory ◽  
Core Cooling

In typical pressurized water reactor (PWR), to cope with beyond design basis events such as station black out (SBO) or small break loss of coolant accident with safety injection system failure, injection from accumulator sustains core cooling by compensating for loss of coolant. Core cooling is sustained by single- or two-phase natural circulation or reflux condensation depending on primary coolant mass inventory. Behavior of the natural circulation in PWR has been investigated in the facilities such as Large Scale Test Facility (LSTF) which is a full-height and full-pressure and thermal-hydraulic simulator of typical four-loop PWR. Two steady-state natural circulation tests were conducted in LSTF at both high and low pressure. These two tests were conducted changing the primary mass inventory as a test parameter, while keeping the other parameters such as core power, steam generator (SG) pressure, and steam generator water level as they are. Mitsubishi Heavy Industries (MHI) plans new natural circulation tests to cover wider range of core power and pressure as test-matrix (including the previous LSTF tests) to validate applicability of the model in wider range of core power and pressure conditions including the SBO conditions. In this paper, the previous LSTF natural circulation tests are reviewed and the new test plan will be described. Additionally, MHI also started a feasibility study to improve the steam generator tube and inlet/outlet plenum model using the M-RELAP5 code [4]. Newly developed model gives reasonable agreement with the previous LSTF tests and applies to the new test conditions. The feasibility findings will also be described in this paper.

Numerical Simulation of Flow Regime Transition From Slug to Wavy Flow in Helically Coiled Tubes

2016 ◽  
Author(s):  
Zhaoxu Li ◽  
Hongye Zhu
Keyword(s):  
Heat Transfer ◽  
Numerical Simulation ◽  
Centrifugal Force ◽  
Slug Flow ◽  
Two Phase Flow ◽  
Flow Regimes ◽  
Phase Flow ◽  
Pressurized Water Reactor ◽  
Two Phase ◽  
Pressurized Water

Two-phase flow in helically coiled tubes is becoming the interest of many investigators because of its importance in various applications, such as nuclear engineering, chemical engineering, refrigerating engineering and power engineering. Compared with U-type tubes used in pressurized water reactor (PWR), helically coiled tubes have advantages in size, heat transfer capacity, thermal stress toleration and two-phase stability. Accordingly the helically coiled tubes have been utilized in the steam generators of the next general reactors, such as gas-cooled reactor, fast breeder reactor and integrated pressurized water reactor. In helically coiled tubes the characteristics of momentum and heat transfer are distinct from those in straight tubes due to the presence of centrifugal force, especially for two-phase flow. Meanwhile, the transitions of flow regime, which is the crucial knowledge for the designers to determine the heat transfer rates and flow resistance, are also significantly affected by the centrifugal force. In this study, two-phase flow regimes in helically coiled tubes are investigated. Computational fluid dynamics (CFD), using fractional volume of fluid (VOF) model, is carried out to simulate wavy and slug flow regimes in helically coiled tubes. The corresponding experiment is also conducted to visualize these flow regimes at different superficial flow velocities. Numerical simulation results actually reflect the influence of centrifugal force on the two-phase flow and show a good agreement with the photographs captured from the experiment. Based on the simulations at different superficial flow velocities, the boundary between the slug and wavy flow regimes is predicted, in addition, compared with that in inclined tubes. The comparison indicates that centrifugal force could induce the appearance of wavy flows in advance and prompt the transition from slug flow to wavy flow.

Characteristics of Downcomer Boiling Phenomena During the Reflood Phase of a Postulated Large Break LOCA for the APR1400

2006 ◽  
Author(s):  
Byong-Jo Yun ◽  
Dong-Jin Euh ◽  
Won-Man Park ◽  
Young-Jung Youn ◽  
Chul-Hwa Song
Keyword(s):  
Two Phase Flow ◽  
Transient Behavior ◽  
Visual Observation ◽  
Test Facility ◽  
Phase Flow ◽  
Pressurized Water Reactor ◽  
Two Phase ◽  
Pressurized Water ◽  
Flow Parameters ◽  
Core Cooling

Downcomer boiling phenomena in a conventional pressurized water reactor have an important effect on the transient behavior of a postulated large-break LOCA (LBLOCA), because it can degrade the hydraulic head of the coolant in the downcomer and consequently affect the reflood flow rate for a core cooling. To investigate the thermal hydraulic behavior in the downcomer region, a test program for a downcomer boiling is being progressed in the reflood phase of a postulated LBLOCA. For this, the test facility was designed as a one side heated rectangular test section which adopts a full-pressure, full-height, and full-size downcomer-gap approach, but with the circumferential length reduced 47.08-fold. The test was performed by dividing it into two-phases: (I) visual observation and acquisition of the global two-phase flow parameters and (II) measurement of the local two-phase flow parameters on the measuring planes along five elevations. In the present paper, the test results of Phase-I and parts of Phase-II were introduced.

TWO-PHASE FLOW INSTABILITIES ANALYSES IN BOILING WATER REACTOR

2016 ◽  
Author(s):  
Antonella Lombardi Costa ◽  
WILMER ARUQUIPA COLOMA ◽  
Antonella Lombardi Costa ◽  
Claubia Pereira ◽  
Maria Veloso ◽  
...  
Keyword(s):  
Two Phase Flow ◽  
Boiling Water ◽  
Flow Instabilities ◽  
Boiling Water Reactor ◽  
Phase Flow ◽  
Two Phase ◽  
Water Reactor
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