scholarly journals Effect of Flow Blockage on the Coolability during Reflood in a 2×2 Rod Bundle

2014 ◽  
Vol 2014 ◽  
pp. 1-9 ◽  
Author(s):  
Kihwan Kim ◽  
Byung-Jae Kim ◽  
Young-Jung Youn ◽  
Hae-Seob Choi ◽  
Sang-Ki Moon ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Test Facility ◽  
Pressurized Water Reactor ◽  
Two Phase ◽  
Flow Area ◽  
Pressurized Water ◽  
Fuel Rods ◽  
Rod Bundle ◽  
Loss Of Coolant Accident ◽  
Two Phase Heat Transfer

During the reflood phase of a large-break loss-of-coolant accident (LBLOCA) in a pressurized-water reactor (PWR), the fuel rods can be ballooned or rearranged owing to an increase in the temperature and internal pressure of the fuel rods. In this study, an experimental study was performed to understand the thermal behavior and effect of the ballooned region on the coolability using a 2 × 2 rod bundle test facility. The electrically heated rod bundle was used and the ballooning shape of the rods was simulated by superimposing hollow sleeves, which have a 90% blockage ratio. Forced reflood tests were performed to examine the transient two-phase heat transfer behavior for different reflood rates and rod powers. The droplet behaviors were also investigated by measuring the velocity and size of droplets near the blockage region. The results showed that the heat transfer was enhanced in the downstream of the blockage region, owing to the reduced flow area of the subchannel, intensification of turbulence, and deposition of the droplet.

Numerical Prediction of Void Distribution in Two-Phase Turbulent Flow in a Sub-Channel

2009 ◽  
Author(s):  
Tsutomu Ikeno ◽  
Tatsuya Sasakawa ◽  
Isao Kataoka
Keyword(s):  
Turbulent Flow ◽  
Transport Equation ◽  
Nucleate Boiling ◽  
Pressurized Water Reactor ◽  
Two Phase ◽  
Simulation Code ◽  
Pressurized Water ◽  
Void Fractions ◽  
Rod Bundle ◽  
Void Distribution

Numerical simulation code for predicting void distribution in two-phase turbulent flow in a sub-channel was developed. The purpose is to obtain a profile of void distribution in the sub-channel. The result will be used for predicting a heat flux at departure from nucleate boiling (DNB) in a rod bundle for the pressurized water reactor (PWR). The fundamental equations were represented by a generalized transport equation, and the transport equation was transformed onto the generalized coordinate system fitted to the rod surface and the symmetric lines in the sub-channel. Using the finite-volume method the transport equation was discretized for the SIMPLE algorism. The flow field and void fraction at the steady state were calculated for different average void fractions. The computational result for atmospheric pressure condition was successfully compared with experimental data. Sensitivity analysis for the PWR condition was performed, and the result showed that the secondary flow slightly contributed to homogenizing the void distribution.

scholarly journals Validation of Advanced Computer Codes for VVER Technology: LB-LOCA Transient in PSB-VVER Facility

2012 ◽  
Vol 2012 ◽  
pp. 1-15 ◽  
Author(s):  
A. Del Nevo ◽  
M. Adorni ◽  
F. D'Auria ◽  
O. I. Melikhov ◽  
I. V. Elkin ◽  
...  
Keyword(s):  
Natural Circulation ◽  
Test Facility ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Mass Inventory ◽  
Computer Codes ◽  
Set Up ◽  
Advanced Computer

The OECD/NEA PSB-VVER project provided unique and useful experimental data for code validation from PSB-VVER test facility. This facility represents the scaled-down layout of the Russian-designed pressurized water reactor, namely, VVER-1000. Five experiments were executed, dealing with loss of coolant scenarios (small, intermediate, and large break loss of coolant accidents), a primary-to-secondary leak, and a parametric study (natural circulation test) aimed at characterizing the VVER system at reduced mass inventory conditions. The comparative analysis, presented in the paper, regards the large break loss of coolant accident experiment. Four participants from three different institutions were involved in the benchmark and applied their own models and set up for four different thermal-hydraulic system codes. The benchmark demonstrated the performances of such codes in predicting phenomena relevant for safety on the basis of fixed criteria.

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.

Numerical Simulation of the Transient Phase Change Heat Transfer in a Pressurized Water Reactor Core During a Loss-of-Coolant Accident

2011 ◽  
Author(s):  
Soo W. Jo ◽  
Yong K. Lee ◽  
Jong C. Jo
Keyword(s):  
Heat Transfer ◽  
Numerical Simulation ◽  
Transient Heat Transfer ◽  
Coolant Temperature ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
The Core ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

Temperature of pressurized water reactor (PWR) core is a key parameter used widely for judging the initiation of emergency operating procedures and severe accident management. Since direct measurement of the fuel cladding surface temperature using thermocouples is not practicable currently, the coolant temperature at the core exit locations is monitored instead. Several experimental researches showed that the CET rise during a loss of coolant accident (LOCA) and its magnitudes were always lower than the actual fuel rod cladding temperature at the same time. In this regard, a theoretical analysis of the transient heat transfer of coolant flow in a PWR core is needed to confirm the findings from the previous experimental works. This paper addresses numerical simulation of the transient boiling-induced multiphase flow through a simplified PWR core model during a LOCA by a commercial computational fluid dynamics (CFD) code. The calculated results are discussed to understand the transient heat transfer mechanism in the core and to provide useful technical information for reactor design and operation.

scholarly journals Major Achievements and Prospect of the ATLAS Integral Effect Tests

2012 ◽  
Vol 2012 ◽  
pp. 1-18 ◽  
Author(s):  
Ki-Yong Choi ◽  
Yeon-Sik Kim ◽  
Chul-Hwa Song ◽  
Won-Pil Baek
Keyword(s):  
Large Scale ◽  
Test Facility ◽  
Pressurized Water Reactor ◽  
Hydraulic Test ◽  
Validation Data ◽  
Pressurized Water ◽  
Loss Of Coolant Accident ◽  
Test Loop ◽  
Loss Of Coolant ◽  
Integral Effect

A large-scale thermal-hydraulic integral effect test facility, ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation), has been operated by KAERI. The reference plant of ATLAS is the APR1400 (Advanced Power Reactor, 1400 MWe). Since 2007, an extensive series of experimental works were successfully carried out, including large break loss of coolant accident tests, small break loss of coolant accident tests at various break locations, steam generator tube rupture tests, feed line break tests, and steam line break tests. These tests contributed toward an understanding of the unique thermal-hydraulic behavior, resolving the safety-related concerns and providing validation data for evaluation of the safety analysis codes and methodology for the advanced pressurized water reactor, APR1400. Major discoveries and lessons found in the past integral effect tests are summarized in this paper. As the demand for integral effect tests is on the rise due to the active national nuclear R&D program in Korea, the future prospects of the application of the ATLAS facility are also discussed.

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.

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