scholarly journals Advanced neutron source reactor thermal-hydraulic test loop facility description

1994 ◽  
Author(s):  
D.K. Felde ◽  
G. Farquharson ◽  
J.H. Hardy ◽  
J.F. King ◽  
M.T. McFee ◽  
...  
Keyword(s):  
Neutron Source ◽  
Hydraulic Test ◽  
Test Loop

Thermo-hydraulic test of the moderator cell of LH2 cold neutron source at BNC

2000 ◽  
Vol 276-278 ◽  
pp. 214-215 ◽  
Author(s):  
T. Grósz ◽  
T. Hargitai ◽  
V.A. Mityukhlyaev ◽  
L. Rosta ◽  
A.P. Serebrov ◽  
...  
Keyword(s):  
Neutron Source ◽  
Cold Neutron ◽  
Hydraulic Test ◽  
Cold Neutron Source ◽  
Moderator Cell

Experimental Study on the Safety Injection Pump (SIP) Failure Accompanied by the Steam Generator Tube Rupture (SGTR)

2018 ◽  
Author(s):  
Yusun Park ◽  
Byoung Uhn Bae ◽  
Jongrok Kim ◽  
Jae Bong Lee ◽  
Hae Min Park ◽  
...  
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Water Levels ◽  
Experimental Result ◽  
Hydraulic Test ◽  
Steam Generator Tube ◽  
Injection Pump ◽  
Management Procedure ◽  
Test Loop ◽  
Integral Effect

The integral effect test to simulate the safety injection pump (SIP) failure accompanied by a steam generator tube rupture (SGTR), named a SGTR-SIP-01 test, was performed to investigate the thermal hydraulic phenomena during a multiple failure accident. In this study, a thermal-hydraulic integral effect facility, ATLAS (Advanced Thermal-hydraulic test Loop for Accident Simulation) was utilized to simulate thermal hydraulic phenomenon which can be occurred in the nuclear power plant, as realistically as possible. In this SGTR-SIP-01 test, a rupture of five steam generator u-tubes on the steam generator hot-side was simulated. Due to the initiation of SGTR, a reactor was tripped by high steam generator level (HSGL) signal. During the transient simulation for SGTR-SIP-01 test, major thermal-hydraulic parameters such as the system pressures, the collapsed water levels, the flows in the primary loops, and the fluid temperatures, were measured and analyzed. Through this experimental result, insights about the accident management procedure can be provided in the case of the multiple failure accident, such as a SGTR accident with a total failure of SIPs. In addition to that, for improvement of the system code which are now on developing such SPACE or MARS-KS code, this test data can be utilized for validation and verification work.

Westinghouse Advanced Loop Tester (WALT) Update

2008 ◽  
Author(s):  
G. Wang ◽  
W. A. Byers ◽  
M. Y. Young ◽  
Z. E. Karoutas
Keyword(s):  
Nuclear Power ◽  
Hot Spot ◽  
Test Facility ◽  
Hydraulic Test ◽  
Pressurized Water ◽  
System Pressure ◽  
Test Loop ◽  
Zinc Addition ◽  
Water Reactors ◽  
The Impact

In order to understand crud formation on the fuel rod cladding surfaces of pressurized water reactors (PWRs), a crud Thermal-Hydraulic test facility referred to as the Westinghouse Advanced Loop Tester (WALT) was built at the Westinghouse Science and Technology Department Laboratories in October 2005. Since then, a number of updates have been made and are described here. These updates include heater rod improvements, system pressure stabilization, and more effective protection systems. After these updates were made, the WALT system has been operated with higher stability and fewer failures. In this test loop, crud can be deposited on the heater rod surface and the character of the crud is similar to what has been observed in the PWRs. In addition, chemistry in the WALT loop can be varied to study its impact on crud morphology and associated parameters. The WALT loop has been successful in generating crud and measuring its thermal impact as a function of crud thickness. Currently, this test facility is supporting an Electric Power Research Institute (EPRI) program to assess the impact of zinc addition to PWR reactor coolant. Meanwhile, the WALT system is also being utilized by Westinghouse to perform dry-out and hot spot tests. These tests support the industry goal of 0 fuel failures by 2010 set by Institute of Nuclear Power Operations (INPO). Another major goal of the Westinghouse tests is to gain a better understanding of unexpected changes in core power distributions in operating reactors known as crud induced power shifts (CIPS) or axial offset anomalies (AOA).

The Applicability of NFI-1 DNB Correlation and Fluid-to-Fluid Similarities to Freon DNB Test

2008 ◽  
Author(s):  
Shumpei Kakinoki ◽  
Keizo Matsuura ◽  
Kenichi Kitagawa ◽  
Isao Kataoka
Keyword(s):  
Heat Flux ◽  
Equivalent Condition ◽  
High Heat Flux ◽  
Diffusivity Coefficient ◽  
Test Results ◽  
Hydraulic Test ◽  
Rod Bundle ◽  
Test Loop ◽  
R 134A ◽  
Subchannel Analysis

Freon thermal hydraulic test is expected to be one of the workable methods to develop high thermal hydraulic performance PWR fuel. That is, high pressure water and high heat flux condition in PWR core can be substituted with lower pressure Freon and lower heat flux by applying appropriate fluid-to-fluid similarity and modeling parameters. Freon DNB tests and mixing tests were carried out against a 4×4 rod bundle configuration where R-134A flowed vertically upwardly. The tests were carried out at Freon thermal hydraulic test loop in Korea Atomic Energy Research Institute (KAERI). The spacer grid used in these tests was modeled on that of conventional PWR fuel, that is, square lattice grid with split type mixing vanes. Diameter of heater rod simulating PWR fuel rod is about 10.7mm and heating length is about 2000 mm. Freon mixing tests were carried out to estimate Turbulence Diffusivity Coefficient (TDC), which was normally used in conventional thermal hydraulic design of nuclear reactor. Freon CHF test results showed that parametric trends agreed with those of existing CHF data. To predict CHF of 4×4 rod bundle, subchannel analysis code Modified COBRA-3C and NFI-1 DNB correlation were applied. TDC value used in subchannel analysis was determined by fitting Freon mixing test data. NFI-1 DNB correlation was developed for predicting DNB heat flux in rod bundle configuration by using water CHF test results at HTRF test loop at Columbia University. The design of spacer grids used in KAERI Freon DNB test was similar to that used in water CHF test at HTRF. Water equivalent flow condition of this R-134A test was estimated using fluid-to-fluid similarities. NFI-1 DNB correlation was applied to this water equivalent condition to estimate water equivalent DNB heat flux. Then R-134A equivalent DNB heat flux was estimated reversely, and compared to Freon DNB test result. The test results were predicted well and applicability of NFI-1 DNB correlation and fluid-to-fluid similarities in 4×4 rod bundle is discussed.

Effects of Cavitation Damage on the Hydraulic Characteristics of a Gate Valve

2013 ◽  
Author(s):  
Vatsal Trivedi ◽  
Amjad Farah ◽  
Glenn Harvel
Keyword(s):  
Gate Valve ◽  
Equation Model ◽  
Hydraulic Loss ◽  
Cfd Model ◽  
Vapor Bubbles ◽  
Hydraulic Test ◽  
Cavitation Damage ◽  
Test Loop ◽  
Computational Fluid Dynamics Cfd ◽  
Loss Coefficients

This paper explores the effects of cavitation damage on the hydraulic performance of a gate valve. Cavitation is the phenomenon in which a fluid evaporates to form vapor bubbles which then subsequently implode. A Computational Fluid Dynamics (CFD) model of water flowing through a partially opened 1 inch (2.54 cm diameter) gate valve was developed using Unigraphics NX 7.5. NX 7.5 utilizes a k-ε turbulent flow 2-equation model. The low pressure and high velocity regions were identified in the CFD model to predict the location of the cavitation site in the valve. In order to simulate the effects of ageing due to cavitation, a gate valve was mechanically damaged in different stages. Damaged and undamaged valves were then inserted in a hydraulic test loop to observe the flow rate and pressure drop across the valves. The valves’ hydraulic loss coefficient was then calculated and compared. The cavitation index of the damaged and the undamaged gate valves was also calculated to predict the likelihood of cavitation. The results show that the aged valve had lower loss coefficients at the corresponding opening positions compared to the ones for the non-aged valve. Also, it was observed that the aged gate valve had less likelihood of cavitation compared to the non-aged gate valve.

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.

Integral Effect Test for a Inadvertent Opening of POSRV Using the ATLAS Facility

2010 ◽  
Author(s):  
Seok Cho ◽  
Ki-Yong Choi ◽  
Hyun-Sik Park ◽  
Kyoung-Ho Kang ◽  
Yeon-Sik Kim ◽  
...  
Keyword(s):  
Nuclear Industry ◽  
Test Facility ◽  
System Response ◽  
Integral Test ◽  
Hydraulic Test ◽  
Pressurized Water ◽  
Reduced Height ◽  
Test Loop ◽  
Integral Effect ◽  
Effect Test

A thermal-hydraulic integral effect test facility for advanced pressurized reactors (PWRs), ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation), has been operated by KAERI (Korea Atomic Energy Research Institute). The reference plant of the ATLAS is a 1400 MWe-class evolutionary pressurized water reactor (PWR), the APR1400 (Advanced Power Reactor 1,400 MWe), which was developed by the Korean industry. The ATLAS has a 1/2 reduced height and a 1/288 volume scaled integral test facility with respect to the APR1400. It has a maximum power capacity of 10% of the scaled nominal core power, and it can simulate full pressure and temperature conditions of the APR1400. The ATLAS could be used to provide experimental data on design-basis accidents including the reflood phase of a large break loss of coolant accident (LBLOCA), small break LOCA (SBLOCA) scenarios including the DVI line and cold leg breaks, a steam generator tube rupture, a main steam line break, a feed line break, etc. An inadvertent opening of POSRV test (SB-POSRV-02) was carried out as one of the SBLOCA spectra. The main objectives of this experimental test were not only to provide physical insight into the system response of the APR1400 reactor during a transient situation but also to present integral effect data for the validation of the SPACE (Safety and Performance Analysis Computer Code), which is now under development by the Korean nuclear industry.

Thermal Conductivity Measurements for Simulated PWR Crud

2013 ◽  
Author(s):  
G. Wang ◽  
W. A. Byers ◽  
M. Y. Young ◽  
J. Deshon ◽  
Z. Karoutas ◽  
...  
Keyword(s):  
Thermal Conductivity ◽  
Regime Theory ◽  
Conductivity Measurements ◽  
Pressurized Water Reactor ◽  
Test Program ◽  
Hydraulic Test ◽  
Pressurized Water ◽  
Fuel Rod ◽  
Test Loop ◽  
Hydraulic Conditions

This paper describes a laboratory test program to measure the thermal conductivity of corrosion product deposits on the surface of a Pressurized Water Reactor (PWR) fuel rod under a variety of thermal hydraulic conditions. This thermal conductivity information is necessary to allow more accurate predictions of fuel rod surface temperatures in the presence of fuel deposits, commonly known as crud. In this paper, a four regime theory and methodology are proposed and utilized for crud thermal conductivity measurements and calculations. The relevant measurements were performed at the Westinghouse Advanced Loop Tester (WALT) facility, which is a single rod crud thermal-hydraulic test loop built at the Westinghouse Science and Technology Center (STC). This facility is described and then selected experiments and calculated results of this study are presented and discussed.

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