Simulated AP600 Response to Small-Break Loss-of-Coolant-Accident and Non-Loss-of-Coolant-Accident Events: Analysis of SPES-2 Integral Test Results

1998 ◽  
Vol 122 (1) ◽  
pp. 19-42 ◽  
Author(s):  
M. T. Friend ◽  
R. F. Wright ◽  
R. Hundal ◽  
L. E. Hochreiter ◽  
M. Ogrins
Keyword(s):  
Test Results ◽  
Integral Test ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

Analyses of ACME integral test results on CAP1400 small-break loss-of-coolant-accident transient

2016 ◽  
Vol 88 ◽  
pp. 375-397 ◽  
Author(s):  
Yu Quan Li ◽  
Hua Jian Chang ◽  
Zi Shen Ye ◽  
Fang Fang Fang ◽  
Yan Shi ◽  
...  
Keyword(s):  
Test Results ◽  
Integral Test ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

Simulating Analysis for Small Break LOCA Test Facilities by Applying the Volume Scaling With Reduced Pressure and Reduced Height

2001 ◽  
Author(s):  
S. K. Moussavian ◽  
M. A. Salehi
Keyword(s):  
Experimental Data ◽  
Pressure Drop ◽  
Test Facility ◽  
Integral Test ◽  
Reduced Pressure ◽  
Simple Loop ◽  
Loss Of Coolant Accident ◽  
Reduced Height ◽  
Simulating Analysis ◽  
Loss Of Coolant

Abstract In this paper first we briefly define the different scaling schemes and scaling logic in which we use these schemes to simulate the Small-Break Loss Of Coolant Accident (SB-LOCA) in test facilities. The simple loop of the test facility is considered and the mass, momentum and energy conservation equations are used for the derivation of the scaling model. The variations of mass flow rate, pressure drop and the void fraction in the loop as functions of the time scale or the inventories are obtained. Finally, the calculated results from the simulating schemes are compared with the experimental data previously obtained in an integral test facility.

Integral Test and Engineering Analysis of Coolant Depletion during a Large-Break Loss-of-Coolant Accident

2005 ◽  
Vol 149 (2) ◽  
pp. 200-216 ◽  
Author(s):  
Yong Soo Kim ◽  
Chang Hwan Park ◽  
Byoung Uhn Bae ◽  
Goon Cherl Park ◽  
Kune Yull Suh ◽  
...  
Keyword(s):  
Engineering Analysis ◽  
Integral Test ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

Experimental Demonstration of Safety of AHWR during Stagnation Channel Break Condition in an Integral Test Loop

2018 ◽  
Vol 4 (2) ◽  
Author(s):  
Mukesh Kumar ◽  
A. K. Nayak ◽  
Sumit V. Prasad ◽  
P. K. Verma ◽  
R. K. Singh ◽  
...  
Keyword(s):  
Nuclear Reactor ◽  
Low Flow ◽  
Experimental Demonstration ◽  
Integral Test ◽  
Water Reactor ◽  
Loss Of Coolant Accident ◽  
Test Loop ◽  
Loss Of Coolant ◽  
Channel Conditions ◽  
Channel Type

Detection of loss of coolant accident (LOCA) and generation of reactor trip signal for shutting down the reactor is very important for safety of a nuclear reactor. Large break LOCA (LBLOCA) is a typical design basis accident in all reactors and has attracted attention of the reactor designers. However, studies reveal that small break loss of coolant accident (SBLOCA) can be more severe as it is difficult to detect with conventional methods to generate reactor trip. SBLOCA in channel-type reactors is essential to consider as it may create stagnation channel conditions in the reactor coolant channel, which may lead to fuel failure, if the reactor is not tripped. Advanced heavy water reactor (AHWR) is a channel-type boiling water reactor, which may experience stagnation channel conditions in case of SBLOCA in feeder pipes. For initiating the trip signals and safe shut down of the reactor in such cases, a novel system comprising of acoustic-based sensors is incorporated in the reactor design. The system detects the peculiar sound of the steam leaked from the main heat transport system (MHTS) and generates reactor trip signal. The experimental demonstration of such new system is essential before its introduction in the reactor. The experimental demonstration of the stagnation channel break, its detection by acoustic-based sensors system, and reactor trip followed by generation of reactor trip signal was performed and presented in the paper. The experiment showed that the trip signal for AHWR can be generated within 5 s with acoustic sensor and 2 s by low flow signal and reactor trip can be ensured in 7 s following a LOCA.

Large-Break Loss-of-Coolant Accident Testing and Simulation for 200-MWe Simplified Boiling Water Reactor

2002 ◽  
Author(s):  
S. T. Revankar ◽  
Y. Xu ◽  
H. J. Yoon ◽  
M. Ishii
Keyword(s):  
Safety System ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Integral Test ◽  
Water Reactor ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Main Steam ◽  
Core Cooling ◽  
Safety Systems

The performance of the safety systems of a new design of the 200-MWe simplified boiling water reactor during a large-break, loss-of-coolant accident transient was investigated through code modeling and integral system testing. The accident considered was a break in the main steam line which is the major design basis accident. RELAP5/MOD3 best estimate reactor thermalhydraulic code was used and its applicability to the reactor safety system evaluation was examined. The integral tests were perfomed to assess the safety systems and the response of the emergency core cooling systems to accident conditions in a scaled facility called PUMA. The details of the safety system behavior are presented. The integral test simulations examined code applicability at the scaled facility level as well as prototype key safety system performance.

IIST Passive Core Cooling on Pressurizer Top Break

2003 ◽  
Author(s):  
Tay-Jian Liu ◽  
Chien-Hsiung Lee
Keyword(s):  
Cooling System ◽  
Test Results ◽  
Pressurized Water Reactor ◽  
Relief Valve ◽  
Pressurized Water ◽  
Hydraulic Behavior ◽  
Loss Of Coolant Accident ◽  
Accident Management ◽  
Loss Of Coolant ◽  
Core Cooling

Two experiments for small-break loss-of-coolant accident on pressurizer top were conducted at the INER Integral System Test (IIST) facility to investigate thermal-hydraulic behavior of a passive core cooling system (PCCS) in a Westinghouse pressurized water reactor (PWR). The test results are compared with the previous IIST tests under the same initial and boundary conditions for a power-operated relief valve (PORV) stuck-open incident. The objectives of this study are to understand of the key thermal-hydraulic phenomena associated with PCCS and to compare the effectiveness of accident management with or without PCCS. The break sizes were scaled down based on one and all three fully-opened PORVs. This paper identified the key phenomena commonly observed and the phenomena unique to a PWR with PCCS.

Direct ECC Bypass Phenomena During LBLOCA Reflood Phase Observed in the MIDAS Test Facility

2002 ◽  
Author(s):  
B. J. Yun ◽  
T. S. Kwon ◽  
D. J. Euh ◽  
I. C. Chu ◽  
C.-H. Song ◽  
...  
Keyword(s):  
Reactor Core ◽  
Flow Distribution ◽  
Test Facility ◽  
Test Results ◽  
Paper Test ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Water Test ◽  
Core Cooling ◽  
Direct Bypass

One of the advanced design features of the APR-1400, direct vessel injection (DVI) system is being considered instead of conventional cold leg injection (CLI) system. It is known that the DVI system greatly enhances the reliability of the emergency core cooling (ECC) system. However, there is still a dispute on its performance in terms of water delivery to the reactor core during the reflood phase of a large-break loss-of-coolant accident (LOCA). Thus, experimental validation is in progress. In this paper, test results of a direct ECC bypass performed in the steam-water test facility called MIDAS (Multi-dimensional Investigation in Downcomer Annulus Simulation) is presented. The test condition is determined, based on the preliminary analysis of TRAC code, by applying the ‘modified linear scaling method’ with the 1/4.93 length scale. From the tests, ECC direct bypass fraction, steam condensation rate and information on the flow distribution in the upper annulus downcomer region is obtained.

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