scholarly journals Assessment of TRACE CCFL Model with SBLOCA Experiment of IIST Facility

2012 ◽  
Vol 2012 ◽  
pp. 1-8 ◽  
Author(s):  
Jung-Hua Yang ◽  
Jong-Rong Wang ◽  
Hao-Tzu Lin ◽  
Chunkuan Shih
Keyword(s):  
Nuclear Energy ◽  
Sensitivity Study ◽  
Test Facility ◽  
Energy Research ◽  
Experiment Data ◽  
Loss Of Coolant Accident ◽  
Study Results ◽  
Loss Of Coolant ◽  
Trace Model ◽  
Good Agreement

In this paper, the TRACE model for IIST facility is developed and verified with the Small Break loss of coolant accident (SBLOCA) experiment of IIST (Institute of Nuclear Energy Research Integral System Test) facility. By using the Wallis and Kutateladze correlations of countercurrent flow limitation (CCFL) model, the TRACE analyses results, such as break flow rate, primary pressure, and the temperature of cold-leg and hot-leg, are consistent with the IIST data. The results show the Kutateladze correlation of CCFL model can well predict the SBLOCA behavior and present good agreement with IIST experiment data in this paper. Besides, the sensitivity study results of Kutateladze correlation in CCFL model are verified and compared with the IIST data.

Development of a Thermal-Hydraulic Code for Chinese Helium-Cooled Ceramic Breeder TBM Cooling System

2013 ◽  
Author(s):  
Jie Wang ◽  
Guanghui Su ◽  
Wenxi Tian ◽  
Suizheng Qiu
Keyword(s):  
Cooling System ◽  
Cooling Systems ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Friction Models ◽  
Cooling Loop ◽  
Thermal Hydraulic Analysis ◽  
Coolant System ◽  
Relap5 Code ◽  
Good Agreement

Helium was chosen as the coolant for divertor cooling loop, Korea helium cooled solid breeder TBM, European helium cooled pebble bed TBM and Chinese helium cooled ceramic breeder TBM. The thermal-hydraulic analysis for the divertor cooling loop and the TBM cooling systems were carried out by RELAP5 and MELOCR codes, which were developed for transient simulation of light water reactor coolant system during postulated accidents. In this study, a thermal-hydraulic and safety analysis code (TSACO) for helium cooling system was developed and calculation of the Chinese HCCB TBM cooling system was presented. Heat transfer and flow friction models for helium were added in the code. First, the code was verified by comparing with the RELAP5 code with the same initial and boundary conditions. The first wall temperature, pressure drop and inlet/outlet temperatures were studied and a good agreement was obtained, then ex-vessel loss of coolant accident for Chinese HCCB-TBM cooling system was investigated using TSACO. The results show that the TBM can be cooled efficiently and the TCWS pressure stays within the design limits with a large margin.

Test and Evaluation of a Filtering System for Retention of Fibres in a Coolant Flow After Loss-of-Coolant Accident

2012 ◽  
Author(s):  
T. Gocht ◽  
W. Kästner ◽  
A. Kratzsch ◽  
M. Strasser
Keyword(s):  
Cooling System ◽  
Reactor Core ◽  
Heat Removal ◽  
Test Facility ◽  
Experimental Investigations ◽  
Time Interval ◽  
Insulation Material ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Core Cooling

In case of an accident the safe heat removal from the reactor core with the installed emergency core cooling system (ECCS) is one of the main features in reactor safety. During a loss-of-coolant accident (LOCA) the release of insulation material fragments in the reactor containment can lead to malfunctions of ECCS. Therefore, the retention of particles by strainers or filtering systems in the ECCS is one of the major tasks. The aim of the presented experimental investigations was the evaluation of a filtering system for the retention of fiber-shaped particles in a fluid flow. The filtering system consists of a filter case with a special lamellar filter unit. The tests were carried out at a test facility with filtering units of different mesh sizes. Insulation material (mineral rock wool) was fragmented to fiber-shaped particles. To simulate the distribution of particle concentration at real plants with large volumes the material was divided into single portions and introduced into the loop with a defined time interval. Material was transported to the filter by the fluid and agglomerated there. The assessment of functionality of the filtering system was made by differential pressure between inlet and outlet of the filtering system and by mass of penetrated particles. It can be concluded that for the tested filtering system no penetration of insulation particles occurred.

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.

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.

Comparative Study of Loss-of-Coolant Accident Using MAAP4.03 and RELAP5/MOD3.2.2

2002 ◽  
Author(s):  
Chang Hwan Park ◽  
Doo Yong Lee ◽  
Ik Jeong ◽  
Un Chul Lee ◽  
Kune Y. Suh ◽  
...  
Keyword(s):  
Flow Model ◽  
Correction Term ◽  
Sensitivity Study ◽  
Severe Accident ◽  
Primary System ◽  
Two Phase ◽  
Flow Rates ◽  
Flow Models ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

Analysis was performed for a large-break loss-of-coolant accident (LOCA) in the APR1400 (Advanced Power Reactor 1400 MWe) with the thermal-hydraulic analysis code RELAP5/ MOD3.2.2 and the severe accident analysis code MAAP4.03. The two codes predicted different sequences for essentially the same initiating condition. As for the break flow and the emergency core cooling (ECC) flow rates, MAAP4.03 predicted considerably higher values in the initial stage than RELAP5/ MOD3.2.2. It was considered that the differing break flow and ECC flow rates would cause the LOCA sequences to deviate from one another between the two codes. Hence, the break flow model in MAAP4.03 was modified with partly implementing the two-phase homogeneous critical flow model and adopting a correction term. The ECC flow model in MAAP4.03 was also varied by changing the hardwired friction factor through the sensitivity study. The modified break flow and ECC flow models yielded more consistent calculational results between RELAP5/MOD3.2.2 and MAAP4.03. It was, however, found that the resultant effect is rather limited unless more mechanistic treatments are done for the primary system in MAAP4.03.

Safety Assessment of Reactor Pressure Vessel Integrity for Loss of Coolant Accident Conditions

2011 ◽  
Vol 134 (1) ◽  
Author(s):  
Dieter Beukelmann ◽  
Wenfeng Guo ◽  
Wieland Holzer ◽  
Robert Kauer ◽  
Wolfgang Münch ◽  
...  
Keyword(s):  
Finite Element ◽  
Pressure Vessel ◽  
Safety Margin ◽  
Reactor Pressure Vessel ◽  
Test Facility ◽  
Temperature Distributions ◽  
Loss Of Coolant Accident ◽  
System Pressure ◽  
Loss Of Coolant ◽  
Reactor Pressure

One of the critical issues for reactor pressure vessel (RPV) structural integrity is related to the pressurized thermal shock (PTS) event. Therefore, within the framework of safety assessments special emphasis is given to the effect of PTS-loadings caused by the nonuniform azimuthal temperature distribution due to cold water plumes or stripes during emergency coolant injection. This paper describes the method used to predict the thermal mechanic boundary conditions (system pressure and wall temperature). Using a system code the pressure and global temperature distributions were calculated, systematically varying the leak size and the location of the coolant water injection. Spatial and temporal temperature distributions in the main circulation pipes and at the RPV wall were predicted by mixing analyses with a computational fluid dynamics (CFD) code. The model used for these calculations was validated by post-test calculations of a UPTF (upper plenum test facility) experiment simulating cold leg injection during a small break loss of coolant accident (LOCA). Comparison with measured temperatures showed that the modeling used is suitable to obtain enveloping results. Fracture mechanics analyses were carried out for circumferential flaw sizes in the weld joint near the core region and between the RPV shell and the flange, as well as for axial flaws in the nozzle corner. Stress intensity factors KI were calculated numerically using the finite element program ansys and analytically on the basis of weight and polynomial influence functions using stresses obtained from elastic finite element analyses. Benchmark tests revealed good agreement between the results from numerical and analytical calculations. For all regions of the RPV investigated and the most severe transients it was demonstrated that a large safety margin against brittle crack initiation exists and brittle fracture of the RPV can be excluded.

Validation of Fluid-Structure Interaction Calculations in a Large-Break Loss of Coolant Accident

2008 ◽  
Author(s):  
Antti Timperi ◽  
Timo Pa¨ttikangas ◽  
Ismo Karppinen ◽  
Ville Lestinen ◽  
Jukka Ka¨hko¨nen ◽  
...  
Keyword(s):  
Structural Analysis ◽  
Fluid Structure Interaction ◽  
Acoustic Medium ◽  
Fluid Structure ◽  
Structure Interaction ◽  
Loss Of Coolant Accident ◽  
Phase Calculation ◽  
Loss Of Coolant ◽  
Pipe Break ◽  
Good Agreement

In the hypothetical Large-Break Loss Of Coolant Accident (LBLOCA), rapid depressurization of the reactor primary circuit causes loads on the reactor internals. This paper presents numerical simulations of a HDR experiment, where LBLOCA of a pressurized water reactor due to a sudden pipe break in the primary loop was studied. In the experiment, Fluid-Structure Interaction (FSI) phenomena caused by the flexibility of the core barrel were studied in particular. Star-CD Computational Fluid Dynamics (CFD) code and ABAQUS structural analysis code were used for three-dimensional calculations. The MpCCI code was used for two-way coupling of the CFD and structural analysis codes in order to take FSI into account. Two-way FSI calculation was also performed with ABAQUS only by modeling water as an acoustic medium. Pressure boundary condition at the pipe break was evaluated with the system code APROS as a two-phase calculation. Comparisons with the experiment were made for fluid pressures and break mass flow as well as for structural displacements and strains. Fairly good agreement was found between the experiment and simulation when coupling of the CFD and structural analysis codes was used. For the acoustic calculation, the results showed good agreement in the early phase of the simulation. In the late phase, structural loads were over-predicted by the acoustic calculation due to the effect of bulk flow of water which is not included in the acoustic model.

Small Break LOCA Analysis of ACR-700 NPP

2006 ◽  
Author(s):  
Limin Zheng ◽  
Sen Shen ◽  
David Wright
Keyword(s):  
Nuclear Power ◽  
Sensitivity Study ◽  
Hydraulic Analysis ◽  
Reactor Outlet ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Subsequent Loss ◽  
Safety Systems ◽  
Limiting Case ◽  
Thermal Hydraulic Analysis

A small break loss of coolant accident (SB-LOCA) analysis to assess a preliminary conceptual design of the ACR-700 PHWR nuclear power plant (NPP) developed by AECL has been performed with CATHENA MOD 3.5d, a PHWR system thermal-hydraulic analysis code. The limiting break size has been found by performing a sensitivity study for three different break locations [i.e. reactor inlet header (RIH), HTS pump suction (PS) pipe and reactor outlet head (ROH)] under the limiting case (i.e. SB-LOCA with subsequent loss of class IV power with all safety systems available). The analysis results indicate that the SB-LOCA acceptance criteria are satisfied.

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