Investigation of PAR Behavior in the REKO-4 Test Facility

2012 ◽  
Author(s):  
B. Simon ◽  
E.-A. Reinecke ◽  
M. Klauck ◽  
D. Heidelberg ◽  
H.-J. Allelein
Keyword(s):  
Experimental Data ◽  
Mitigation Strategy ◽  
Hydrogen Combustion ◽  
Test Facility ◽  
Reactor Accident ◽  
Code Validation ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Reactor Building

Passive auto-catalytic recombiners (PARs) play a key role in the hydrogen mitigation strategy of European LWRs. In order to avoid possible threats related to hydrogen combustion, PARs are installed to remove hydrogen released during a loss-of-coolant accident. The possible impact of hydrogen explosions became evident during the reactor accident in Fukushima (Japan) in March 2011, where leaked hydrogen ignited and largely destroyed the upper part of the reactor building. The mitigation strategy is based and verified by computational accident assessments. Code validation against experimental data is vital in order to achieve reliable results.

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 Experimental and Numerical Results of LIFUS5/Mod3 Series E Test on In-Box LOCA Transient for WCLL-BB

Energies ◽  
2021 ◽  
Vol 14 (24) ◽  
pp. 8527
Author(s):  
Marica Eboli ◽  
Francesco Galleni ◽  
Nicola Forgione ◽  
Nicolò Badodi ◽  
Antonio Cammi ◽  
...  
Keyword(s):  
Experimental Data ◽  
Safety Concern ◽  
Computational Cost ◽  
Water System ◽  
Test Facility ◽  
Relevant Parameter ◽  
Research Activities ◽  
Loss Of Coolant Accident ◽  
Parameter Ranges ◽  
Physical Phenomena

The in-box LOCA (Loss of Coolant Accident) represents a major safety concern to be addressed in the design of the WCLL-BB (water-cooled lead-lithium breeding blanket). Research activities are ongoing to master the phenomena and processes that occur during the postulated accident, to enhance the predictive capability and reliability of numerical tools, and to validate computer models, codes, and procedures for their applications. Following these objectives, ENEA designed and built the new separate effects test facility LIFUS5/Mod3. Two experimental campaigns (Series D and Series E) were executed by injecting water at high pressure into a pool of PbLi in WCLL-BB-relevant parameter ranges. The obtained experimental data were used to check the capabilities of the RELAP5 system code to reproduce the pressure transient of a water system, to validate the chemical model of PbLi/water reactions implemented in the modified version of SIMMER codes for fusion application, to investigate the dynamic effects of energy release on the structures, and to provide relevant feedback for the follow-up experimental campaigns. This work presents the experimental data and the numerical simulations of Test E4.1. The results of the test are presented and critically discussed. The code simulations highlight that SIMMER code is able to reproduce the phenomena connected to PbLi/water interaction, and the relevant test parameters are in agreement with the acquired experimental signals. Moreover, the results obtained by the first approach to SIMMER-RELAP5 code-coupling demonstrate its capability of and strength for predicting the transient scenario in complex geometries, considering multiple physical phenomena and minimizing the computational cost.

Investigation of Chemical Effects on Emergency Core Cooling Filtration Head Loss After Loss of Coolant Accident

2009 ◽  
Author(s):  
Jong Woon Park ◽  
Byung Gi Park ◽  
Chang Hyun Kim
Keyword(s):  
Calcium Silicate ◽  
Sodium Phosphate ◽  
Building Materials ◽  
Head Loss ◽  
Aluminum Surface ◽  
Loss Of Coolant Accident ◽  
Filter Screen ◽  
Loss Of Coolant ◽  
Reactor Building ◽  
Core Cooling

Integrated tests of head loss through an emergency core cooling filter screen are conducted, simulating reactor building environmental conditions for thirty days after a loss of coolant accident. A test apparatus with five individual loops each of whose chamber is established to test chemical product formation and measure the head loss through a sample filter. The screen area at each chamber is 78.54cm2 and reactor building materials can be scaled down according to specific plant condition. A series of tests have been performed to investigate the effects of reactor building spray, existence of calcium-silicate with tri-sodium phosphate (TSP), and composition of materials. The results showed that head loss across the chemical bed with even a small amount of calcium-silicate insulation instantaneously increased as soon as TSP was added to the test solution. Also, the head loss across the filter screen is strongly affected by spray duration and the head loss increase is rapid at the early stage, because of high dissolution and precipitation of aluminum and zinc. After passivation of aluminum and zinc by corrosion, the head loss increase is much slowed down and is mainly induced by materials such as calcium, silicon, and magnesium leached from NUKON™ and concrete. Furthermore, it is newly found that the spay buffer agent, tri-sodium phosphate, to form protective coating on the aluminum surface and reduce aluminum leaching is not effective for a large amount of aluminum and a long spray.

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.

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.

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.

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.

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.

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