Experimental Investigation on Gas Mixing and Stratification in Containment Influenced by External Cooling

2020 ◽  
Author(s):  
Ying Li ◽  
Wei Xue ◽  
Xuewu Cao ◽  
Lili Tong
Keyword(s):  
Large Scale ◽  
Initial Pressure ◽  
Stable Stratification ◽  
Test Facility ◽  
Gas Mixing ◽  
Additional Time ◽  
Integral Test ◽  
External Cooling ◽  
Comprehensive Test ◽  
Accident Conditions

Abstract The distribution of hydrogen inside the containment is a key issue in assessing the evolution of the postulated accident. For safety analysis and codes validation purposes, a large scale comprehensive test facility has been built to investigate the containment thermal-hydraulic characteristics under accident conditions. In this paper, a test was performed to experimentally investigate the distribution of the hydrogen inside the containment and the influence of the external cooling on gas mixing and stratification. The paper presents the experimental results of the integral test performed in this facility. During the experiments, helium was used to simulate hydrogen. Helium and steam are released together and allowed to take additional time to form a relatively stable stratification, then followed by external cooling. The initial pressure of the experiments is around 0.1MPa(a) and the initial Froude number is around 333. The results showed that a helium-enriched stratification emerged in the upper containment due to the density difference after the injection. External cooling caused condensation and intense convective flow. As a result, an overall increase in helium concentration was observed with a decrease in concentration gradient.

Break Size Effects on CET Response in an Upper Head SBLOCA Transient

2012 ◽  
Author(s):  
S. Gallardo ◽  
A. Querol ◽  
G. Verdú
Keyword(s):  
Pressure Vessel ◽  
Large Scale ◽  
Injection System ◽  
Test Facility ◽  
Energy Agency ◽  
Pressurized Water ◽  
The Core ◽  
Upward Direction ◽  
Water Reactors ◽  
Accident Conditions

In the transients produced during Small Break Loss-Of-Coolant Accidents (SBLOCA), the maximum Peak Cladding Temperature (PCT) in the core could suffer rapid excursions which might strongly affect the core integrity. Most Pressurized Water Reactors (PWR) have Core Exit Thermocouples (CETs) to detect core overheating by considering that superheated steam flows in the upward direction when core uncovery occurs during SBLOCAs. Operators may start Accident Management (AM) actions to mitigate such accident conditions when the CET temperature exceeds a certain value. However, in a Vessel Upper Head SBLOCA, a significant delay in time and temperature rise of CETs from core heat-up can be produced. This work is developed in the frame of OECD/NEA ROSA Project Test 6-1 (SB-PV-9 in JAEA) handled in the Large Scale Test Facility (LSTF) of the Japan Atomic Energy Agency (JAEA). Test 6-1 simulated a PWR pressure vessel Upper-Head SBLOCA with a break size equivalent to 1.9% of the cold leg break under the assumption of total failure of High Pressure Injection System (HPIS). The paper shows several analyses about the geometry variables (size, location, flow paths and Upper Head nodalization) which can influence on the pressure vessel Upper Head SBLOCA model performed using the thermal-hydraulic code TRACE5.

scholarly journals Modeling of Kerena Emergency Condenser

2017 ◽  
Vol 38 (4) ◽  
pp. 29-51 ◽  
Author(s):  
Rafał Bryk ◽  
Holger Schmidt ◽  
Thomas Mull ◽  
Thomas Wagner ◽  
Ingo Ganzmann ◽  
...  
Keyword(s):  
Large Scale ◽  
Cold Water ◽  
Test Stand ◽  
Empirical Correlation ◽  
Core Flooding ◽  
Integral Test ◽  
Loss Of Coolant Accident ◽  
Sophisticated Model ◽  
Accident Conditions ◽  
Semi Empirical

Abstract KERENA is an innovative boiling water reactor concept equipped with several passive safety systems. For the experimental verification of performance of the systems and for codes validation, the Integral Test Stand Karlstein (INKA) was built in Karlstein, Germany. The emergency condenser (EC) system transfers heat from the reactor pressure vessel (RPV) to the core flooding pool in case of water level decrease in the RPV. EC is composed of a large number of slightly inclined tubes. During accident conditions, steam enters into the tubes and condenses due to the contact of the tubes with cold water at the secondary side. The condensed water flows then back to the RPV due to gravity. In this paper two approaches for modeling of condensation in slightly inclined tubes are compared and verified against experiments. The first approach is based on the flow regime map. Depending on the regime, heat transfer coefficient is calculated according to specific semi-empirical correlation. The second approach uses a general, fully-empirical correlation. The models are developed with utilization of the object-oriented Modelica language and the open-source OpenModelica environment. The results are compared with data obtained during a large scale integral test, simulating loss of coolant accident performed at Integral Test Stand Karlstein (INKA). The comparison shows a good agreement.Due to the modularity of models, both of them may be used in the future in systems incorporating condensation in horizontal or slightly inclined tubes. Depending on his preferences, the modeller may choose one-equation based approach or more sophisticated model composed of several exchangeable semi-empirical correlations.

Scaling Application of a Hot Leg SBLOCA Scenario to a Nuclear Power Plant

2014 ◽  
Author(s):  
Andrea Querol ◽  
Sergio Gallardo ◽  
Gumersindo Verdú
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Large Scale ◽  
Injection System ◽  
Test Facility ◽  
Experimental Conditions ◽  
Integral Test ◽  
Pressurized Water ◽  
Choked Flow

Several experimental facilities, such as the Large Scale Test Facility (LSTF) of the Japan Atomic Energy Agency (JAEA), have been built to reproduce some accidental scenarios because full-scale testing is usually impossible to perform. One of the objectives of these Integral Test Facilities (ITFs) is to obtain measured data to be compared to simulations in order to test the capability of the thermalhydraulic codes to reproduce experimental conditions. The applicability of these experimental results to a full-size power plant system depends on the scaling criteria adopted. The present paper is focused on the simulation and the scaling of the Test 1-2 in the frame of the OECD/NEA ROSA Project to a Nuclear Power Plant (NPP). This test simulates a hot leg 1% Small Break Loss-Of-Coolant Accident (SBLOCA) in a Pressurized Water Reactor (PWR) under the actuation of High Pressure Injection (HPI) system and Accumulator Injection System (AIS). A scaled-up NPP TRACE5 input has been developed from a LSTF TRACE5 model validated by authors in previous works. The scaled-up model has been developed conserving the power-to-volume scaling ratios of LSTF components, initial and boundary conditions. Lengths and diameters of hot legs have been scaled from LSTF model trying to conserve Froude number. A comparison between both TRACE5 models (LSTF and scaled-up NPP) is performed (system pressures, discharged inventory and collapsed liquid levels). Special TRACE5 models such as Choked flow model and OFFTAKE model have been tested. A 3D VESSEL component has been tested in comparison to 1D TEE component to simulate the hot leg where the SBLOCA is located and varying the break orientation (downwards and upwards). Finally, a sensitivity analysis has been made to determine the effect of the break size in the SBLOCA range.

scholarly journals Integral Test Facility PKL: Experimental PWR Accident Investigation

2012 ◽  
Vol 2012 ◽  
pp. 1-16 ◽  
Author(s):  
Klaus Umminger ◽  
Lars Dennhardt ◽  
Simon Schollenberger ◽  
Bernhard Schoen
Keyword(s):  
Systematic Investigation ◽  
Test Facility ◽  
Pressurized Water Reactor ◽  
Integral Test ◽  
Pressurized Water ◽  
Height Scale ◽  
Water Reactors ◽  
Accident Conditions ◽  
Secondary Side ◽  
Reactor Pressure

Investigations of the thermal-hydraulic behavior of pressurized water reactors under accident conditions have been carried out in the PKL test facility at AREVA NP in Erlangen, Germany for many years. The PKL facility models the entire primary side and significant parts of the secondary side of a pressurized water reactor (PWR) at a height scale of 1 : 1. Volumes, power ratings and mass flows are scaled with a ratio of 1 : 145. The experimental facility consists of 4 primary loops with circulation pumps and steam generators (SGs) arranged symmetrically around the reactor pressure vessel (RPV). The investigations carried out encompass a very broad spectrum from accident scenario simulations with large, medium, and small breaks, over the investigation of shutdown procedures after a wide variety of accidents, to the systematic investigation of complex thermal-hydraulic phenomena. This paper presents a survey of test objectives and programs carried out to date. It also describes the test facility in its present state. Some important results obtained over the years with focus on investigations carried out since the beginning of the international cooperation are exemplarily discussed.

scholarly journals PANDA: A Multipurpose Integral Test Facility for LWR Safety Investigations

2012 ◽  
Vol 2012 ◽  
pp. 1-9 ◽  
Author(s):  
Domenico Paladino ◽  
Jörg Dreier
Keyword(s):  
Large Scale ◽  
Electric Utilities ◽  
Test Facility ◽  
Experimental Investigations ◽  
Primary System ◽  
Integral Test ◽  
Hydraulic Facility ◽  
International Projects ◽  
Regulatory Bodies ◽  
The Eu

The PANDA facility is a large scale, multicompartmental thermal hydraulic facility suited for investigations related to the safety of current and advanced LWRs. The facility is multipurpose, and the applications cover integral containment response tests, component tests, primary system tests, and separate effect tests. Experimental investigations carried on in the PANDA facility have been embedded in international projects, most of which under the auspices of the EU and OECD and with the support of a large number of organizations (regulatory bodies, technical dupport organizations, national laboratories, electric utilities, industries) worldwide. The paper provides an overview of the research programs performed in the PANDA facility in relation to BWR containment systems and those planned for PWR containment systems.

scholarly journals Remarks on Consistent Development of Plant Nodalizations: An Example of Application to the ROSA Integral Test Facility

2012 ◽  
Vol 2012 ◽  
pp. 1-17 ◽  
Author(s):  
J. Freixa ◽  
A. Manera
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Large Scale ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Test Facility ◽  
Integral Test ◽  
Energy Agency ◽  
Water Reactors ◽  
Regulatory Commission

Experimental results obtained at integral test facilities (ITFs) are used in the validation process of system codes for the transient analyses of light water reactors (LWRs). The expertise and guidelines derived from this work are later applied to transient analyses of nuclear power plants (NPPs). However, the boundary conditions at the NPPs will always differ from those at the ITF, and hence, the soundness of the ITF model needs to be maximized. An unaltered ITF nodalization should prove to be able to simulate as many tests as possible, before any conclusion is derived to NPP analyses. The STARS group at the Paul Scherrer Institut (PSI) actively participates in several international programs, where ITFs are being used (e.g., ROSA, PKL). Several tests carried out at the ROSA large-scale test facility operated by the Japan Atomic Energy Agency (JAEA) have been simulated in recent years by using the United States Nuclear Regulatory Commission (US-NRC) system code TRACE. In this paper, 5 different posttest analyses are presented, along with the evolution of the employed TRACE nodalization and the process followed to track the consistency of the nodalization modifications. The ROSA TRACE nodalization provided results in a reasonable agreement with all 5 experiments.

Analysis of VVER-1000 Large and Small Break LOCA Experiments With RELAP5/MOD3.2

2004 ◽  
Author(s):  
Maxym Rychkov ◽  
Utkarsh Chikkanagoudar ◽  
Bal Raj Sehgal
Keyword(s):  
Experimental Data ◽  
Test Data ◽  
Large Scale ◽  
Nuclear Safety ◽  
Test Facility ◽  
Integral Test ◽  
Power Scale ◽  
Accident Scenarios

A RELAP5 model for the analysis of the PSB-VVER test facility was developed by the EREC in Russia. The PSB-VVER is a large-scale integral test facility to model the VVER-1000 type NPP. The volume and power scale in this test facility is 1:300 and the elevation scale is 1:1, which corresponds to the elevation mark of the reactor prototype. At the Division of Nuclear Safety, RIT, Sweden, we have modified the PSB-VVER facility’s RELAP5 model in order to analyze two of the transient tests performed on the PSB-VVER facility, which serve as the validation matrix described by NEA/CSNI. The objective of the work conducted was to validate the results obtained from RELAP5’s calculation with the supplied experimental data from the PSB-VVER test facility. Two accident scenarios have been calculated and analyzed. After being verified against the “11% UP LOCA” test data, the RELAP5/MOD3.2 model was used for a so-called “blind” transient calculation of the test “2×25% HL LOCA” and the results obtained were compared with the experimental data provided after the calculation.

Scaling analysis of core pressure drop in reduced height integral test facility

Kerntechnik ◽  
2018 ◽  
Vol 83 (3) ◽  
pp. 178-180
Author(s):  
P. Ju ◽  
B. Long ◽  
L. Li ◽  
Q. Su ◽  
X. Wu ◽  
...  
Keyword(s):  
Pressure Drop ◽  
Test Facility ◽  
Scaling Analysis ◽  
Integral Test ◽  
Reduced Height
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