scholarly journals The Integral Test Facility Karlstein

2012 ◽  
Vol 2012 ◽  
pp. 1-12 ◽  
Author(s):  
Stephan Leyer ◽  
Michael Wich
Keyword(s):  
Power Supply ◽  
Test Facility ◽  
Experimental Program ◽  
Core Flooding ◽  
Integral Test ◽  
Large Power ◽  
Passive Safety Systems ◽  
Loss Of Coolant Accident ◽  
Accident Scenarios ◽  
Safety Systems

The Integral Test Facility Karlstein (INKA) test facility was designed and erected to test the performance of the passive safety systems of KERENA, the new AREVA Boiling Water Reactor design. The experimental program included single component/system tests of the Emergency Condenser, the Containment Cooling Condenser and the Passive Core Flooding System. Integral system tests, including also the Passive Pressure Pulse Transmitter, will be performed to simulate transients and Loss of Coolant Accident scenarios at the test facility. The INKA test facility represents the KERENA Containment with a volume scaling of 1 : 24. Component heights and levels are in full scale. The reactor pressure vessel is simulated by the accumulator vessel of the large valve test facility of Karlstein—a vessel with a design pressure of 11 MPa and a storage capacity of 125 m3. The vessel is fed by a benson boiler with a maximum power supply of 22 MW. The INKA multi compartment pressure suppression Containment meets the requirements of modern and existing BWR designs. As a result of the large power supply at the facility, INKA is capable of simulating various accident scenarios, including a full train of passive systems, starting with the initiating event—for example pipe rupture.

Safety Assessment and Accident Analysis of VVER-1000/466B With Active and Passive Safety Systems for Belene NPP

2014 ◽  
Author(s):  
Luben Sabotinov ◽  
Borislav Dimitrov ◽  
Giovanni B. Bruna
Keyword(s):  
Cooling System ◽  
Safety Evaluation ◽  
Safety Issues ◽  
Passive Safety Systems ◽  
Loss Of Coolant Accident ◽  
Depth Analysis ◽  
Active Safety Systems ◽  
Accident Scenarios ◽  
Core Cooling ◽  
Safety Systems

The paper presents the methodology adopted to assess the Interim Safety Analysis Report (ISAR) of the Belene NPP in the framework of the contract between the Bulgarian Nuclear Regulatory Authority (BNRA) and RISKAUDIT (IRSN&GRS). It stresses the in-depth analysis carried-out for several relevant-to-safety issues and illustrates in some detail the investigation of the Large Break Loss of Coolant Accident (LB LOCA) with loss of power and failure of the active part of the Emergency Core Cooling System (High Pressure and Low Pressure Safety Injection pumps), performed with the French best estimate thermal-hydraulic code CATHARE. The role, problems and efficiency of the passive and active safety systems during the accident scenarios are discussed. Finally, the main conclusions of the safety evaluation of the Belene NPP project are summarized.

Investigation of effects of nonavailability of passive safety systems on the reactor behaviour during LOCA Scenario in AP600

Kerntechnik ◽  
2021 ◽  
Vol 86 (3) ◽  
pp. 244-255
Author(s):  
S. H. Abdel-Latif ◽  
A. M. Refaey
Keyword(s):  
Passive Safety ◽  
Primary System ◽  
Water Storage Tank ◽  
The Core ◽  
Passive Safety Systems ◽  
Loss Of Coolant Accident ◽  
Passive Safety System ◽  
Stage 4 ◽  
Core Cooling ◽  
Safety Systems

Abstract The AP600 is a Westinghouse Advanced Passive PWR with a two–loop 1 940 MWt. This reactor is equipped with advanced passive safety systems which are designed to operate automatically at desired set-points. On the other hand, the failure or nonavailability to operate of any of the passive safety systems may affect reactor safety. In this study, modeling and nodalization of primary and secondary loops, and all passive reactor cooling systems are conducted and a 10-inch cold leg break LOCA is analyzed using ATHLET 3.1A Code. During loss of coolant accident in which the passive safety system failure or nonavailability are considered, four different scenarios are assumed. Scenario 1 with the availability of all passive systems, scenario 2 is failure of one of the accumulators to activate, scenario 3 is without actuation of the automatic depressurization system (ADS) stages 1–3, and scenario 4 is without actuation of ADS stage 4. Results indicated that the actuation of passive safety systems provide sufficient core cooling and thus could mitigate the accidental consequence of LOCAs. Failure of one accumulator during LOCA causes early actuation of ADS and In-Containment Refueling Water Storage Tank (IRWST). In scenario 3 where the LOCA without ADS stages 1–3 actuations, the depressurization of the primary system is relatively slow and the level of the core coolant drops much earlier than IRWST actuation. In scenario 4 where the accident without ADS stage-4 activation, results in slow depressurization and the level of the core coolant drops earlier than IRWST injection. During the accident process, the core uncovery and fuel heat up did not happen and as a result the safety of AP600 during a 10-in. cold leg MBLOCA was established. The relation between the cladding surface temperature and the primary pressure with the actuation signals of the passive safety systems are compared with that of RELAP5/Mode 3.4 code and a tolerable agreement was obtained.

AP1000® Passive Cooling Containment Analysis of a Double-Ended LBLOCA With a 3D Gothic Model

2018 ◽  
Author(s):  
Samanta Estevez-Albuja ◽  
Gonzalo Jimenez ◽  
Kevin Fernández-Cosials ◽  
César Queral ◽  
Zuriñe Goñi
Keyword(s):  
Cooling System ◽  
Heat Removal ◽  
Water Tank ◽  
Local Pressure ◽  
Water Storage Tank ◽  
Passive Safety Systems ◽  
Containment Model ◽  
Loss Of Coolant Accident ◽  
Ac Power ◽  
Safety Systems

In order to enhance Generation II reactors safety, Generation III+ reactors have adopted passive mechanisms for their safety systems. In particular, the AP1000® reactor uses these mechanisms to evacuate heat from the containment by means of the Passive Containment Cooling System (PCS). The PCS uses the environment atmosphere as the ultimate heat sink without the need of AC power to work properly during normal or accidental conditions. To evaluate its performance, the AP1000 PCS has been usually modeled with a Lumped Parameters (LP) approach, coupled with another LP model of the steel containment vessel to simulate the accidental sequences within the containment building. However, a 3D simulation, feasible and motivated by the current computational capabilities, may be able to produce more detailed and accurate results. In this paper, the development and verification of an integral AP1000® 3D GOTHIC containment model, taking into account the shield building, is briefly presented. The model includes all compartments inside the metallic containment liner and the external shield building. Passive safety systems, such as the In-containment Refueling Water Storage Tank (IRWST) with the Passive Residual Heat Removal (PRHR) heat exchanger and the Automatic Depressurization System (ADS), as well as the PCS, are included in the model. The model is tested against a cold leg Double Ended Guillotine Break Large Break Loss of Coolant Accident (DEGB LBLOCA) sequence, taking as a conservative assumption that the PCS water tank is not available during the sequence. The results show a pressure and temperature increase in the containment in consonance with the current literature, but providing a greater detail of the local pressure and temperature in all compartments.

scholarly journals Experimental investigation of LWR passive safety systems performance at the INKA test facility

2019 ◽  
Vol 137 ◽  
pp. 01035
Author(s):  
Rafał Bryk ◽  
Thomas Mull ◽  
Holger Schmidt
Keyword(s):  
Test Facility ◽  
Passive Safety ◽  
Light Water Reactors ◽  
Passive Systems ◽  
Actual Performance ◽  
Passive Safety Systems ◽  
Original Plant ◽  
Water Reactors ◽  
Original Objective ◽  
Safety Systems

INKA is a test facility designed by Framatome and built in the technical center in Karlstein. The original objective for establishing this test rig was the investigation of the performance of the passive safety systems developed in a new Framatome Boiling Water Reactor (BWR) design – KERENA. INKA was constructed in the scale of 1:1 in heights while the total volume of the containment was replicated in 1:24. Since the geometries of particular safety systems are faithfully reflected, their actual performance in the original plant can be investigated at the full scale. Due to the unquestionable interest of the nuclear community in the inherent safety, not only new BWR and PWR designs are equipped with the passive systems, but also particular passive solutions are considered to be applied into the already existing Light Water Reactors (LWR). In this context and due to the fact that both, single component tests and integral tests can be conducted at INKA, the facility can be employed for a demonstration/qualification of a large range of passive safety systems foreseen for quite different types of LWRs. Hence, the goal of the EASY project was the experimental confirmation of the passive systems performance and the analysis of their interactions between each other in the integral tests. Besides, the overarching target of all tests performed at INKA is provision of data for codes validation. This paper presents major outcomes and conclusions drawn on the basis of EASY project results.

Scaling Analysis of the Pressurizer Surge Line in Nuclear Plant Integral Test Facilities

2012 ◽  
Author(s):  
Cheng-Cheng Deng ◽  
Hua-Jian Chang ◽  
Ben-Ke Qin ◽  
Han Wang ◽  
Lian Chen
Keyword(s):  
Reactor Core ◽  
Nuclear Plant ◽  
Test Facility ◽  
Scaling Analysis ◽  
Two Phase ◽  
Integral Test ◽  
Loss Of Coolant Accident ◽  
Initial Injection ◽  
Surge Line ◽  
Core Cooling

During small break loss of coolant accident (SBLOCA) of AP1000 nuclear plant, the behavior of pressurizer surge line has an important effect on the operation of ADS valves and the initial injection of IRWST, which may happen at a time when the reactor core reaches its minimum inventory. Therefore, scaling analysis of the PRZ surge line in nuclear plant integral test facilities is important. Four scaling criteria of surge line are developed, which respectively focus on two-phase flow pattern transitions, counter-current flow limitation (CCFL) behavior, periodic draining and filling and maintaining system inventory. The relationship between the four scaling criteria is discussed and comparative analysis of various scaling results is performed for different length scale ratios of test facilities. The results show that CCFL phenomenon and periodic draining and filling behavior are relatively more important processes and the surge line diameter ratios obtained by the two processes’ scaling criteria are close to each other. Thus, an optimal scaling analysis considering both CCFL phenomenon and periodic draining and filling process of PRZ surge line is given, which is used to provide a practical reference to choose appropriate scale of the surge line for the ACME (Advanced Core-cooling Mechanism Experiment) test facility now being built in China.

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 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.

Experimental Investigation of Non-Condensable Gases Effect on Novovoronezh NPP-2 Steam Generator Condensation Power Under the Condition of Passive Safety Systems Operation

2009 ◽  
Author(s):  
I. I. Kopytov ◽  
S. G. Kalyakin ◽  
V. M. Berkovich ◽  
A. V. Morozov ◽  
O. V. Remizov
Keyword(s):  
Steam Generator ◽  
Flow Rate ◽  
Computational Simulation ◽  
Heat Removal ◽  
Test Facility ◽  
Passive Safety ◽  
Primary Circuit ◽  
Water Radiolysis ◽  
Passive Safety Systems ◽  
Safety Systems

The design substantiation of the heat removal efficiency from Novovoronezh NPP-2 (NPP-2006 project with VVER-1200 reactor) reactor core in the event of primary circuit leaks and operation of passive safety systems only (among these are the systems of hydroaccumulators of the 1st and 2nd stages and passive heat removal system) has been performed based on computational simulation of the related processes in the reactor and containment. The computational simulation has been performed with regard to the detrimental effect of non-condensable gases on steam generator (SG) condensation power. Nitrogen arriving at the circuit with the actuation of hydroaccumulators of the 1st stage and products of water radiolysis are the main sources of non-condensable gases in the primary circuit. The feature of Novovoronezh NPP-2 passive safety systems operation is that during the course of emptying of the 2nd stage hydroaccumulators system (HA-2) the gas-steam mixture spontaneously flows out from SG cold headers into the volume of HA-2 tanks. The flow rate of gas-steam mixture during the operation of HA-2 system is equal to the volumetric water discharge from hydroaccumulators. The calculations carried out by different integral thermal hydraulic codes revealed that this volume flow rate of gas-steam mixture from SG to the HA-2 system would suffice to eliminate the “poisoning” of SG piping and to maintain necessary condensation power. In support of the calculation results, the experiments were carried out at the HA2M-SG test facility constructed at IPPE. The test facility incorporates a VVER steam generator model of volumetric-power scale of 1:46. Steam to the HA2M-SG test facility is supplied fed from the IPPE heat power plant. Gas addition to steam coming to the SG model is added from high pressure gas cylinders. Nitrogen and helium are used in the experiments for simulating hydrogen. The report presents the basic results of experimental investigations aimed at the evaluation of SG condensation power under the inflow of gas-steam mix with different gases concentration to the tube bundle, both under the simulation of gas-steam mixture outflow from SG cold header to the HA-2 system and without outflow. As a result of the research performed at the HA2M-SG test facility, it has been substantiated experimentally that in the event of an emergency leak steam generators have condensation power sufficient for effective heat removal from the reactor provided by PHR system.

Scaling Assessment for the Integral Effect Tests: 2 — Open System Depressurization

2013 ◽  
Author(s):  
Xiaoyu Cai
Keyword(s):  
Open System ◽  
Computer Code ◽  
Passive Safety ◽  
Prototype System ◽  
Integral Test ◽  
Pressurized Water ◽  
Passive Safety Systems ◽  
Loss Of Coolant Accident ◽  
Water Reactors ◽  
Westinghouse Electric

In this paper a method is described for using NOTRUMP models to corroborate the Hierarchical Two-Tiered Scaling (H2TS) methodology that has been used for design of the APEX and SPES-2 test facilities. These facilities were built for the Westinghouse Electric Corporation to obtain data on the performance of the passive safety systems of the advanced pressurized water reactors. Similarity between the prototype system and the scaled test facilities is investigated for the open system depressurization phenomena in the postulated small break loss of coolant accident transient. The objective of this analysis is to provide a basis that an integral test for the passive safety system will provide valid experimental data for the high-ranked phenomena that may occur during the hypothetical SBLOCA transients. In this way, the experiment will capture the phenomena that would be expected to occur in the plant, and hence provide data that can be used to validate computer code.

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.

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