Application of Thermal Hydraulic and Severe Accident Code SOCRAT/V3 to Bottom Water Reflood Experiment QUENCH-LOCA-1

2016 ◽  
Vol 2 (2) ◽  
Author(s):  
Alexander Vasiliev ◽  
Juri Stuckert
Keyword(s):  
Nuclear Power ◽  
Hydrogen Generation ◽  
Cooling System ◽  
Reactor Core ◽  
Severe Accident ◽  
Loss Of Coolant Accident ◽  
Post Test ◽  
Core Cooling ◽  
Modeling Code ◽  
Code Quality

This study aims to (1) use the thermal hydraulic and severe fuel damage (SFD) best-estimate computer modeling code SOCRAT/V3 for post-test calculation of QUENCH-LOCA-1 experiment and (2) estimate the SOCRAT code quality of modeling. The new QUENCH-LOCA bundle tests with different cladding materials will simulate a representative scenario for a loss-of-coolant-accident (LOCA) nuclear power plant (NPP) accident sequence in which the overheated (up to 1050°C) reactor core would be reflooded from the bottom by the emergency core cooling system (ECCS). The test QUENCH-LOCA-1 was successfully performed at the KIT, Karlsruhe, Germany, on February 2, 2012, and was the first test for this series after the commissioning test QUENCH-LOCA-0 conducted earlier. The SOCRAT/V3-calculated results describing thermal hydraulic, hydrogen generation, and thermomechanical behavior including rods ballooning and burst are in reasonable agreement with the experimental data. The results demonstrate the SOCRAT code’s ability for realistic calculation of complicated LOCA scenarios.

scholarly journals Evaluation of passive autocatalytic recombiners operation efficiency by means of the lumped parameter approach

Nukleonika ◽  
2015 ◽  
Vol 60 (2) ◽  
pp. 339-345 ◽  
Author(s):  
Tomasz Bury
Keyword(s):  
Hydrogen Generation ◽  
Cooling System ◽  
Component Mixture ◽  
Pressurized Water ◽  
Water Reactor ◽  
Safety Issues ◽  
Loss Of Coolant Accident ◽  
Lumped Parameter ◽  
Core Cooling ◽  
Parameter Approach

Abstract The problem of hydrogen behavior in containment buildings of nuclear reactors belongs to thermal-hydraulic area. Taking into account the size of systems under consideration and, first of all, safety issues, such type of analyses cannot be done by means of full-scale experiments. Therefore, mathematical modeling and numerical simulations are widely used for these purposes. A lumped parameter approach based code HEPCAL has been elaborated in the Institute of Thermal Technology of the Silesian University of Technology for simulations of pressurized water reactor containment transient response. The VVER-440/213 and European pressurised water reactor (EPR) reactors containments are the subjects of analysis within the framework of this paper. Simulations have been realized for the loss-of-coolant accident scenarios with emergency core cooling system failure. These scenarios include core overheating and hydrogen generation. Passive autocatalytic recombiners installed for removal of hydrogen has been taken into account. The operational efficiency of the hydrogen removal system has been evaluated by comparing with an actual hydrogen concentration and flammability limit. This limit has been determined for the three-component mixture of air, steam and hydrogen. Some problems related to the lumped parameter approach application have been also identified.

scholarly journals Thermal Hydraulic Analysis of a Nuclear Reactor due to Loss of Coolant Accident with and without Emergency Core Cooling System

2020 ◽  
Vol 01 (02) ◽  
pp. 53-60
Author(s):  
Pronob Deb Nath ◽  
Kazi Mostafijur Rahman ◽  
Md. Abdullah Al Bari
Keyword(s):  
Nuclear Reactor ◽  
Cooling System ◽  
Computational Study ◽  
Reactor Core ◽  
Primary Circuit ◽  
Feed Water ◽  
Pressurized Water ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Core Cooling

This paper evaluates the thermal hydraulic behavior of a pressurized water reactor (PWR) when subjected to the event of Loss of Coolant Accident (LOCA) in any channel surrounding the core. The accidental break in a nuclear reactor may occur to circulation pipe in the main coolant system in a form of small fracture or equivalent double-ended rupture of largest pipe connected to primary circuit line resulting potential threat to other systems, causing pressure difference between internal parts, unwanted core shut down, explosion and radioactivity release into environment. In this computational study, LOCA for generation III+ VVER-1200 reactor has been carried out for arbitrary break at cold leg section with and without Emergency Core Cooling System (ECCS). PCTRAN, a thermal hydraulic model-based software developed using real data and computational approach incorporating reactor physics and control system was employed in this study. The software enables to test the consequences related to reactor core operations by monitoring different operating variables in the system control bar. Two types of analysis were performed -500% area break at cold leg pipe due to small break LOCA caused by malfunction of the system with and without availability of ECCS. Thermal hydraulic parameters like, coolant dynamics, heat transfer, reactor pressure, critical heat flux, temperature distribution in different sections of reactor core have also been investigated in the simulation. The flow in the reactor cooling system, steam generators steam with feed-water flow, coolant steam flow through leak level of water in different section, power distribution in core and turbine were plotted to analyze their behavior during the operations. The simulation showed that, LOCA with unavailability of Emergency Core Cooling System (ECCS) resulted in core meltdown and release of radioactivity after a specific time.

Modeling of Thermal Hydraulics Features of Top Water Reflood Experiment PARAMETER-SF3 Using SOCRAT/V2 Code

2009 ◽  
Author(s):  
Arcadii E. Kisselev ◽  
Valerii F. Strizhov ◽  
Alexander D. Vasiliev ◽  
Vladimir I. Nalivayev ◽  
Nikolay Ya. Parshin
Keyword(s):  
Experimental Data ◽  
Nuclear Power ◽  
Cooling System ◽  
Severe Accident ◽  
Numerical Code ◽  
Thermal Hydraulics ◽  
Vver Fuel ◽  
Design Basis ◽  
Best Estimate ◽  
Core Cooling

The PARAMETER-SF3 test conditions simulated a severe LOCA (Loss of Coolant Accident) nuclear power plant sequence in which the overheated up to 1700÷2300K core would be reflooded from the top and the bottom in occasion of ECCS (Emergency Core Cooling System) recovery. The test was successfully conducted at the NPO “LUTCH”, Podolsk, Russia, in October 31, 2008, and was the third of four experiments of series PARAMETER-SF. PARAMETER facility of NPO “LUTCH”, Podolsk, is designed for studies of the VVER fuel assemblies behavior under conditions simulating design basis, beyond design basis and severe accidents. The test bundle was made up of 19 fuel rod simulators with a length of approximately 3.12 m (heated rod simulators) and 2.92 m (unheated rod simulator). Heating was carried out electrically using 4-mm-diameter tantalum heating elements installed in the center of the rods and surrounded by annular UO2 pellets. The rod cladding was identical to that used in VVERs: Zr1%Nb, 9.13 mm outside diameter, 0.7 mm wall thickness. After the maximum cladding temperature of about 1900K was reached in the bundle during PARAMETER-SF3 test, the top flooding was initiated. The thermal hydraulic and SFD (Severe Fuel Damage) best estimate numerical complex SOCRAT/V2 was used for the calculation of PARAMETER-SF3 experiment. The counter-current flow limitation (CCFL) model was implemented to best estimate numerical code SOCRAT/V2 developed for modeling thermal hydraulics and severe accident phenomena in a reactor. Thermal hydraulics in PARAMETER-SF3 experiment played very important role and its adequate modeling is important for the thermal analysis. The results obtained by the complex SOCRAT/V2 were compared with experimental data concerning different aspects of thermal hydraulics behavior including the CCFL phenomenon during the reflood. The temperature experimental data were found to be in a good agreement with calculated results. It is indicative of the adequacy of modeling the complicated thermo-hydraulic behavior in the PARAMETER-SF3 test.

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.

Application of Thermal Hydraulic Code SOCRAT/V2 to Top Water Reflood Experiment PARAMETER-SF3

2009 ◽  
Author(s):  
Alexander D. Vasiliev
Keyword(s):  
Experimental Data ◽  
Nuclear Power ◽  
Cooling System ◽  
Severe Accident ◽  
Numerical Code ◽  
Thermal Hydraulics ◽  
Vver Fuel ◽  
Design Basis ◽  
Best Estimate ◽  
Core Cooling

The PARAMETER-SF3 test conditions simulated a severe LOCA (Loss of Coolant Accident) nuclear power plant sequence in which the overheated up to 1700–2300K core would be reflooded from the top and the bottom in occasion of ECCS (Emergency Core Cooling System) recovery. The test was successfully conducted at the NPO “LUTCH”, Podolsk, Russia, in October 31, 2008, and was the third of four experiments of series PARAMETER-SF. PARAMETER facility of NPO “LUTCH”, Podolsk, is designed for studies of the VVER fuel assemblies behavior under conditions simulating design basis, beyond design basis and severe accidents. The test bundle was made up of 19 fuel rod simulators with a length of approximately 3.12 m (heated rod simulators) and 2.92 m (unheated rod simulator). Heating was carried out electrically using 4-mm-diameter tantalum heating elements installed in the center of the rods and surrounded by annular UO2 pellets. The rod cladding was identical to that used in VVERs: Zr1%Nb, 9.13 mm outside diameter, 0.7 mm wall thickness. After the maximum cladding temperature of about 1900K was reached in the bundle during PARAMETER-SF3 test, the top flooding was initiated. The thermal hydraulic and SFD (Severe Fuel Damage) best estimate numerical complex SOCRAT/V2 was used for the calculation of PARAMETER-SF3 experiment. The counter-current flow limitation (CCFL) model was implemented to best estimate numerical code SOCRAT/V2 developed for modeling thermal hydraulics and severe accident phenomena in a reactor. Thermal hydraulics in PARAMETER-SF3 experiment played very important role and its adequate modeling is important for the thermal analysis. The results obtained by the complex SOCRAT/V2 were compared with experimental data concerning different aspects of thermal hydraulics behavior including the CCFL phenomenon during the reflood. The temperature experimental data were found to be in a good agreement with calculated results. It is indicative of the adequacy of modeling the complicated thermo-hydraulic behavior in the PARAMETER-SF3 test.

scholarly journals CFD Analysis of Passive Autocatalytic Recombiner

2011 ◽  
Vol 2011 ◽  
pp. 1-9
Author(s):  
B. Gera ◽  
P. K. Sharma ◽  
R. K. Singh ◽  
K. K. Vaze
Keyword(s):  
Nuclear Power ◽  
Cooling System ◽  
Maximum Temperature ◽  
Convection Flow ◽  
Heat Of Reaction ◽  
Loss Of Coolant Accident ◽  
Flow Through ◽  
Core Cooling ◽  
Catalytic Recombination ◽  
Steam Heat

In water-cooled nuclear power reactors, significant quantities of hydrogen could be produced following a postulated loss-of-coolant accident (LOCA) along with nonavailability of emergency core cooling system (ECCS). Passive autocatalytic recombiners (PAR) are implemented in the containment of water-cooled power reactors to mitigate the risk of hydrogen combustion. In the presence of hydrogen with available oxygen, a catalytic reaction occurs spontaneously at the catalyst surfaces below conventional ignition concentration limits and temperature and even in presence of steam. Heat of reaction produces natural convection flow through the enclosure and promotes mixing in the containment. For the assessment of the PAR performance in terms of maximum temperature of catalyst surface and outlet hydrogen concentration an in-house 3D CFD model has been developed. The code has been used to study the mechanism of catalytic recombination and has been tested for two literature-quoted experiments.

scholarly journals Early quantification of reliability for a safety critical and control system: a case study of reactor core cooling system

2018 ◽  
Vol 7 (2.12) ◽  
pp. 248
Author(s):  
Vinay Kumar ◽  
Suraj Gupta ◽  
Anil Kumar Tripathi
Keyword(s):  
Control System ◽  
Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Cooling System ◽  
Reactor Core ◽  
Safety Critical ◽  
Core Cooling ◽  
And Control

Using Probabilistic Reliability analysis for Quantifying reliability of a system is already a common practice in Reliability Engineering community. This method plays an important role in analyzing reliability of nuclear plants and its various components. In Nuclear Power Plants Reactor Core Cooling System is a component of prime importance as its breakdown can disrupt Cooling System of power plant. In this paper, we present a framework for early quantification of Reliability and illustrated with a Safety Critical and Control System as case study which runs in a Nuclear Power Plant.  

The Development of the Evolutionary BWR

2006 ◽  
Author(s):  
A. Murase ◽  
M. Nakamaru ◽  
M. Kuroki ◽  
Y. Kojima ◽  
S. Yokoyama
Keyword(s):  
Cooling System ◽  
Reactor Core ◽  
Heat Removal ◽  
Safety Performance ◽  
Severe Accident ◽  
Main Role ◽  
Isolation System ◽  
Near Term ◽  
Core Cooling ◽  
Control Rods

Considering the delay of the fast breeding reactor (FBR) development, it is expected that the light water reactor will still play the main role of the electric power generation in the 2030’s. Accordingly, Toshiba has been developing a new conceptual ABWR as the near-term BWR. We tentatively call it AB1600. The AB1600 has introduced the hybrid active/passive safety system in order to improve countermeasure against severe accident (SA). At the same time, we have made the simplification of the overall plant systems in order to improve economy. The simplification of the AB1600 is based on the proven technologies. To retain the safety performance superior or equivalent to the current ABWR and to strengthen the countermeasure against SA, the AB1600 has introduced the passive systems such as the passive containment cooling system (PCCS), the gravity driven core cooling system (GDCS) and the isolation condenser (IC). While we retain the safety performance superior or equivalent to the current ABWR, we have made the simplification of the safety systems. We could eliminate the high pressure core flooder system (HPCF) and the reactor core isolation system (RCIC) by extending the height of reactor pressure vessel (RPV) two meters. To achieve simplification of reactor systems, we have reduced the number of fuel bundles and the number of control rods by adopting large bundle that has a bundle pitch 1.2 times wider than that of the current ABWR. In the 1600MWe class, the number of fuel bundles could be reduced to 600 from 872 of the current ABWR, and the number of control rods could be reduced to 137 from 205 of the current ABWR. Because the reactor internal pump (RIP) of the current ABWR has sufficient performance capacity and the improvement of fuel characteristics from the current fuel enables the operation at lower core flow, the number of RIPs could be decreased from ten to eight. Furthermore, we have reduced the number of divisions of emergency core cooling system (ECCS)/heat removal system to two from three of the current ABWR. This configuration change contributes to reduce the amount of resources of not only reactor systems but also auxiliary systems. In the previous paper, the AB1600 had four low pressure flooder systems (LPFLs). We have studied about the possibility of reduction of LPFLs to two from four by providing the LPFL with alternative injection lines. This change is expected to contribute to reduce the total number of ECCS pumps and the capacity of emergency AC power.

Severe Accident Analysis for Reactor Core Applying SiC to Fuel Claddings and Channel Boxes

2018 ◽  
Author(s):  
Hideki Horie ◽  
Yutaka Takeuchi ◽  
Kenya Takiwaki ◽  
Fumie Sebe ◽  
Kazuo Kakiuchi ◽  
...  
Keyword(s):  
High Temperature ◽  
Oxidation Reaction ◽  
Nuclear Power ◽  
Reaction Rate ◽  
Hydrogen Generation ◽  
Reactor Core ◽  
The Other ◽  
Severe Accident ◽  
Fuel Cladding ◽  
The Difference

Development of a fuel cladding or a channel box applying silicon carbide (SiC), which has high accident tolerance, in place of zircaloy (Zry) or Steel Use Stainless (SUS) composing current light water reactors, has being proceeded with after the accident of Fukushima Daiichi Nuclear Power Plant (1F). When applying SiC to core structures of a nuclear power plant such as fuel cladding, it is expected that the difference of high temperature oxidation characteristics in the severe accident (SA) conditions would mitigate progression of core damage comparing with the current Zry fuel core. This study performed SA analyses considering high temperature chemical reaction characteristics of SiC by using SA analysis code “MAAP”, and thermal hydraulics analysis code “TRAC Toshiba version (TRAC)”, and compared the difference between SiC and Zry. Both codes originally have no model of oxidation reaction for SiC. Hence, a new model for SiC in addition to the current model for Zry was incorporated into “MAAP”. On the other hand, “TRAC” adjusted reaction rate by changing oxidation reaction coefficients in the current Zry oxidation reaction models such as Baker-Just and Cathcart correlations in order to simulate SiC-water/steam reaction. In analysis using “MAAP”, seven accident sequences from representative Probabilistic Risk Assessment ones were selected to evaluate the difference of SA behavior between two materials. As a result, in the case of replacing current Zry of fuel claddings and channel boxes into SiC, an amount of hydrogen generation reduced to about 1/6 than the case of Zry. In addition to that, in the case of replacing SUS structures in the reactor core into SiC, an amount of hydrogen generation moreover reduced to about 1/6 than the above result, which means just about 2% of an amount in the original case. On the other hand, in analysis using “TRAC”, the accident sequence for unit 3 of 1F (1F3) was selected, and reaction rate in the oxidation reaction model was examined as parameter. In the case of 1.0 time of the reaction rate, which means an original reaction rate, maximum fuel cladding temperature exceeded 2000K in 50 hour after reactor scram. However, using the reaction rate below 0.01 to the original one, the fuel cladding temperature didn’t exceed 1,600K.

scholarly journals Evaluation of Heat Removal from RBMK-1500 Core Using Control Rods Cooling Circuit

2008 ◽  
Vol 2008 ◽  
pp. 1-8
Author(s):  
A. Kaliatka ◽  
E. Uspuras ◽  
M. Vaisnoras
Keyword(s):  
Nuclear Power ◽  
Cooling System ◽  
Reactor Core ◽  
Heat Removal ◽  
Thermal Inertia ◽  
Severe Accident ◽  
Cooling Circuit ◽  
High Thermal Inertia ◽  
The Moment ◽  
Control Rods

The Ignalina nuclear power plant is a twin unit with two RBMK-1500, graphite moderated, boiling water, multichannel reactors. After the decision was made to decommission the Ignalina NPP, Unit 1 was shut down on December 31, 2004, and Unit 2 is to be operated until the end of 2009. Despite of this fact, severe accident management guidelines for RBMK-1500 reactor at Ignalina NPP are prepared. In case of beyond design basis accidents, it can occur that no water sources are available at the moment for heat removal from fuel channels. Specificity of RBMK reactor is such that the channels with control rods are cooled with water supplied by the system totally independent from the reactor cooling system. Therefore, the heat removal from RBMK-1500 reactor core using circuit for cooling of rods in control and protection system can be used as nonregular mean for reactor cooldown in case of BDBA. The heat from fuel channels, where heat is generated, through graphite bricks is transferred in radial direction to cooled CPS channels. This article presents the analysis of possibility to remove heat from reactor core in case of large LOCA by employing CPS channels cooling circuit. The analysis was performed for Ignalina NPP with RBMK-1500 reactor using RELAP5-3D and RELAP5 codes. Results of the analysis have shown that, in spite of high thermal inertia of graphite, this heat removal from CPS channels allows to slow down effectively the core heat-up process.

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