scholarly journals Assessment of ASTEC 2.1 capability to predict reactor core behaviour at the late phase of severe accident progression based on QUENCH-12

2020 ◽  
Vol 1555 ◽  
pp. 012033
Author(s):  
Petya Vryashkova ◽  
Pavlin Groudev
Keyword(s):  
Late Phase ◽  
Reactor Core ◽  
Severe Accident

Code Simulation of Quenching of a High Temperature Debris Bed: Model Improvement and Validation With Experimental Results

2012 ◽  
Author(s):  
A. Bachrata ◽  
F. Fichot ◽  
G. Repetto ◽  
M. Quintard ◽  
J. Fleurot
Keyword(s):  
Late Phase ◽  
Reactor Core ◽  
Injection Rate ◽  
Mitigation Strategies ◽  
Experimental Results ◽  
Severe Accident ◽  
Simulation Tools ◽  
Variation Of Parameters ◽  
Model Improvement ◽  
Accident Sequences

The loss of coolant accidents with core degradation e.g. TMI-2 and Fukushima demonstrated that the nuclear safety analysis has to cover accident sequences involving a late reflood activation in order to develop appropriate and reliable mitigation strategies for both, existing and advanced reactors. The reflood (injection of water) is possible if one or several water sources become available during the accident. In a late phase of accident, no well-defined coolant paths would exist and a large part of the core would resemble to a debris bed e.g. particles with characteristic length-scale: 1 to 5 mm, as observed in TMI-2. The French “Institut de Radioprotection et de Sûreté Nucléaire” (IRSN) is developing experimental programs (PEARL and PRELUDE) and simulation tools (ICARE-CATHARE and ASTEC) to study and optimize the severe accident management strategy and to assess the probabilities to stop the progress of in-vessel core degradation at a late stage of an accident. The purpose of this paper is to propose a consistent thermo-hydraulic model of reflood of severely damaged reactor core for ICARE-CATHARE code. The comparison of the calculations with PRELUDE experimental results is presented. It is shown that the quench front exhibits either a 1D behavior or a 2D one, depending on injection rate or bed characteristics. The PRELUDE data cover a rather large range of variation of parameters for which the developed model appears to be quite predictive.

Analysis of Late Phase Severe Accident Phenomena in CANDU Plant

2017 ◽  
Vol 3 (2) ◽  
Author(s):  
D. Dupleac
Keyword(s):  
Heat Transfer ◽  
Fluid Dynamics ◽  
Late Phase ◽  
Severe Accident ◽  
Ansys Fluent ◽  
Water Depletion ◽  
Sources Of Uncertainty ◽  
Parametric Studies ◽  
Computational Fluid Dynamics Cfd ◽  
Containment Pressure

The paper overviews the analytical studies performed at Politehnica University of Bucharest on the analysis of late phase severe accident phenomena in a Canada Deuterium Uranium (CANDU) plant. The calculations start from a dry debris bed at the bottom of calandria vessel. Both SCDAPSIM/RELAP code and ansys-fluent computational fluid dynamics (CFD) code are used. Parametric studies are performed in order to quantify the effect of several identified sources of uncertainty on calandria vessel failure: metallic fraction of zirconium inside the debris, containment pressure, timing of water depletion inside calandria vessel, steam circulation in calandria vessel above debris bed, debris temperature at moment of water depletion inside calandria vessel, calandria vault nodalization, and the gap heat transfer coefficient.

scholarly journals Analysis of humanoid robotics for nuclear disaster management incorporated with biomechanics

2019 ◽  
Vol 34 (3) ◽  
pp. 291-298
Author(s):  
Kyung Jang ◽  
Tae Woo
Keyword(s):  
Reactor Core ◽  
Nuclear Industry ◽  
Severe Accident ◽  
Physical Aspect ◽  
Multidisciplinary Research ◽  
Nuclear Disaster ◽  
Collapsed Building ◽  
Very High ◽  
Mechanical Studies

The humanoid is investigated for the mechanical and physical aspect in the nuclear disaster, especially for a severe accident, which includes the core melting. There are some mechanical studies of the leg and hand of the humanoid in which the human mimicking features are described. The management of the task is accomplished by the three regional preparations. The robot is made of the radiation-resistance substance. Therefore, it could work on the normal task of a human for the removal of the broken debris in a collapsed building. However, there is a limitation for the use in the reactor core building due to very high temperature of the nuclear fuel. The regional classification of the site is studied for the practical purposes. The post-accident analysis is accompanied with multidisciplinary research for the humanoid development in the nuclear industry.

Assessment of RELAP/SCDAPSIM for Turbine Trip Transient in NuScale-SMR

2018 ◽  
Author(s):  
Katarzyna Skolik ◽  
Anuj Trivedi ◽  
Marina Perez-Ferragut ◽  
Chris Allison
Keyword(s):  
Natural Circulation ◽  
System Modeling ◽  
Thermal Power ◽  
Reactor Core ◽  
Operating Conditions ◽  
Severe Accident ◽  
Circulation System ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Secondary Steam

The NuScale Small Modular Reactor (SMR) is an integrated Pressurized Water Reactor (iPWR) with the coolant flow based on the natural circulation. The reactor core consists of 37 fuel assemblies similar to those used in typical PWRs, but only half of their length to generate 160MW thermal power (50 MWe). Current study involves the development of a NuScale-SMR model based on its Design Certification Application (DCA) data (from NRC) using RELAP/SCDAPSIM. The turbine trip transient (TTT) was simulated and analysed. The objective was to assess this version of the code for natural circulation system modeling capabilities and also to verify the input model against the publicly available TTT results obtained using NRELAP5. This successful benchmark confirms the reliability of the thermal hydraulic model and allows authors to use it for further safety and severe accident analyses. The reactor core channels, pressurizer, riser and downcomer pipes as well as the secondary steam generator tubes and the containment were modeled with RELAP5 components. SCDAP core and control components were used for the fuel elements in the core. The final input deck achieved the steady state with the operating conditions comparable to those reported in the DCA. RELAP/SCDAPSIM predictions are found to be satisfactory and comparable to the reference study. It confirms the code code capabilities for natural circulation system transients.

Natural Circulation and Creep Damage Analyses of Severe Accident for a Typical PWR Plant

2009 ◽  
Author(s):  
Osamu Kawabata ◽  
Masao Ogino
Keyword(s):  
Steam Generator ◽  
Natural Circulation ◽  
Reactor Core ◽  
Creep Damage ◽  
Reactor Vessel ◽  
Reactor System ◽  
Severe Accident ◽  
Steam Generator Tube ◽  
Hot Gas ◽  
Core Meltdown

When the primary reactor system remain pressurized during core meltdown for a typical PWR plant, loop seals formed in the primary reactor system would lead to natural circulations in hot leg and steam generator. In this case, the hot gas released from the reactor core moves to a steam generator, and a steam generator tube would be failed with cumulative creep damage. From such phenomena, a high-pressure scenario during core meltdown may lead to large release of fission products to the environment. In the present study, natural circulation and creep damage in the primary reactor system accompanying the hot gas generation in the reactor core were discussed and the combining analysis with MELCOR and FLUENT codes were performed to examine the natural circulation behavior. For a typical 4 loop PWR plant, MELCOR code which can analyze for the severe accident progression was applied to the accident analyses from accident initiation to reactor vessel failure for the accident sequence of the main steam pipe break which is maintained at high pressure during core meltdown. In addition, using the CFD code FLUENT, fluid dynamics in the reactor vessel plenum, hot leg and steam generator of one loop were simulated with three-dimensional coordinates. And the hot gas natural circulation flow and the heat transfer to adjoining structures were analyzed using results provided by the MELCOR code as boundary conditions. The both ratios of the natural circulation flow calculated in the hot leg and the steam generator using MELCOR code and FLUENT code were obtained to be about 2 (two). And using analytical results of thermal hydraulic analysis with both codes, creep damage analysis at hottest temperature points of steam generator tube and hot leg were carried out. The results in both cases showed that a steam generator tube would be failed with creep rupture earlier than that of hot leg rupture.

Evaluation of Dump Tank Coolability in PHWRs During Late-Phase Severe Accident

2019 ◽  
Vol 5 (4) ◽  
Author(s):  
Pradeep Pandey ◽  
Parimal P. Kulkarni ◽  
Arun Nayak ◽  
Sumit V. Prasad
Keyword(s):  
Heavy Water ◽  
Late Phase ◽  
Atomic Power Station ◽  
Severe Accident ◽  
Large Temperature ◽  
Tank Wall ◽  
Power Station ◽  
Melt Temperature ◽  
Decay Heat ◽  
Water Reactors

In some of the older design of pressurized heavy water reactors (PHWRs), such as in Rajasthan Atomic Power Station (RAPS) and Madras Atomic Power Station (MAPS), in case of a severe accident, the debris/corium may cause failure of the dump port of calandria and relocate into the dump tank. The sensible and decay heat of debris/corium makes the heavy water in dump tank to boil off leaving the dry debris in dump tank. The dry debris remelt with time and the molten corium, thus, formed has the potential to breach the dump tank and move into the containment cavity, which is highly undesirable. Hence, as an accident management strategy, water is being flooded outside the dump tank using fire water hook-up lines to remove the heat from corium to cool and stabilize it and terminate the accident progression, similar to in vessel retention. However, the question is “is the molten corium coolable by this technique.” The coolability of the molten corium in dump tank as in the reactor is assessed by conducting experiments in a scaled facility using a simulant material having comparable thermophysical properties with that of corium. Melting of dry debris resting on dump tank bottom marks the beginning of the experimental investigation for present analysis. Decay heat is simulated by a set of immersed heaters inside the melt. Temperature profiles at different locations in dump tank and in the melt pool are obtained as a function of time to demonstrate the coolability with decay heat. Large temperature gradient is observed within the corium, involving high melt center temperature and low tank wall temperature suggesting formation of crust which insulates the dump tank wall from hot corium.

Experimental and Numerical Investigation of Solidification of Gallium in an Initially Emptied Horizontal Pipe Flow

2020 ◽  
Author(s):  
Shawn Somers-Neal ◽  
Alex Pegarkov ◽  
Edgar Matida ◽  
Vinh Tang ◽  
Tarik Kaya
Keyword(s):  
Numerical Model ◽  
Molten Metal ◽  
Reactor Core ◽  
Experimental Tests ◽  
Severe Accident ◽  
Penetration Length ◽  
Horizontal Pipe ◽  
Cooling Pipe ◽  
Severe Accidents ◽  
Core Meltdown

Abstract In a reactor core meltdown under postulated severe accidents, the molten material (corium) could be ejected or relocated through existing vessel penetrations (cooling pipe connections), thus potentially contaminating other locations in the power plant. There exists, however, a potential for plugging of melt flow due to its complete solidification, providing the availability of an adequate heat sink. Therefore, a numerical model was created to simulate the flow of molten metal through an initially empty horizontal pipe. The numerical model was verified using a previously developed analytical model and validated against experimental tests with gallium (low melting temperature) as a substitute for corium. The numerical model was able to predict the penetration length (length of distance travelled by the molten metal) after a complete blockage occurred with an average percent error range of 9%. Since the numerical model has been verified and validated, the model can be updated to predict the penetration length in the cooling pipe in case of a severe accident.

Effects of Ruthenium Release in Oxidizing Conditions on a BWR Source Term

2009 ◽  
Author(s):  
Atso Suopaja¨rvi ◽  
Teemu Ka¨rkela¨ ◽  
Ari Auvinen ◽  
Ilona Lindholm
Keyword(s):  
Pressure Vessel ◽  
Source Term ◽  
Reactor Core ◽  
Severe Accident ◽  
Ruthenium Oxides ◽  
The Core ◽  
The Standard Model ◽  
Reactor Building ◽  
Release Model ◽  
Scoping Studies

The release of ruthenium in oxygen-rich conditions from the reactor core during a severe accident may lead to formation of significantly more volatile ruthenium oxides than produced in steam atmosphere. The effect of volatile ruthenium release in a case a reference BWR nuclear plant was studied to get rough-estimates of the effects on the spreading of airborne ruthenium inside the containment and reactor building and the fission product source term. The selected accident scenario starting during shutdown conditions with pressure vessel upper head opened was a LOCA with a break in the bottom of the RPV. The results suggest that there is a remarkable amount of airborne Ru in the containment atmosphere, unlike with the standard MELCOR Ru release model which predicts no airborne Ru at all in the selected case. The total release of ruthenium from the core can be 5000 times the release predicted by the standard model. Based on the performed plant scoping studies it seems reasonable to take the release of volatile ruthenium oxides into account when assessing source terms for plants during shutdown states.

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