MELPROG-POW/MOD1: A two-dimensional, mechanistic code for analysis of reactor core melt progression and vessel attack under severe accident conditions

1989 ◽  
Author(s):  
S. S. Dosanjh
Keyword(s):  
Reactor Core ◽  
Severe Accident ◽  
Two Dimensional ◽  
Accident Conditions

Best Estimate Analysis of PHEBUS FPT1 Test Using RELAP/SCDAPSIM Code

2012 ◽  
Author(s):  
Tadas Kaliatka ◽  
Eugenijus Ušpuras ◽  
Virginijus Vileiniškis
Keyword(s):  
Reactor Core ◽  
Fission Products ◽  
Severe Accident ◽  
Cooperative Research ◽  
Best Estimate ◽  
Fuel Bundle ◽  
Calculation Results ◽  
Estimate Method ◽  
Analysis Computer ◽  
Accident Conditions

The PHEBUS-FP program is an outstanding example of an international cooperative research program that is yielding valuable data for validating severe accident analysis computer codes. The main objective of the PHEBUS FPT1 experiment was to study the processes in the overheated reactor core, release of fission products and their subsequent transport and deposition under conditions representative of a severe accident of a Pressurised Water Reactor. The FPT1 test could be divided in the bundle degradation, aerosol, washing and chemistry phases. The objective of this article is the best estimate analysis of the bundle degradation phase. GRS (Germany) best estimate method with the statistic tool SUSA used for uncertainty and sensitivity analysis of calculation results and RELAP/SCDAPSIM code, designed to predict the behaviour of reactor systems during severe accident conditions, was used for the simulation of this test. The RELAP/SCDAPSIM calculation results were compared with the experimental measurements and calculations results, received by employing ICARE module of ASTEC V2 code. The performed analysis demonstrated, that the best estimate method, employing RELAP/SCDAPSIM and SUSA codes, is capable to model main severe accidents phenomena in the fuel bundle during the overheating and melting of reactor core.

3D FEA Elastoplastic Structural Analysis of Hypothetically Cracked Monju FBR Core Support Structures

2014 ◽  
Author(s):  
Daniel Garcia-Rodriguez ◽  
Shinichiro Matsubara
Keyword(s):  
Structural Reliability ◽  
Reactor Core ◽  
Operating Conditions ◽  
Severe Accident ◽  
Support Structures ◽  
Element Analysis ◽  
The Core ◽  
Fea Model ◽  
3D Fea ◽  
Accident Conditions

In this work the structural reliability of the circumferentially cracked core support mount of Monju Fast Breeder Reactor (FBR) is analyzed using Finite Element Analysis (FEA). The 3D shell model employed was derived after detailed evaluation of the core support mount behavior with a specific 3D solid model. First, elastoplastic static analysis results show that, under nominal operating conditions, the overall structure would be able to survive a total loss of the core support mount. Second, using the double elastic slope method it was inferred that earthquake loading integrity could be warranted up to a crack representing more than 50% of the total circumference. Both results highlight the ample primary loading margins taken in the design of Monju’s reactor core support structures. Furthermore, the developed 3D shell FEA model will be applied to study other extreme cases such as those under severe accident conditions.

Creation of Automatic System for Monitoring, Forecasting and Control of Condition of Lead-Bismuth (Lead) Coolant and Surfaces of Circuits of Nuclear Power Plants

2009 ◽  
Author(s):  
P. N. Martynov ◽  
R. Sh. Askhadullin ◽  
A. A. Simakov ◽  
A. Yu. Chaban’ ◽  
M. E. Chernov ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Reactor Core ◽  
Heat Removal ◽  
Severe Accident ◽  
Medium Size ◽  
Safety Level ◽  
Nitride Fuel ◽  
Active Safety Systems ◽  
Accident Conditions

Lead-bismuth coolant is preferable for the medium size reactors, since, in contrast to the sodium coolant, it does not interact with water and air, it is radiation resistant, insignificantly activated and it is not combustible [1]. Combination of natural properties of lead-based coolants, mono-nitride fuel, fast reactor neutronics and design approaches used for the reactor core and heat removal system brings SVBR 75/100 NPP [2] to achieve a new safety level and assures its stability without operation of active safety systems even under severe accident conditions. Analysis of possible sequences of the events even under conditions of such severe accidents as addition of total excess reactivity or all pumps trip accompanied by safety system failure leads to the conclusion on that power unit with SVBR 75/100 reactor plant (RP) has high safety level.

MELCOR Study of VVER-1000 Behavior in Case of Overheated Reactor Core Quenching

2014 ◽  
Author(s):  
Pavlin P. Groudev ◽  
Antoaneta E. Stefanova ◽  
Petya I. Vryashkova
Keyword(s):  
Parametric Study ◽  
Cold Water ◽  
Reactor Core ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Fuel Cladding ◽  
Accident Management ◽  
Management Guidance ◽  
Fuel Behavior ◽  
Accident Conditions

This paper presents the results obtained with the MELCOR computer code from a simulation of fuel behavior in case of severe accident for the VVER-1000 reactor core. The examination is focused on investigation the influence of some important parameters, such as porosity, on fuel behavior starting from oxidation of the fuel cladding, fusion product release in the primary circuit after rupture of the fuel cladding, melting of the fuel and reactor core internals and its further relocation to the bottom of the reactor vessel. In the first analyses are modeled options for investigation of melt blockage and debris during the relocation. In the performed analyses are investigated the uncertainty margin of reactor vessel failure based on modeling of the reactor core and an investigation of its behavior. For this purposes it have been performed sensitivity analyses for VVER-1000 reactor core with gadolinium fuel type for parametric study the influence of porosity debris bed. The second analyses is focused on investigation of influence of cold water injection on overheated reactor core at different core exit temperatures, based on severe accident management guidance operator actions. For this purpose was simulated the same SBO scenario with injection of cold water by a high pressure pump in cold leg (quenching from the bottom of reactor core) at different core exit temperatures from 1200 °C to 1500 °C. The aim of the analysis is to track the evolution of the main parameters of the simulated accident. The work was performed at the Institute for Nuclear Research and Nuclear Energy (INRNE) in the frame of severe accident research. The performed analyses continue the effort in the modeling of fuel behavior during severe accidents such as Station Blackout sequence for VVER-1000 reactors based on parametric study. The work is oriented towards the investigation of fuel behavior during severe accident conditions starting from the initial phase of fuel damaging through melting and relocation of fuel elements and reactor internals until the late in-vessel phase, when melt and debris are relocated almost entirely on the bottom head of the reactor vessel. The received results can be used in support of PSA2 as well as in support of analytical validation of Sever Accident Management Guidance for VVER-1000 reactors. The main objectives of this work area better understanding of fuel behavior during severe accident conditions as well as plant response in such situations.

The Gamma Radiation Emitted by a PWR Core Under Severe Accident Conditions

2014 ◽  
Author(s):  
Carsten Brachem ◽  
Jörg Konheiser ◽  
Uwe Hampel
Keyword(s):  
Gamma Radiation ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Feasibility Studies ◽  
Severe Accident ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Radiation Distribution ◽  
University Of Applied Sciences ◽  
Accident Conditions

The gamma radiation emitted by a nuclear reactor core might contain information about the reactor state. This information may be used in a monitoring system for severe accidents. The Technische Universität Dresden and the Zittau/Görlitz University of Applied Sciences are currently carrying out feasibility studies for the development of such a system in a collaborative effort. As one part of such feasibility studies we performed Monte Carlo simulations on a simplified model of a generic pressurized water reactor. For a set of states which represent scenarios of a coolant level decrease and core melt, the gamma radiation distribution outside the reactor pressure vessel has been computed. The results are presented in this paper. They indicate that different coolant levels yield different gamma radiation distributions, and that an accumulation of corium inside the lower head is detectable from the outside.

scholarly journals Simulation of the Westinghouse AP1000 Response to SBLOCA Using RELAP/SCDAPSIM

2014 ◽  
Vol 2014 ◽  
pp. 1-9 ◽  
Author(s):  
Ayah Elshahat ◽  
Timothy Abram ◽  
Judith Hohorst ◽  
Chris Allison
Keyword(s):  
Driving Forces ◽  
Natural Circulation ◽  
Safety System ◽  
Operating Conditions ◽  
Severe Accident ◽  
Passive Safety ◽  
System Component ◽  
Pressurized Water ◽  
Loss Of Coolant Accident ◽  
Accident Conditions

Great interest is given now to advanced nuclear reactors especially those using passive safety components. The Westinghouse AP1000 Advanced Passive pressurized water reactor (PWR) is an 1117 MWe PWR designed to achieve a high safety and performance record. The AP1000 safety system uses natural driving forces, such as pressurized gas, gravity flow, natural circulation flow, and convection. In this paper, the safety performance of the AP1000 during a small break loss of coolant accident (SBLOCA) is investigated. This was done by modelling the AP1000 and the passive safety systems employed using RELAP/SCDAPSIM code. RELAP/SCDAPSIM is designed to describe the overall reactor coolant system (RCS) thermal hydraulic response and core behaviour under normal operating conditions or under design basis or severe accident conditions. Passive safety components in the AP1000 showed a clear improvement in accident mitigation. It was found that RELAP/SCDAPSIM is capable of modelling a LOCA in an AP1000 and it enables the investigation of each safety system component response separately during the accident. The model is also capable of simulating natural circulation and other relevant phenomena. The results of the model were compared to that of the NOTRUMP code and found to be in a good agreement.

Ruthenium release modelling in air and steam atmospheres under severe accident conditions using the MAAP4 code

2012 ◽  
Vol 246 ◽  
pp. 157-162 ◽  
Author(s):  
Emilie Beuzet ◽  
Jean-Sylvestre Lamy ◽  
Hadrien Perron ◽  
Eric Simoni ◽  
Gérard Ducros
Keyword(s):  
Severe Accident ◽  
Accident Conditions
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