scholarly journals Prevention of early containment failure due to high pressure melt ejection and direct containment heating for advanced light water reactors: Advanced Reactor Severe Accident Program

1990 ◽  
Author(s):  
J Carter, III ◽  
M Fontana ◽  
S Additon ◽  
R Summitt ◽  
R Henry ◽  
...  
Keyword(s):  
High Pressure ◽  
Severe Accident ◽  
Light Water Reactors ◽  
Light Water ◽  
Water Reactors ◽  
Direct Containment Heating ◽  
Melt Ejection

scholarly journals Stand-Alone Containment Analysis of the Phébus FPT Tests with the ASTEC and the MELCOR Codes: The FPT-0 Test

2017 ◽  
Vol 2017 ◽  
pp. 1-25 ◽  
Author(s):  
Bruno Gonfiotti ◽  
Sandro Paci
Keyword(s):  
Fission Products ◽  
Severe Accident ◽  
Sensitivity Analyses ◽  
Light Water Reactors ◽  
Control Rod ◽  
Light Water ◽  
Severe Accidents ◽  
Fuel Bundle ◽  
Water Reactors

The integral Phébus tests were probably one of the most important experimental campaigns performed to investigate the progression of severe accidents in light water reactors. In these tests, the degradation of a PWR fuel bundle was investigated employing different control rod materials and burn-up levels in strongly or weakly oxidizing conditions. From the results of such tests, numerical codes such as ASTEC and MELCOR have been developed to describe the evolution of a severe accident. After the termination of the experimental Phébus campaign, these two codes were furthermore expanded. Therefore, the aim of the present work is to reanalyze the first Phébus test (FPT-0) employing the updated ASTEC and MELCOR versions to ensure that the new improvements introduced in such codes allow also a better prediction of these Phébus tests. The analysis focuses on the stand-alone containment aspects of this test, and the paper summarizes the main thermal-hydraulic results and presents different sensitivity analyses carried out on the aerosols and fission products behavior. This paper is part of a series of publications covering the four executed Phébus tests employing a solid PWR fuel bundle: FPT-0, FPT-1, FPT-2, and FPT-3.

Iodine-steel reactions under severe accident conditions in light-water reactors

1996 ◽  
Vol 166 (3) ◽  
pp. 357-365 ◽  
Author(s):  
F. Funke ◽  
G.-U. Greger ◽  
S. Hellmann ◽  
A. Bleier ◽  
W. Morell
Keyword(s):  
Severe Accident ◽  
Light Water Reactors ◽  
Light Water ◽  
Water Reactors ◽  
Accident Conditions

The Impact of Transient Behavior of Corium in the Lower Head of a Reactor Vessel for In-Vessel Melt Retention Strategies

2016 ◽  
Author(s):  
L. Carénini ◽  
F. Fichot
Keyword(s):  
Nuclear Power ◽  
Metal Layer ◽  
Heat Fluxes ◽  
Severe Accident ◽  
Total Power ◽  
Light Water Reactors ◽  
Light Water ◽  
Accident Management ◽  
Water Reactors ◽  
The Impact

One of the main goals of severe accident management strategies is to mitigate radiological releases to people and environment. To choose the most appropriate strategy, one needs to know the probability of its success taking into account the associated uncertainties. In the field of corium and debris behavior and coolability, research programs are still on going and the possibilities to efficiently cool and retain corium and debris inside the Reactor Pressure Vessel (RPV) then inside the containment are difficult to evaluate. This leads to uncertainties in safety assessments particularly when margins to RPV or containment failure are too weak. In Vessel Melt Retention (IVMR) strategies for Light Water Reactors (PWR, BWR, VVER) intend to stabilize and retain the core melt in the RPV (as it happened during the TMI-2 accident). This would reduce significantly the threats to the last barrier (the containment) and therefore reduce the risk of release of radioactive elements to the environment. This type of Severe Accident Management (SAM) strategy has already been incorporated recently in the SAM guidance (SAMG) of several operating medium size Light Water Reactors (reactor below 500MWe (like VVER440)) and is part of the SAMG strategies for some Gen III+ PWRs of higher power like the AP1000. A European project coordinated by IRSN and gathering 23 organizations (Utilities, Technical Support Organizations, Nuclear Power Plant vendors, Research Institutes…) has been launched in 2015 with as main objective the evaluation of feasibility of IVMR strategies for Light Water Reactors (PWR, VVER, BWR) of total power around 1000MWe (which represent a significant part of the European Nuclear Power Plants fleet). This paper intends to show how it is possible to introduce transient evolutions of the stratified corium pool in the evaluation of the heat flux profile along the vessel wall. Indeed, due to chemical reactions in the U–Zr–O–Fe molten pool, separation between non-miscible metallic and oxide phases may occur, modifying the thermal load applied to the RPV. If stabilized stratified corium configurations are well defined and modeled, transient evolutions of material layers in the corium pool are still difficult to predict. The evaluations presented are based on calculations performed with the severe accident integral code ASTEC (Accident Source Term Evaluation Code) for a typical PWR reactor. The modeling of transient evolution of corium layers leads to configurations with a thin light metal layer on top of the oxidic one, increasing the so called “focusing effect” (intense heat fluxes on the RPV walls adjacent to the top metal layer). A sensitivity study on some uncertain parameters is proposed to evaluate the impact on the kinetics of layers inversion. Depending on the duration of these transient heat fluxes, the mechanical strength of the RPV could be challenged.

Modelling of QUENCH 03 and QUENCH 06 Experiments Using RELAP/SCDAPSIM Code

2013 ◽  
Author(s):  
Tadas Kaliatka ◽  
Eugenijus Ušpuras ◽  
Algirdas Kaliatka
Keyword(s):  
Severe Accident ◽  
Light Water Reactors ◽  
Core Flooding ◽  
Case Scenario ◽  
Worst Case ◽  
Light Water ◽  
Input Parameters ◽  
Calculation Results ◽  
Water Reactors ◽  
Quench Test

An important accident management measure for controlling severe accident transients in Light Water Reactors is the injection of water to cool the degrading core. Flooding of the overheated core, which causes quenching of the fuel rods, is considered a worst-case scenario regarding hydrogen generation rates which should not exceed safety-relevant critical values. Within the frame of the QUENCH test-program the loss of coolant accidents with the following flooding of overheated core in Light Water Reactors is analysed using an experimental facility. The modelling of QUENCH-03 and QUENCH-06 experiments was performed with RELAP/SCDAPSIM computer code. The observed calculation results showed that thermal properties of shroud materials (heat losses through the shroud) and electrical power of fuel simulators are the main source of uncertainty in the calculations. The main idea of this article is modification of input parameters to receive the best agreement with the measurements for the selected QUENCH test. Modified input parameters are used in the input deck for another QUENCH test. The good agreement between calculation results and measurements of both QUENCH tests demonstrated the correctness of modified parameters and legitimacy with the real physical processes.

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