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.

Influence of Boron Carbide on Core Degradation During Severe Accidents in LWRs

2012 ◽  
Author(s):  
Martin Steinbrück
Keyword(s):  
Boron Carbide ◽  
Nuclear Power ◽  
Source Term ◽  
Current Knowledge ◽  
Steel Structures ◽  
Severe Accident ◽  
Control Rod ◽  
Severe Accidents ◽  
Fuel Bundle ◽  
Water Reactors

Boron carbide (B4C) is widely used as neutron absorbing control rod material in light water reactors (LWRs). It was also applied in all units of the Fukushima Dai-ichi nuclear power plant. Although the melting temperature of B4C is 2450 °C, it initiates local, but significant melt formation in the core at temperatures around 1250 °C due to eutectic interactions with the surrounding steel structures. The B4C containing melt relocates and hence transports material and energy to lower parts of the fuel bundle. It is chemically aggressive and may attack other structure materials. Furthermore, the absorber melt is oxidized by steam very rapidly and thus contributes to the hydrogen source term in the early phase of a severe accident. After failure of the control rod cladding B4C reacts with the oxidizing atmosphere. This reaction produces CO, CO2, boron oxide and boric acids, as well as significant amount of hydrogen. It is strongly exothermic, thus causing considerable release of energy. No or only insignificant formation of methane was observed in all experiments with boron carbide. The paper will summarize the current knowledge on boron carbide behavior during severe accidents, and will try, also in the light of the Fukushima accidents, to draw some common conclusions on the behavior of B4C during severe accidents with the main focus on the consequences for core degradation and hydrogen source term.

ASTEC: An Integral Code for Simulation of Severe Light Water Reactor Accidents

2006 ◽  
Author(s):  
N. Reinke ◽  
K. Neu ◽  
H.-J. Allelein
Keyword(s):  
Fission Products ◽  
Feasibility Studies ◽  
Current Status ◽  
Light Water ◽  
Fast Running ◽  
Severe Accidents ◽  
Loss Of Coolant Accident ◽  
Station Blackout ◽  
Water Reactors ◽  
Reactor Types

The integral code ASTEC (Accident Source Term Evaluation Code) commonly developed by IRSN and GRS is a fast running programme, which allows the calculation of entire sequences of severe accidents (SA) in light water reactors from the initiating event up to the release of fission products into the environment, thereby covering all important in-vessel and containment phenomena. Thus, the main fields of ASTEC application are intended to be accident sequence studies, uncertainty and sensitivity studies, probabilistic safety analysis level 2 studies as well as support to experiments. The modular structure of ASTEC allows running each module independently and separately, e.g. for separate effects analyses, as well as a combination of multiple modules for coupled effects testing and integral analyses. Among activities concentrating on the validation of individual ASTEC modules describing specific phenomena, the applicability to reactor cases marks an important step in the development of the code. Feasibility studies on plant applications have been performed for several reactor types such as the German Konvoi PWR 1300, the French PWR 900, and the Russian VVER-1000 and −440 with sequences like station blackout, small- or medium-break loss-of-coolant accident, and loss-of-feedwater transients. Subject of this paper is a short overview on the ASTEC code system and its current status with view to the application to severe accidents sequences at several PWRs, exemplified by selected calculations.

scholarly journals Steam Oxidation of Silicon Carbide at High Temperatures for the Application as Accident Tolerant Fuel Cladding, an Overview

Thermo ◽  
2021 ◽  
Vol 1 (2) ◽  
pp. 151-167
Author(s):  
Hai V. Pham ◽  
Masaki Kurata ◽  
Martin Steinbrueck
Keyword(s):  
Silicon Carbide ◽  
High Temperatures ◽  
Nuclear Power ◽  
Steam Oxidation ◽  
Severe Accident ◽  
Light Water Reactors ◽  
Severe Accidents ◽  
Potential Benefits ◽  
Water Reactors ◽  
Accident Conditions

Since the nuclear accident at Fukushima Daiichi Nuclear Power Station in 2011, a considerable number of studies have been conducted to develop accident tolerant fuel (ATF) claddings for safety enhancement of light water reactors. Among many potential ATF claddings, silicon carbide is one of the most promising candidates with many superior features suitable for nuclear applications. In spite of many potential benefits of SiC cladding, there are some concerns over the oxidation/corrosion resistance of the cladding, especially at extreme temperatures (up to 2000 °C) in severe accidents. However, the study of SiC steam oxidation in conventional test facilities in water vapor atmospheres at temperatures above 1600 °C is very challenging. In recent years, several efforts have been made to modify existing or to develop new advanced test facilities to perform material oxidation tests in steam environments typical of severe accident conditions. In this article, the authors outline the features of SiC oxidation/corrosion at high temperatures, as well as the developments of advanced test facilities in their laboratories, and, finally, give some of the current advances in understanding based on recent data obtained from those advanced test facilities.

Computational Study for Inventory Estimation of Se-79, Tc-99, Sn-126, and Cs-135 in High-Level Radioactive Wastes From Spent Nuclear Fuels of Light Water Reactors

2011 ◽  
Author(s):  
Keisuke Okumura ◽  
Shiho Asai ◽  
Yukiko Hanzawa ◽  
Tsutomu Okamoto ◽  
Hideya Suzuki ◽  
...  
Keyword(s):  
Parametric Study ◽  
Computational Study ◽  
Estimation Method ◽  
Fission Products ◽  
Radioactive Wastes ◽  
Light Water Reactors ◽  
Nuclear Fuels ◽  
Light Water ◽  
Water Reactors ◽  
High Level

Inventory estimation of long-lived fission products (LLFPs) in high-level radioactive wastes (HLW) from spent nuclear fuels of light water reactors is important for a safety assessment of their disposal. In order to develop an inventory estimation method of difficult-to-measure LLFPs (Se-79, Tc-99, Sn-126, and Cs-135), a parametric study was carried out by using a sophisticated burnup calculation code and data. In the parametric study, fuel specifications and irradiation conditions are changed in the conceivable range. The considered parameters are fuel assembly types (PWR / BWR), U-235 enrichment, moderator temperature, void fraction, power density, and so on. From the calculated results, we clarify the burnup characteristics of the target LLFPs and their possible ranges of generations. Finally, candidates of the key nuclide are proposed for the scaling factor method of HLW.

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
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