Investigation of fission products distribution effected by spray system in PWR containment during severe accidents

2022 ◽  
Vol 169 ◽  
pp. 108940
Author(s):  
Hongping Sun ◽  
Jian Deng ◽  
Yuejian Luo ◽  
Ming Zhang ◽  
Youyou Xu ◽  
...  
Keyword(s):  
Fission Products ◽  
Spray System ◽  
Severe Accidents

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 Numerical Simulation of Decontamination of Airborne Fission Products during In-Vessel Release Phase by Containment Spray

2018 ◽  
Vol 2018 ◽  
pp. 1-19
Author(s):  
Khurram Mehboob
Keyword(s):  
Droplet Size ◽  
Failure Time ◽  
Ph Value ◽  
Removal Rate ◽  
Droplet Diameter ◽  
Fission Products ◽  
Product Release ◽  
Spray System ◽  
Spray Solution ◽  
Proposed Model

The containment spray system (CSS) has a significant role in limiting the risk of radioactive exposure to the environment. In this work, the optimal droplet size and pH value of spray water to prevent the fission product release have been evaluated to improve the performance of the spray system during in-vessel release phase. A semikinetic model has been developed and implemented in MATLAB. The sensitivity and removal rate of airborne isotopes with the spray system have been simulated versus the spray activation and failure time, droplet size, and pH value. The alkaline (Na2S2O3) spray solution and spray water with pH 9.5 have similar scrubbing properties for iodine. However, the removal rate from the CSS has been found to be an approximately inverse square of droplet diameter (1/d2) for Na2S2O3 and higher pH of spray water. The numerical results showed that 450 μm–850 μm droplet with 9.5 pH and higher or the alkaline (Na2S2O3) solution with 0.2 m3/s–0.35 m3/s flow rate is optimal for effective scrubbing of in-containment fission products. The proposed model has been validated with TOSQAN experimental data.

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.

scholarly journals Modeling of Spray System Operation under Hydrogen and Steam Emissions in NPP Containment during Severe Accident

2011 ◽  
Vol 2011 ◽  
pp. 1-13 ◽  
Author(s):  
Vadim E. Seleznev ◽  
Vladimir V. Aleshin ◽  
Sergey N. Pryalov
Keyword(s):  
Boundary Conditions ◽  
Nuclear Power ◽  
Dirichlet Boundary ◽  
Dirichlet Boundary Conditions ◽  
Severe Accident ◽  
System Operation ◽  
Spray System ◽  
Severe Accidents ◽  
Full Cross Section ◽  
Reactor Building

The paper describes one of the variants of mathematical models of a fluid dynamics process inside the containment, which occurs in the conditions of operation of spray systems in severe accidents at nuclear power plant. The source of emergency emissions in this case is the leak of the coolant or rupture at full cross-section of the main circulating pipeline in a reactor building. Leak or rupture characteristics define the localization and the temporal law of functioning of a source of emergency emission (or accrued operating) of warmed up hydrogen and steam in the containment. Operation of this source at the course of analyzed accident models should be described by the assignment of the relevant Dirichlet boundary conditions. Functioning of the passive autocatalytic recombiners of hydrogen is described in the form of the complex Newton boundary conditions.

Numerical Simulation of the Behavior of Fission Products in the Primary Circuit of the VVER During the LOCA Severe Accident

2009 ◽  
Author(s):  
S. V. Tsaun ◽  
V. V. Bezlepkin ◽  
A. E. Kiselev ◽  
I. A. Potapov ◽  
V. F. Strizhov ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Hydrogen Generation ◽  
Fission Products ◽  
Severe Accident ◽  
Primary Circuit ◽  
Decay Heat ◽  
The Core ◽  
Severe Accidents ◽  
Loss Of Coolant ◽  
Mass Activity

The methods and models for the analysis of the radiological consequences of the design basis and severe accidents in a Nuclear Power Plant (NPP) are presented in this paper when using the system code SOCRAT. The system code SOCRAT/V3 was elaborated for a realistic analysis of radiological consequences of severe accidents in a NPP. The following models of the fission products (FP) behavior are included into the code SOCRAT/V3: (i) the condensation and the evaporation of the FP in the gaseous phase and (ii) the sedimentation, the evaporation, and the coagulation of the aerosol-shape FP. The latter processes are governed by gravity, Brownian and turbulent diffusion, thermophoresis, turbophoresis and so forth. The behavior of the FP during the loss-of-coolant accidents (LOCA) is presented to demonstrate the possibilities of the code SOCRAT/V3. The main stages of the accident (the core dryout, the core reflooding, the core degradation, the hydrogen generation, the FP release, etc.) are described. Corresponding estimations of the mass, activity, and decay heat of the suspended, settled and released into containment the FP (Xe, Te, Cs, CsI, Cs2MoO4, and so forth) are represent as well.

scholarly journals Analysis of Core Degradation and Fission Products Release in Phebus FPT1 Test at IRSN by Detailed Severe Accidents Analysis Code, IMPACT/SAMPSON

2003 ◽  
Vol 40 (8) ◽  
pp. 591-603 ◽  
Author(s):  
Takashi IKEDA ◽  
Masafumi TERADA ◽  
Hidetoshi KARASAWA ◽  
Katsuhiko NAKAHARA ◽  
Makoto YAMAGISHI
Keyword(s):  
Fission Products ◽  
Severe Accidents

Study on mitigation of in-vessel release of fission products in severe accidents of PWR

2010 ◽  
Vol 240 (11) ◽  
pp. 3888-3897 ◽  
Author(s):  
G.F. Huang ◽  
L.L. Tong ◽  
J.X. Li ◽  
X.W. Cao
Keyword(s):  
Fission Products ◽  
Severe Accidents

Effects of Filtered Containment Venting on Fission Product Releases During CANDU Reactor Severe Accidents

2017 ◽  
Vol 3 (2) ◽  
Author(s):  
A. C. Morreale ◽  
L. S. Lebel ◽  
M. J. Brown
Keyword(s):  
Fission Product ◽  
Noble Gas ◽  
Fission Products ◽  
Nuclear Industry ◽  
Severe Accident ◽  
Accident Analysis ◽  
Employee Safety ◽  
Severe Accidents ◽  
Loss Of Coolant Accident ◽  
Containment Integrity

Severe accidents are of increasing concern in the nuclear industry worldwide since the accidents at Fukushima Daiichi (March 2011). These events have significant consequences that must be mitigated to ensure public and employee safety. Filtered containment venting (FCV) systems are beneficial in this context as they would help to maintain containment integrity while also reducing radionuclide releases to the environment. This paper explores the degree to which filtered containment venting would reduce fission product releases during two Canada Deuterium Uranium (CANDU) 6 severe accident scenarios, namely a station blackout (SBO) and a large loss of coolant accident (LLOCA) (with limited emergency cooling). The effects on the progression of the severe accident and radionuclide releases to the environment are explored using the Modular Accident Analysis Program (MAAP)–CANDU integrated severe accident analysis code. The stylized filtered containment venting system model employed in this study avoids containment failure and significantly reduces radionuclide releases by 95–97% for non-noble gas fission products. Filtered containment venting is shown to be a suitable technology for the mitigation of severe accidents in CANDU, maintaining containment integrity and reducing radionuclide releases to the environment.

New Approach for Environmental Protection During Severe Accidents at Nuclear Power Plants

2009 ◽  
Author(s):  
S. A. Kulyukhin ◽  
N. B. Mikheev ◽  
L. N. Falkovskii ◽  
L. A. Reshetov ◽  
M. Ya. Zvetkova ◽  
...  
Keyword(s):  
Environmental Protection ◽  
Nuclear Power ◽  
Power Plants ◽  
Heat Removal ◽  
Passive Mode ◽  
Pressure Release ◽  
Spray System ◽  
Severe Accidents ◽  
Heat Removal System ◽  
New Generation

This paper reports new approaches for increasing environmental protection during severe accidents at NPPs. For NPPs with two protective shells and pressure release system such as WWER-1000 we suggest a new comprehensive, passive-mode environmental protection system of decontamination of the radioactive air-steam mixture from the containment and the intercontainment area, which includes the “wet” stage (scrubbers, etc.), the “dry” stage (sorption module), and also an ejector, which in a passive mode is capable of solving the multi-purpose task of decontamination of the air-steam mixture. For Russian WWER-440/V-230 NPPs we suggest three protection levels: 1) a jet-vortex condenser; 2) the spray system; 3) a sorption module. For modern designs of new generation NPPs, which do not provide for pressure release systems, we proposed a new passive filtering system together with the passive heat-removal system, which can be used during severe accidents in case all power supply units become unavailable.

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