Representation of the Fission Products Behavior during Severe Accidents into a Global Experiment: Dimensioning of the Phebus RR Experimental Set-Up

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.

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

Science and Technology of Nuclear Installations ◽

10.1155/2017/1450648 ◽

2017 ◽

Vol 2017 ◽

pp. 1-25 ◽

Cited By ~ 1

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.

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Use of fission products in a pilot set-up to initiate the oxidation of paraffin

The International Journal of Applied Radiation and Isotopes ◽

10.1016/0020-708x(59)90337-0 ◽

1959 ◽

Vol 7 (1) ◽

pp. 73

Author(s):

G. Icanid ◽

A. Dragut ◽

J. Drimus ◽

V. Dumitresco ◽

I. Stoian

Keyword(s):

Fission Products ◽

Set Up

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Approaches to the Establishment and Justification of the Size of Emergency Planning Areas around NPPs

Nuclear and Radiation Safety ◽

10.32918/nrs.2020.1(85).01 ◽

2020 ◽

pp. 4-14

Author(s):

V. Bogorad ◽

O. Slepchenko ◽

I. Kalyta ◽

T. Lytvynska ◽

S. Chupryna ◽

...

Keyword(s):

Emergency Planning ◽

Nuclear Facilities ◽

Emergency Response System ◽

Material Resources ◽

Severe Accidents ◽

Radionuclide Composition ◽

Monitoring Zone ◽

Set Up ◽

Effects Of Radiation ◽

Definition Of

Emergency planning zones around facilities that use radiation and nuclear technologies are set up in order to prevent or minimize deterministic effects and reduce risks of stochastic effects of radiation exposure in case of nuclear or radiological emergency. In international practice, in order to enhance effectiveness of response to emergencies, it is common to set up emergency planning zones for nuclear facilities in accordance with their level of threat for the environment and public. As the establishment of such zones requires certain material resources, the size of the zones is to be justified. As of today, sanitary-protective zone and monitoring zone in Ukraine could be considered, to a certain extent, as analogues of such emergency planning zones. However, the main functional load of these zones are relevant more for routine operation of nuclear facilities than for the issues of emergency planning and do not have such a specific for the emergency planning zones functions as ensuring necessary infrastructure for conducting such urgent protective actions as evacuation, sheltering, notification, relocation and others in case of severe accidents at NPPs. Absence of transparent numerical criteria for determining the size of emergency planning areas, on one hand, and necessity to determine such areas as required by the key IAEA publication GSR Part 7, as well as due to understanding the importance of emergency planning areas in the state emergency response system, on the other hand, make the task on establishing approaches to the definition of the scope of emergency release, which is a framework for emergency zoning, crucial. The paper proposes to discuss two approaches to defining the scope of the boundary release, according to which the size of the emergency planning areas is to be defined. The paper presents results of assessing the size of the emergency planning areas around NPPs based on the scope and radionuclide composition of the boundary release.

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AN ASSOCIATION RULE MINING MODEL FOR THE ASSESSMENT OF THE CORRELATIONS BETWEEN THE ATTRIBUTES OF SEVERE ACCIDENTS

Journal of Civil Engineering and Management ◽

10.3846/jcem.2020.12316 ◽

2020 ◽

Vol 26 (4) ◽

pp. 315-330 ◽

Cited By ~ 2

Author(s):

Bilal Umut Ayhan ◽

Neşet Berkay Doğan ◽

Onur Behzat Tokdemir

Keyword(s):

Analytical Model ◽

Association Rule ◽

Association Rule Mining ◽

Factual Knowledge ◽

Rule Mining ◽

Preventative Measures ◽

Severe Accidents ◽

Set Up ◽

Construction Accidents ◽

Computational Systems

Identifying the correlations between the attributes of severe accidents could be vital to preventing them. If such relationships were known dynamically, it would be possible to take preventative actions against accidents. The paper aims to develop an analytical model that is adaptable for each type of data to create preventative measures that will be suitable for any computational systems. The present model collectively shows the relationships between the attributes in a coherent manner to avoid severe accidents. In this respect, Association Rule Mining (ARM) is used as the technique to identify the correlations between the attributes. The research adopts a positivist approach to adhere to the factual knowledge concerning nine different accident types through case studies and quantitative measurements in an objective nature. ARM was exemplified with nine different types of construction accidents to validate the adaptability of the proposed model. The results show that each accident type has different characteristics with varying combinations of the attribute, and analytical model accomplished to accommodate variation through the dataset. Ultimately, professionals can identify the cause-effect relationships effectively and set up preventative measures to break the link between the accident causing factors.

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

Volume 2: Structural Integrity; Safety and Security; Advanced Applications of Nuclear Technology; Balance of Plant for Nuclear Applications ◽

10.1115/icone17-75159 ◽

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.

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

Journal of Nuclear Science and Technology ◽

10.1080/18811248.2003.9715396 ◽

2003 ◽

Vol 40 (8) ◽

pp. 591-603 ◽

Cited By ~ 8

Author(s):

Takashi IKEDA ◽

Masafumi TERADA ◽

Hidetoshi KARASAWA ◽

Katsuhiko NAKAHARA ◽

Makoto YAMAGISHI

Keyword(s):

Fission Products ◽

Severe Accidents

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Failure Fraction Calculation of the TRISO-Coated Particle Using X-Ray Computed Tomography

Science and Technology of Nuclear Installations ◽

10.1155/2021/2420526 ◽

2021 ◽

Vol 2021 ◽

pp. 1-6

Author(s):

Libing Zhu ◽

Jianxun Zhao ◽

Xincheng Xiang ◽

Yu Zhou ◽

Xiangang Wang

Keyword(s):

Fission Products ◽

Geometrical Parameters ◽

Geometrical Shape ◽

Performance Simulation ◽

X Ray ◽

The Real ◽

Coated Particle ◽

Fission Gas ◽

X Ray Computed ◽

Set Up

The geometrical shape of the TRISO-coated particle is closely related to its performance and safety. In this paper, models were set up to study the failure fraction of TRISO particle, considering the real asphericity induced by manufacturing uncertainties. TRISO is simplified as a pressure vessel model, and micro X-ray CT was employed to detect the real geometrical shape. Key geometrical parameters, thickness and volume of the real particle, were then obtained with the 3D measurement method and input into PANAMA code (a German code for fuel performance simulation). Release fraction of fission gas and failure fraction of the TRISO-coated particle were revised with the aforementioned parameters with more accuracy and compared with those of the spherical particle. Obvious increment of failure fraction of the particle is found, which may contribute to the release of fission products.

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