A Model of the Fission Products Release from the Melt Pool during Severe Accident in a Liquid Metal Cooled Fast Reactor

2021 ◽  
Vol 68 (2) ◽  
pp. 152-158
Author(s):  
E. V. Usov ◽  
V. I. Chukhno ◽  
I. A. Klimonov ◽  
V. D. Ozrin ◽  
N. A. Mosunova ◽  
...  
Author(s):  
A. A. Gubaidullin ◽  
B. R. Sehgal

In the last phase of the core degradation, an oxidic melt pool of mainly UO2, ZrO2, and unoxidized Zircaloy and stain-less steel will form in the lower head of the RPV (Theofanous et al., 1996). A molten metal layer (composed mainly of Fe and Zr) will rest on the top of the crust of the oxidic pool. A thin oxidic crust layer of frozen core material is formed on the vessel’s inside wall. In this bounding configuration, thermal loads to the RPV walls are determined by natural convection heat transfer driven by internal heat sources. Decay heat from fission products is assumed to be generated uniformly within the oxidic pool and generally no heat generation is considered in the upper metallic layer. For example, in a hypothetical severe accident scenario for an AP600-like reactor, the following values can be expected: volumetric heat generation Qv ∼ 1 MW/m3, volume of the oxidic pool V ∼ 10 m3, radius R = 2 m, temperatures in the oxidic pool T ∼ 2700°C, temperatures in the metal layer T ∼ 2000°C, maximum depth ratio of the metal layer to the oxidic pool L12 ∼ 0:3, properties of the oxidic pool, depending on melt composition, as characterized by the Prandtl number, Pr ∼ 0:6, properties of the metallic layer Pr < 0:1, the intensity of convective motion, as characterized by the Rayleigh number, Ra ∼ 1015–1016 (Theofanous et al., 1996). The time scale of core melt pool formation is estimated as 1/2 to 1 hour (Sehgal, 1999). Indeed, these estimates could vary, depending very much on the accident scenario and the type of reactor.


Author(s):  
Bruno Gonfiotti ◽  
Sandro Paci

The estimation of Fission Products (FPs) release from the containment system of a nuclear plant to the external environment during a Severe Accident (SA) is a quite complex task. In the last 30–40 years several efforts were made to understand and to investigate the different phenomena occurring in such a kind of accidents in the primary coolant system and in the containment. These researches moved along two tracks: understanding of involved phenomenologies through the execution of different experiments, and creation of numerical codes capable to simulate such phenomena. These codes are continuously developed to reflect the actual SA state-of-the-art, but it is necessary to continuously check that modifications and improvements are able to increase the quality of the obtained results. For this purpose, a continuous verification and validation work should be carried out. Therefore, the aim of the present work is to re-analyze the Phébus FPT-1 test employing the ASTEC (F) and MELCOR (USA) codes. The analysis focuses on the stand-alone containment aspects of the test, and three different modellisations of the containment vessel have been developed showing that at least 15/20 Control Volumes (CVs) are necessary for the spatial schematization to correctly predict thermal-hydraulics and the aerosol behavior. Furthermore, the paper summarizes the main thermal-hydraulic results, and presents different sensitivity analyses carried out on the aerosols and FPs behavior.


2018 ◽  
Vol 2018 ◽  
pp. 1-12
Author(s):  
Taeseok Kim ◽  
Wonjun Choi ◽  
Joongoo Jeon ◽  
Nam Kyung Kim ◽  
Hoichul Jung ◽  
...  

During a hypothesized severe accident, a containment building is designed to act as a final barrier to prevent release of fission products to the environment in nuclear power plants. However, in a bypass scenario of steam generator tube rupture (SGTR), radioactive nuclides can be released to environment even if the containment is not ruptured. Thus, thorough mitigation strategies are needed to prevent such unfiltered release of the radioactive nuclides during SGTR accidents. To mitigate the consequence of the SGTR accident, this study was conducted to devise a conceptual approach of installing In-Containment Relief Valve (ICRV) from steam generator (SG) to the free space in the containment building and it was simulated by MELCOR code for numerical analysis. Simulation results show that the radioactive nuclides were not released to the environment in the ICRV case. However, the containment pressure increased more than the base case, which is a disadvantage of the ICRV. To minimize the negative effects of the ICRV, the ICRV linked to Reactor Drain Tank (RDT) and cavity flooding was performed. Because the overpressurization of containment is due to heat of ex-vessel corium, only cavity flooding was effective for depressurization. The conceptual design of the ICRV is effective in mitigating the SGTR accident.


2021 ◽  
Vol 7 (4) ◽  
pp. 319-325
Author(s):  
Anastasiya V. Dragunova ◽  
Mikhail S. Morkin ◽  
Vladimir V. Perevezentsev

To timely detect failed fuel elements, a reactor plant should be equipped with a fuel cladding tightness monitoring system (FCTMS). In reactors using a heavy liquid-metal coolant (HLMC), the most efficient way to monitor the fuel cladding tightness is by detecting gaseous fission products (GFP). The article describes the basic principles of constructing a FCTMS in liquid-metal-cooled reactors based on the detection of fission products and delayed neutrons. It is noted that in a reactor plant using a HLMC the fuel cladding tightness is the most efficiently monitored by detecting GFPs. The authors analyze various aspects of the behavior of fission products in a liquid-metal-cooled reactor, such as the movement of GFPs in dissolved and bubble form along the circuit, the sorption of volatile FPs in the lead coolant (LC) and on the surfaces of structural elements, degassing of the GFPs dissolved in the LC, and filtration of cover gas from aerosol particles of different nature. In addition, a general description is given of the conditions for the transfer of GFPs in a LC environment of the reactor being developed. Finally, a mathematical model is presented that makes it possible to determine the calculated activity of reference radionuclides in each reactor unit at any time after the fuel element tightness failure. Based on this model, methods for monitoring the fuel cladding tightness by the gas activity in the gas volumes of the reactor plant will be proposed.


2016 ◽  
Vol 298 ◽  
pp. 218-228 ◽  
Author(s):  
E. Merzari ◽  
P. Fischer ◽  
H. Yuan ◽  
K. Van Tichelen ◽  
S. Keijers ◽  
...  

Author(s):  
Jiayun Wang ◽  
Wei Lu ◽  
Pei Wen Gu

IVR (In-Vessel Retention) strategy is designed as the key severe accident mitigation feature for CAP1400. This paper studies the core melt and relocation progression, which is the base of the melt pool analysis and assessment in the plenum. The MAAP and CFD code are used together to obtain the main insights of the phenomena during core melting. The MAAP code is adopted to have an overall understanding of the progress with the lumped calculation, while the CFD code is used as the tool to study the local failure of the complex structure such as shroud and barrel with finite element simulation. Based on the analysis, the core will heat up after uncovered, and the upper region will melt first to form the core melt pool, as there is still water exist in the active fuel region at the time of upper part rods melting, the debris would be refrozen to form crust to block the relocation. As the melt pool increasing, the shroud is melt-through from the corner, and melts would drop to fill the gap volume between the shroud and barrel before relocation to lower plenum. Furthermore, the barrel will be melted later and the debris relocation to the lower plenum from the core sideward. The melts will touch the lower core support plate before water in the plenum depleted, which provides large mass of metal to be melted into the pool, avoiding large heat flux to challenge the RPV in the pool forming stage.


2014 ◽  
Vol 64 ◽  
pp. 220-229 ◽  
Author(s):  
Acacia Brunett ◽  
Richard Denning ◽  
Marissa Umbel ◽  
Whitney Wutzler

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