Simeco Tests in a Melt Stratified Pool

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.

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

Fluids ◽  
2019 ◽  
Vol 4 (2) ◽  
pp. 75 ◽  
Author(s):  
Leiv Storesletten ◽  
D. Andrew S. Rees

The onset of convection in an inclined porous layer which is heated internally by a uniform distribution of heat sources is considered. We investigate the combined effects of inclination, anisotropy and internal heat generation on the linear instability of the basic parallel flow. When the Rayleigh number is sufficiently large, instability occurs and a convective motion is set up. It turns out that the preferred motion at convection onset depends quite strongly on the anisotropy ratio, ξ , and the inclination angle. When ξ < 1 the preferred motion is in the form of longitudinal rolls for all inclinations. When ξ > 1 transverse rolls are preferred for small inclinations but, at high inclinations, longitudinal rolls are preferred. At intermediate inclinations the preferred roll orientation varies smoothly between these two extremes.


2018 ◽  
Vol 4 (4) ◽  
Author(s):  
Jianfeng Mao ◽  
Shiyi Bao ◽  
Zhiming Lu ◽  
Lijia Luo ◽  
Zengliang Gao

The so-called in-vessel retention (IVR) was considered as a severe accident management strategy and had been certified by Nuclear Regulatory Commission (NRC) in U.S. as a standard measure for severe accident management since 1996. In the core meltdown accident, the reactor pressure vessel (RPV) integrity should be ensured during the prescribed time of 72 h. However, in traditional concept of IVR, several factors that affect the RPV failure were not considered in the structural safety assessment, including the effect of corium crust on the RPV failure. Actually, the crust strength is of significant importance in the context of a severe reactor accident in which molten core material melts through the reactor vessel and collects on the lower head (LH) of the RPV. Consequently, the RPV integrity is significantly influenced by the crust. A strong, coherent crust anchored to the RPV walls could allow the yet-molten corium to fall away from the crust as it erodes the RPV, therefore thermally decoupling the melt pool from the coolant and sharply reducing the cooling rate. Due to the thermal resistance of the crust layer, it somewhat prevents further attack of melt pool from the RPV. In the present study, the effect of crust on RPV structural behaviors was examined under multilayered crust formation conditions with consideration of detailed thermal characteristics, such as high-temperature gradient across the wall thickness. Thereafter, systematic finite element analyses and subsequent damage evaluation with varying parameters were performed on a representative RPV to figure out the possibility of high temperature induced failures with the effect of crust layer.


Author(s):  
Seung D. Lee ◽  
Hyoung M. Son ◽  
Kune Y. Suh ◽  
Joy L. Rempe ◽  
F. Bill Cheung ◽  
...  

Natural convection plays an important role in determining the thermal load from debris accumulated in the reactor vessel lower head during a severe accident in nuclear power plants. The natural convection heat transfer involving internal heat generation is generally represented by the modified Rayleigh number, Ra’, which quantifies the internal heat source and hence the strength of the buoyancy force. The SIGMA RP (Simulant Internal Gravitated Material Apparatus Rectangular Pool) tests are concerned with high Ra’ turbulent natural convection in a rectangular pool. The internal heating was realized by cable-type heaters in the test to simulate uniform heat generation. Ra’ was found to be in the range of 1×109–1×1012. Experiments were conducted using air with the Prandtl number, Pr, of about 0.7 as the working fluid. The test results confirmed feasibility of the direct heating method to simulate uniform volumetric heat generation in the pool.


Author(s):  
Masanori Naitoh ◽  
Marco Pellegrini ◽  
Hiroaki Suzuki ◽  
Hideo Mizouchi ◽  
Hidetoshi Okada

This paper describes analysis results of the early phase accident progression of the Fukushima Daiichi Nuclear Power Plant (NPP) Unit 1 by the severe accident analysis code SAMPSON. The isolation condensers were the only devices for decay heat removal at Unit 1, but they stopped after the loss of AC and DC powers. Since there were no decay heat removal for about 14 hours after their termination until the start of alternative water injection into the core by the fire engine, the core melt and the reactor pressure vessel (RPV) bottom failure occurred resulting in large amount of fission products release into the environment. The original SAMPSON was improved by adding new modellings for the phenomena which have been deemed specific to the Fukushima Daiichi NPP: (1) deterioration of SRV gaskets and (2) buckling of in-core-monitor housings which caused the early steam leakage from the core into the drywell, and (3) melt of the in-core-monitor housings in the lower plenum of the RPV. The analysis results showed that (1) 55.3% of UO2 of the initial loading and 66.1% of the core material including UO2, zircaloy, steel and control materials had melted down into the pedestal of the drywell, (2) the amount of Hydrogen generated by Zr-H2O reaction was 686 kg, (3) amount of Cs element released from fuels was 61 kg which was 72% of the total Cs element which was included in fuels at the initiation of the accident, and (4) 18.3% of the corium which fell into the pedestal was one large lump and the 81.7% was particulate corium.


Author(s):  
Tobias Jankowski ◽  
Marco K. Koch

In case of a postulated severe accident scenario in a nuclear power plant coolant might be released from the cooling circuit into the containment atmosphere. Due to the pressure difference between the cooling circuit and the containment, the fluid will be released gaseous and might contain fission products, depending on the severe accident scenario. The effluent vapour might condensate on the colder containment structures and flow down to the floor, where water pools will be formed. Because of the decay heat of the fission products or due to a pressure decrease in the containment, which might be induced by the operation of filtered containment venting systems, the water pools could start boiling. During this boiling process gas bubbles will rise to the water pool surface and release former solved fission products by bubble bursting into the containment atmosphere. Those released fission products will dry due to their decay heat and remain as very fine particles for a long time in the containment atmosphere. In the context of filtered containment venting system operation a detailed knowledge of those particles is among others of interest for the development of e. g. accident management measures. A correlation for the estimation of the so-called droplet re-entrainment, which is defined as the ratio of the droplet mass flow, released from the bursting bubbles, and the gas flow, which streams through the effective pool surface, is developed and will be validated based on several single effect tests. The considered tests were conducted in test facilities of different scale and investigated the influence of different parameters on the release of aerosols by the effect of droplet re-entrainment. The calculation results are in a good agreement with the experimental data, indicating that parameters with a relevant influence on the droplet re-entrainment are considered in the developed correlation with a suitable impact.


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