scholarly journals Experimental Investigation of Iodine Removal in a Submerged Venturi Scrubber

2021 ◽  
Vol 12 (1) ◽  
pp. 38
Author(s):  
Jawaria Ahad ◽  
Amjad Farooq ◽  
Masroor Ahmad ◽  
Khalid Waheed ◽  
Kamran Rasheed Qureshi ◽  
...  
Keyword(s):  
Removal Efficiency ◽  
Nuclear Power ◽  
Fission Products ◽  
Severe Accident ◽  
Operating Parameters ◽  
Nuclear Accidents ◽  
Sodium Thiosulphate ◽  
Venturi Scrubber ◽  
Vis Spectroscopy ◽  
Accident Scenarios

Severe nuclear accidents can cause over-pressurization and serious damage to the containment of a nuclear power plant, which can result in the release of radioactivity into the environment. Filtered containment venting systems are a nuclear safety system that is designed to control over-pressurization and prevent radioactive fission products from spreading into the environment in the case of a severe accident. Iodine is one of the most harmful products among this list of fissionable products, as it can cause thyroid cancer. The removal of iodine is very important in order to ensure the safety of people and the environment. Thus, an indigenous lab scale setup of this system was developed at PIEAS to conduct research on iodine removal. It is comprised of a compressor for replicating high-pressure accident scenarios, a heater to keep iodine in a vapor form, a dosing pump for the injection of iodine, and a venturi scrubber, submerged in the scrubbing column, containing a solution of 0.2% sodium thiosulphate and 0.5% sodium hydroxide. Inlet and outlet samples were trapped in 0.1 M KOH solution and analyzed via UV-VIS spectroscopy. Operating parameters play an important role in the working of a venturi scrubber. The throat velocity was varied to determine its influence on the removal efficiency of iodine. An increase in removal efficiency was observed with an increase in throat velocity. A removal efficiency of >99% was achieved, which fulfilled the requirements for FCVS.

scholarly journals A Conceptual Approach to Eliminate Bypass Release of Fission Products by In-Containment Relief Valve under SGTR Accident

2018 ◽  
Vol 2018 ◽  
pp. 1-12
Author(s):  
Taeseok Kim ◽  
Wonjun Choi ◽  
Joongoo Jeon ◽  
Nam Kyung Kim ◽  
Hoichul Jung ◽  
...  
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Fission Products ◽  
Mitigation Strategies ◽  
Severe Accident ◽  
Base Case ◽  
Conceptual Approach ◽  
Relief Valve ◽  
Radioactive Nuclides

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.

Analyzing Droplet Size Distributions Inside a Self-Priming Venturi Scrubber for Filtered Containment Venting Systems

2018 ◽  
Author(s):  
P. Papadopoulos ◽  
T. Lind ◽  
H.-M. Prasser
Keyword(s):  
Droplet Size ◽  
High Speed ◽  
Nuclear Power ◽  
Power Plants ◽  
Gas Flow ◽  
Tube Bundle ◽  
Droplet Size Distribution ◽  
Severe Accident ◽  
Design Element ◽  
Venturi Scrubber

After the accident in the Fukushima Daiichi nuclear power plant, the interest of adding Filtered Containment Venting Systems (FCVS) on existing nuclear power plants to prevent radioactive releases to the environment during a severe accident has increased. Wet scrubbers are one possible design element which can be part of an FCVS system. The efficiency of this scrubber type is thereby depending, among others, on the thermal-hydraulic characteristics inside the scrubber. The flow structure is mainly established by the design of the gas inlet nozzle. The venturi geometry is one of the nozzle types that can be found in nowadays FCVS. It acts in two different steps on the removal process of the contaminants in the gas stream. Downstream the suction opening in the throat of the venturi, droplets are formed by atomization of the liquid film. The droplets are contributing to the capture of aerosols and volatile gases from the mixture coming from the containment. Studies state that the majority of the contaminants is scrubbed within this misty flow regime. At the top of the venturi, the gas stream is injected into the pool. The pressure drop at the nozzle exit leads to the formation of smaller bubbles, thus increasing the interfacial area concentration in the pool. In this work, the flow inside a full-scale venturi scrubber has been optically analyzed using shadowgraphy with a high-speed camera. The venturi nozzle was installed in the TRISTAN facility at PSI which was originally designed to investigate the flow dynamics of a tube rupture inside a full-length scale steam generator tube bundle. The data analysis was focused on evaluating the droplet size distribution and the Sauter mean diameter under different gas flow rates and operation modes. The scrubber was operated in two different ways, submerged and unsubmerged. The aim was to include the effect on the droplet sizes of using the nozzle in a submerged operation mode.

Influence of Gas Properties on Gas-Liquid Two-Phase Flow

2020 ◽  
Author(s):  
Miki Saito ◽  
Taizo Kanai ◽  
Satoshi Nishimura ◽  
Yoshihisa Nishi
Keyword(s):  
Removal Efficiency ◽  
Nuclear Power ◽  
Two Phase Flow ◽  
Wire Mesh ◽  
Severe Accident ◽  
Phase Flow ◽  
Stagnant Water ◽  
Two Phase ◽  
Bubble Characteristics ◽  
Gas Properties

Abstract Understanding the mechanism of fission product (FP) removal by pool scrubbing is essential for improving the prediction accuracy of FP emissions concerning severe accident (SA) in a nuclear power plant. Since FP migrates from a gas-phase to a liquid-phase via a gas-liquid interface, the FP removal efficiency by pool scrubbing is largely affected by the flow regime of gas-liquid two-phase flow. In order to gain a deeper understanding of the influence of gas properties on flow regimes, experiments were performed by injecting helium (He) and nitrogen (N2) gas mixtures of several volumetric ratios through a pool of stagnant water. The result suggests clear effects of gas compositions on gas-liquid two-phase flow, where both void and holdup fractions were found to increase with N2 fraction in the supplied gas. The results were compared with previous studies, and a detailed analysis of bubble characteristics for different compositions of gases was performed using a wire-mesh sensor (WMS). This paper also illustrates further research aspects needed to discuss the effect of its results on FP removal efficiency in a SA, and to acquire comprehensive physics behind such gas property influences on two-phase flow.

scholarly journals Analysis of Release Model Effect in the Transport of Fission Products Simulating the FPT3 Test Using MELCOR 2.1 and MELCOR 2.2

2021 ◽  
Vol 13 (14) ◽  
pp. 7964
Author(s):  
Alain Flores y Flores ◽  
Danilo Ferretto ◽  
Tereza Marková ◽  
Guido Mazzini
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Fission Products ◽  
Nuclear Industry ◽  
Severe Accident ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Accident Simulation ◽  
Experimental Facilities ◽  
The Difference

The severe accident integral codes such as Methods for Estimation of Leakages and Consequences of Releases (MELCOR) are complex tools used to simulate and analyse the progression of a severe accident from the onset of the accident up to the release from the containment. For this reason, these tools are developed in order to simulate different phenomena coupling models which can simulate simultaneously the ThermoHydraulic (TH), the physics and the chemistry. In order to evaluate the performance in the prediction of those complicated phenomena, several experimental facilities were built in Europe and all around the world. One of these facilities is the PHEBUS built by Institut de Radioprotection et de Sûrete Nucléaire (IRSN) in Cadarache. The facility reproduces the severe accident phenomena for a pressurized water reactor (PWR) on a volumetric scale of 1:5000. This paper aims to continue the assessment of the MELCOR code from version 2.1 up to version 2.2 underlying the difference in the fission product transport. The assessment of severe accident is an important step to the sustainability of the nuclear energy production in this period where the old nuclear power plants are more than the new reactors. The analyses presented in this paper focuses on models assessment with attention on the influence of B4C oxidation on the release and transport of fission products. Such phenomenon is a concern point in the nuclear industry, as was highlighted during the Fukushima Daiichi accident. Simulation of the source term is a key point to evaluate the severe accident hazard along with other safety aspects.

Causes and Radiological Consequences of the Chernobyl and Fukushima Nuclear Accidents

2018 ◽  
Vol 4 (2) ◽  
Author(s):  
L. Sihver ◽  
N. Yasuda
Keyword(s):  
Local Time ◽  
Emergency Response ◽  
Nuclear Power ◽  
Power Plants ◽  
Severe Accident ◽  
Nuclear Accidents ◽  
Emergency Response System ◽  
Response System ◽  
Accident Management ◽  
Fukushima Daiichi

In this paper, the causes and the radiological consequences of the explosion of the Chernobyl reactor occurred at 1:23 a.m. (local time) on Apr. 26, 1986, and of the Fukushima Daiichi nuclear disaster following the huge Tsunami caused by the Great East Japan earthquake at 2.46 p.m. (local time) on Mar. 11, 2011 are discussed. The need for better severe accident management (SAM), and severe accident management guidelines (SAMGs), are essential in order to increase the safety of the existing and future operating nuclear power plants (NPPs). In addition to that, stress tests should, on a regular basis, be performed to assess whether the NPPs can withstand the effects of natural disasters and man-made failures and actions. The differences in safety preparations at the Chernobyl and Fukushima Daiichi will therefore be presented, as well as recommendations concerning improvements of safety culture, decontamination, and disaster planning. The need for a high-level national emergency response system in case of nuclear accidents will be discussed. The emergency response system should include fast alarms, communication between nuclear power plants, nuclear power authorities and the public people, as well as well-prepared and well-established evacuation plans and evacuation zones. The experiences of disaster planning and the development of a new improved emergency response system in Japan will also be presented together with the training and education program, which have been established to ensure that professional rescue workers, including medical staff, fire fighters, and police, as well as the normal populations including patients, have sufficient knowledge about ionizing radiation and are informed about the meaning of radiation risks and safety.

Preliminary Study of Influence of Severe Accident Mitigation Countermeasures to In-Vessel Release of Fisson Products

2010 ◽  
Author(s):  
Gaofeng Huang ◽  
Xuewu Cao ◽  
Jingxi Li
Keyword(s):  
Nuclear Power ◽  
Fission Products ◽  
Early Time ◽  
Severe Accident ◽  
Base Case ◽  
Primary Loop ◽  
Radiological Consequence ◽  
Secondary Loop ◽  
Station Blackout ◽  
Preliminary Study

During the severe accident in a nuclear power plant, large amounts of fission products release with accident progression, which includes in-vessel release and ex-vessel release. Mitigation of release of fission products is the need of alleviating radiological consequence in severe accident. Mitigation countermeasures to in-vessel release of fission products are studied, including feed-bleed in primary loop, feed-bleed in secondary loop and cooling of ex-vessel. Representative high pressure melt accident of station blackout is chosen, and different entry condition of countermeasures is assumed. The results show that: (1) Feed-bleed in primary loop is an effective countermeasure to mitigate in-vessel release of fission products. With early time to implement the countermeasure, in-vessel release fraction of fission products is low. (2) Feed-bleed in secondary loop is also an effective countermeasure to mitigate in-vessel release of fission products. Low in-vessel release fraction of fission products is produced with early time of countermeasure implemented. (3) Cooling of ex-vessel is not an effective countermeasure to control in-vessel release of fission products, the in-vessel release fraction in this case is almost equal to base case that uses none countermeasure.

Removal Efficiency of Iodine at Saturated Steam in Submerged Venturi Scrubber

2014 ◽  
Author(s):  
Majid Ali ◽  
Changqi Yan ◽  
Haifeng Gu ◽  
Khurram Mehboob ◽  
Athar Rasool
Keyword(s):  
Removal Efficiency ◽  
Flow Rate ◽  
Nuclear Power ◽  
Gas Flow ◽  
Radioactive Iodine ◽  
Saturated Steam ◽  
Venturi Scrubber ◽  
Compressed Gas ◽  
Water Saturated ◽  
Maximum Removal

Sever accident due to molten core of Nuclear Power Plant causes the production of steam which carries the radioactive iodine. It is important to retain the radioactive iodine from contaminated gas and steam before it is released into the environment. The purpose of this research is to investigate the removal efficiency of iodine in a submerged venturi scrubber for saturated steam at 100°C. Venturi Scrubber is submerged in a venturi tank filled with liquid which is alkaline by adding sodium hydroxide (NaOH) and sodium thiosulphate (Na2S2O3) in scrubbing water. Saturated Steam of 100 °C is injected into an experimental loop. Iodine removal efficiency is investigated for saturated steam at various compressed gas flow rate 330, 420, and 510 kg/s. Inlet and outlet concentration are measured at the sampling points of an experimental loop to calculate the iodine removal efficiency. The maximum removal efficiency of 99.4% is achieved at gas mass flow rate of 510 kg/s.

Severe Accident Analysis With Spatial Discretized Model by MAAP: Part 1 — Parametric Study on Fukushima-Daiichi Unit-2

2018 ◽  
Author(s):  
Kenichi Kanda ◽  
Yoshihisa Nishi ◽  
Kazuma Abe ◽  
Satoshi Nishimura ◽  
Koichi Nakamura ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Water Injection ◽  
Fission Products ◽  
Severe Accident ◽  
Accident Analysis ◽  
Detailed Model ◽  
Parametric Studies ◽  
Fukushima Daiichi ◽  
Alternative Water ◽  
Small Parts

Accident analyses of the Fukushima-Daiichi unit-2 nuclear power plant were performed with MAAP (Modular Accident Analysis Program) version 5.03. We assumed RCIC, SRV operation and alternative water injection in order to reproduce the measured pressure and temperature values in RPV and PCV. From parametric studies, it was found that the analysis results were in good agreement with the measured data. In this paper, the results of the parametric studies are reported. Furthermore, spatial discretization of compartments (such as rooms in the reactor building, etc.) into small parts successfully demonstrated the transient distribution and deposition of fission products (FPs) across the rooms. Such special discretization is particularly important for the forensic investigation of severe accidents and the deposited amount in the R/B might be estimated by using this detailed model.

Severe accident simulation for VVER-1000 reactor using ASTEC-V2.1.1.3

Kerntechnik ◽  
2021 ◽  
Vol 86 (6) ◽  
pp. 454-469
Author(s):  
S. H. Abdel-Latif
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Fission Products ◽  
Reactor Pressure Vessel ◽  
Severe Accident ◽  
Potential Core ◽  
Safety Requirements ◽  
The Core ◽  
Active Safety Systems ◽  
Reactor Pressure

Abstract The station black-out (SBO) is one of the main accident sequences to be considered in the field of severe accident research. To evaluate a nuclear power plant’s behavior in the context of this accident, the integral ASTEC-V2.1.1.3 code “Accident Source Term Evaluation Code” covers sequences of SBO accidents that may lead to a severe accident. The aim of this work is to discuss the modelling principles for the core melting and in-vessel melt relocation phenomena of the VVER-1000 reactor. The scenario of SBO is simulated by ASTEC code using its basic modules. Then, the simulation is performed again by the same code after adding and activating the modules; ISODOP, DOSE, CORIUM, and RCSMESH to simulate the ex-vessel melt. The results of the two simulations are compared. As a result of SBO, the active safety systems are not available and have not been able to perform their safety functions that maintain the safety requirements to ensure a secure operation of the nuclear power plant. As a result, the safety requirements will be violated causing the core to heat-up. Moreover potential core degradation will occur. The present study focuses on the reactor pressure vessel failure and relocation of corium into the containment. It also discusses the transfer of Fission Products (FPs) from the reactor to the containment, the time for core heat-up, hydrogen production and the amount of corium at the lower plenum reactor pressure vessel is determined.

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