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.

Analysis of Accident Progression of Fukushima Daiichi NPP by the SAMPSON Code

2014 ◽  
Author(s):  
Masanori Naitoh ◽  
Marco Pellegrini ◽  
Hiroaki Suzuki ◽  
Hideo Mizouchi ◽  
Hidetoshi Okada
Keyword(s):  
Nuclear Power ◽  
Water Injection ◽  
Heat Removal ◽  
Fission Products ◽  
Severe Accident ◽  
Core Material ◽  
Decay Heat ◽  
The Core ◽  
Fukushima Daiichi ◽  
Fire Engine

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.

Study on Cooling Process in a Reactor Vessel of Sodium-Cooled Fast Reactor Under Severe Accident: Velocity Measurement Experiments Simulating Operation of Decay Heat Removal Systems

2020 ◽  
Author(s):  
Mitsuyo Tsuji ◽  
Kosuke Aizawa ◽  
Jun Kobayashi ◽  
Akikazu Kurihara ◽  
Yasuhiro Miyake
Keyword(s):  
Natural Circulation ◽  
Heat Removal ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Decay Heat ◽  
The Core ◽  
Severe Accidents ◽  
Piv Measurement ◽  
Image Velocimetry ◽  
Experimental Apparatus

Abstract In Sodium-cooled Fast Reactors (SFRs), it is important to optimize the design and operate decay heat removal systems for safety enhancement against severe accidents which could lead to core melting. It is necessary to remove the decay heat from the molten fuel which relocated in the reactor vessel after the severe accident. Thus, the water experiments using a 1/10 scale experimental apparatus (PHEASANT) simulating the reactor vessel of SFR were conducted to investigate the natural circulation phenomena in a reactor vessel. In this paper, the natural circulation flow field in the reactor vessel was measured by the Particle Image Velocimetry (PIV) method. The PIV measurement was carried out under the operation of the dipped-type direct heat exchanger (DHX) installed in the upper plenum when 20% of the core fuel fell to the lower plenum and accumulated on the core catcher. From the results of PIV measurement, it was quantitatively confirmed that the upward flow occurred at the center region of the lower and the upper plenums. In addition, the downward flows were confirmed near the reactor vessel wall in the upper plenum and through outermost layer of the simulated core in the lower plenum. Moreover, the relationship between the temperature field and the velocity field was investigated in order to understand the natural circulation phenomenon in the reactor vessel. From the above results, it was confirmed that the natural circulation cooling path was established under the dipped-type DHX operation.

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.

scholarly journals Standalone Containment Analysis of Four Phébus Tests with the ASTEC and the MELCOR Codes

2019 ◽  
Vol 2019 ◽  
pp. 1-19
Author(s):  
Bruno Gonfiotti ◽  
Sandro Paci
Keyword(s):  
Nuclear Power ◽  
Fission Products ◽  
Severe Accident ◽  
Primary Circuit ◽  
Solid Base ◽  
Pressurized Water ◽  
Control Rod ◽  
Experimental Campaign ◽  
Fuel Bundle ◽  
Software Codes

After the severe accident (SA) occurred at the Three-Miles Island Nuclear Power Plant (NPP), important efforts on the investigation of the different phenomena during this kind of accidents have been started. Several experimental campaigns investigating one phenomenon at time or the combination of two or more phenomena have been performed. Today, the Phébus experimental campaign is probably the most important activity on the evaluation of the coupling among different phenomena. Four out of five tests investigated the degradation of an intact Pressurized Water Reactor (PWR) fuel bundle and the subsequent transport of Fission Products (FP) and Structural Materials (SM) through the primary circuit and into the containment, while the fifth test was only the degradation of a bed of PWR fuel bundle debris. These tests were performed between 1990 and 2010 at the CEA Cadarache laboratories (France) in a 5000:1 scaled facility. The main four tests varied the employed control rod materials, the fuel burn-up, and the oxidizing conditions of the atmosphere (strongly or weakly). The outcomes of this experimental campaign created a solid base for the understanding of the involved phenomena and allowed the development of models and software codes capable of simulating the evolution of a SA in a real NPP. ASTEC and MELCOR were two of the main SA codes profiting from the results of this Phébus campaign. These two codes were further improved in the latest years to account for the findings obtained in more recent experimental campaigns. A continuous verification and validation work is then necessary to check how the newer code’s versions reproduce the tests performed in these older experimental campaigns such as Phébus one. The present work is intended to be the final step of a series of publications covering the activities carried out at University of Pisa with the ASTEC and the MELCOR SA codes on the four Phébus tests employing an intact PWR fuel bundle. Because of the complexity and the extent of these tests, only the containment aspects were considered in the precedent works, i.e., only the thermal-hydraulics transient and its coupling with the FP and SM behavior. Then, general conclusions based on the outcomes of these precedent works are summarized in this work.

Review on Water Radiolysis in the Fukushima Daiichi Accident: Potential Cause of Hydrogen Generation and Explosion

2014 ◽  
Author(s):  
Genn Saji
Keyword(s):  
Hydrogen Generation ◽  
Hydrogen Gas ◽  
Severe Accident ◽  
Normal Operation ◽  
Water Radiolysis ◽  
Chain Reactions ◽  
Decay Heat ◽  
Severe Accidents ◽  
Heat Curve ◽  
Fukushima Daiichi

Although the water radiolysis, decomposition of water by radiation, is a well-known phenomenon the exact mechanism is not well characterized especially for severe accidents. The author first reviewed the water radiolysis phenomena in LWRs during normal operation to severe accidents (e.g., TMI- and Chernobyl accidents) and performed a scoping estimation of the amount of radiological hydrogen generation, accumulation and release for the Fukushima Daiichi accident. The estimation incorporates the decay heat curve after a reactor trip combined with G-values. As much as 450 cubic meters-STP of accumulated hydrogen gas is estimated to be located inside the PCV just prior to the hydrogen explosion which occurred a day after the reactor trip in Unit 1. When a set of radiological chain reactions are incorporated the resultant reverse reactions substantially reduce the hydrogen generation, even when removal of molecular products (i.e., oxygen and hydrogen) is assumed stripped rapidly from boiling water through bubbles. Even in the most favorable configuration a typical amount of hydrogen gas reduces to only several tens of cubic meters. Finally, the author tested a new mechanism, “radiation-induced electrolysis,” which had been applied to his corrosion studies for last several years. His theory has been verified with the published in-pile test data, although he has never tried to apply this to his severe accident study. The predicted results indicated that the total inventory of hydrogen gas inside RPV may reach as much as 1000 cubic meters in just 3 hours during the SBO due to a high decay heat soon after the reactor trip through this process.

scholarly journals FISSION-PRODUCT RELEASES FROM A PHWR TERMINAL DEBRIS BED

2016 ◽  
Vol 5 (1) ◽  
pp. 95-105 ◽  
Author(s):  
M.J. Brown ◽  
D.G. Bailey
Keyword(s):  
Fission Product ◽  
Heavy Water ◽  
Fission Products ◽  
Severe Accident ◽  
Water Reactor ◽  
Decay Heat ◽  
The Core ◽  
Heavy Water Reactor ◽  
Release Model ◽  
Tank Water

During an unmitigated severe accident in a pressurized heavy water reactor (PHWR) with horizontal fuel channels, the core may disassemble and relocate to the bottom of the calandria vessel. The resulting heterogeneous in-vessel terminal debris bed (TDB) would likely be quenched by any remaining moderator, and some of the decay heat would be conducted through the calandria vessel shell to the surrounding reactor vault or shield tank water. As the moderator boiled off, the solid debris bed would transform into a more homogeneous molten corium pool located between top and bottom crusts. Until recently, the severe accident code MAAP-CANDU assumed that unreleased volatile and semi-volatile fission products remained in the TDB until after calandria vessel failure, due to low diffusivity through the top crust and the lack of gases or steam to flush released fission products from the debris. However, national and international experimental results indicate this assumption is unlikely; instead, high- and medium-volatility fission products would be released from a molten debris pool, and their volatility and transport should be taken into account in TDB modelling. The resulting change in the distribution of fission products within the reactor and containment, and the associated decay heat, can have significant effects upon the progression of the accident and fission-product releases to the environment. This article describes a postulated PHWR severe accident progression to generate a TDB and the effects of fission-product releases from the terminal debris, using the simple release model in the MAAP-CANDU severe accident code. It also provides insights from various experimental programs related to fission-product releases from core debris, and their applicability to the MAAP-CANDU TDB model.

The Development of a Mechanistic Model of MCCI Under Severe Accidents

2013 ◽  
Author(s):  
Wei Wei ◽  
Kelin Qi ◽  
Fuchang Shan ◽  
Yanfang Chen ◽  
Fude Guo
Keyword(s):  
Nuclear Power ◽  
Large Scale ◽  
Mechanistic Model ◽  
Severe Accident ◽  
Severe Accidents ◽  
Large Scale Analysis ◽  
Loss Of Coolant ◽  
Calculation Results ◽  
Station Blackout ◽  
Core Concrete

This paper describes a mechanistic model of the molten core-concrete interaction (MCCI) process under severe accidents, and selects the Daya-Bay nuclear power plant as the research object to calculate and analyze the process of the MCCI when the station blackout (SBO), or loss of coolant (LOCA) severe accident serial is happened. The calculation results of this procedure are compared with the large-scale analysis programs MELCOR to verify the reasonableness and correctness of the model. The results indicate that the model present in this paper can simulate the MCCI process correctly and reasonably under multi-serial severe accidents.

Evaluation of Iodine Source Term in PWR under Loss of Coolant Accident

2012 ◽  
Vol 482-484 ◽  
pp. 1115-1119 ◽  
Author(s):  
Khurram Mehboob ◽  
Xin Rong Cao
Keyword(s):  
Nuclear Power ◽  
Source Term ◽  
Fission Products ◽  
Retention Factor ◽  
Severe Accident ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Term Evaluation ◽  
Computer Based ◽  
Instantaneous Release

During the severe accident in nuclear power plant (NPP), large amounts of fission products are released with accident progression, including In-vessel and Ex-vessel release. Thus, the Source term evaluation is essential for the probability risk assessment (PRA) and is still imperative for the licensing and operation of NPPs. Iodine is one of the most reactive fission products emitting in a large amount to containment and have a severe impact on health and sounding environment. Therefore, the iodine source term has been evaluated for 1000MW Reactor, by considering the TMI-2 as the reference reactor. The modeling and simulation of released radioactivity have been carried out by developing a MATLAB computer-based program. For post 1100 operation days, with the instantaneous release of radioactivity to the containment has been studied under LOCA. The dependency of radioiodine on ventilation exhaust rates has been studied in normal, emergency and isolation mode of containment. Moreover, the containment retention factor is also evaluated in said states of containment.

Uncertainty studies on hydrogen source term with MAAP5 code

Kerntechnik ◽  
2021 ◽  
Vol 86 (2) ◽  
pp. 152-163
Author(s):  
T.-C. Wang ◽  
M. Lee
Keyword(s):  
Hydrogen Production ◽  
Nuclear Power ◽  
Hydrogen Generation ◽  
Pearson Correlation ◽  
Latin Hypercube Sampling ◽  
Severe Accident ◽  
Model Parameters ◽  
Power Company ◽  
Sensitivity Studies ◽  
The Impact

Abstract In the present study, a methodology is developed to quantify the uncertainties of special model parameters of the integral severe accident analysis code MAAP5. Here, the in-vessel hydrogen production during a core melt accident for Lungmen Nuclear Power Station of Taiwan Power Company, an advanced boiling water reactor, is analyzed. Sensitivity studies are performed to identify those parameters with an impact on the output parameter. For this, multiple calculations of MAAP5 are performed with input combinations generated from Latin Hypercube Sampling (LHS). The results are analyzed to determine the 95th percentile with 95% confidence level value of the amount of in-vessel hydrogen production. The calculations show that the default model options for IOXIDE and FGBYPA are recommended. The Pearson Correlation Coefficient (PCC) was used to determine the impact of model parameters on the target output parameters and showed that the three parameters TCLMAX, FCO, FOXBJ are highly influencing the in-vessel hydrogen generation. Suggestions of values of these three parameters are given.

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