scholarly journals Experimental Study on a Hydrogen Stratification Induced by PARs Installed in a Containment

Energies ◽  
2020 ◽  
Vol 13 (21) ◽  
pp. 5552
Author(s):  
Jongtae Kim ◽  
Seongho Hong ◽  
Ki Han Park ◽  
Jin Heok Kim ◽  
Jeong Yun Oh
Keyword(s):  
Experimental Study ◽  
High Temperature ◽  
Power Supply ◽  
Nuclear Power ◽  
Severe Accident ◽  
Experimental Simulation ◽  
Exhaust Gas ◽  
Metal Oxidation ◽  
Fuel Cladding ◽  
Analysis Model

Hydrogen can be produced in undesired ways such as a high temperature metal oxidation during an accident. In this case, the hydrogen must be carefully managed. A hydrogen mitigation system (HMS) should be installed to protect a containment of a nuclear power plant (NPP) from hazards of hydrogen produced by an oxidation of the fuel cladding during a severe accident in an NPP. Among hydrogen removal devices, passive auto-catalytic recombiners (PARs) are currently applied to many NPPs because of passive characteristics, such as not requiring a power supply nor an operators’ manipulations. However, they offer several disadvantages, resulting in issues related to hydrogen control by PARs. One of the issues is a hydrogen stratification in which hydrogen is not well-mixed in a compartment due to the high temperature exhaust gas of PARs and accumulation in the lower part. Therefore, experimental simulation on hydrogen stratification phenomenon by PARs is required. When the hydrogen stratification by PARs is observed in the experiment, the verification and improvement of a PAR analysis model using the experimental results can be performed, and the hydrogen removal characteristics by PARs installed in an NPP can be evaluated using the improved PAR model.

Severe Accident Analysis for Reactor Core Applying SiC to Fuel Claddings and Channel Boxes

2018 ◽  
Author(s):  
Hideki Horie ◽  
Yutaka Takeuchi ◽  
Kenya Takiwaki ◽  
Fumie Sebe ◽  
Kazuo Kakiuchi ◽  
...  
Keyword(s):  
High Temperature ◽  
Oxidation Reaction ◽  
Nuclear Power ◽  
Reaction Rate ◽  
Hydrogen Generation ◽  
Reactor Core ◽  
The Other ◽  
Severe Accident ◽  
Fuel Cladding ◽  
The Difference

Development of a fuel cladding or a channel box applying silicon carbide (SiC), which has high accident tolerance, in place of zircaloy (Zry) or Steel Use Stainless (SUS) composing current light water reactors, has being proceeded with after the accident of Fukushima Daiichi Nuclear Power Plant (1F). When applying SiC to core structures of a nuclear power plant such as fuel cladding, it is expected that the difference of high temperature oxidation characteristics in the severe accident (SA) conditions would mitigate progression of core damage comparing with the current Zry fuel core. This study performed SA analyses considering high temperature chemical reaction characteristics of SiC by using SA analysis code “MAAP”, and thermal hydraulics analysis code “TRAC Toshiba version (TRAC)”, and compared the difference between SiC and Zry. Both codes originally have no model of oxidation reaction for SiC. Hence, a new model for SiC in addition to the current model for Zry was incorporated into “MAAP”. On the other hand, “TRAC” adjusted reaction rate by changing oxidation reaction coefficients in the current Zry oxidation reaction models such as Baker-Just and Cathcart correlations in order to simulate SiC-water/steam reaction. In analysis using “MAAP”, seven accident sequences from representative Probabilistic Risk Assessment ones were selected to evaluate the difference of SA behavior between two materials. As a result, in the case of replacing current Zry of fuel claddings and channel boxes into SiC, an amount of hydrogen generation reduced to about 1/6 than the case of Zry. In addition to that, in the case of replacing SUS structures in the reactor core into SiC, an amount of hydrogen generation moreover reduced to about 1/6 than the above result, which means just about 2% of an amount in the original case. On the other hand, in analysis using “TRAC”, the accident sequence for unit 3 of 1F (1F3) was selected, and reaction rate in the oxidation reaction model was examined as parameter. In the case of 1.0 time of the reaction rate, which means an original reaction rate, maximum fuel cladding temperature exceeded 2000K in 50 hour after reactor scram. However, using the reaction rate below 0.01 to the original one, the fuel cladding temperature didn’t exceed 1,600K.

Identification of Failure Modes Under Design Extension Conditions

2015 ◽  
Author(s):  
Naoto Kasahara ◽  
Izumi Nakamura ◽  
Hideo Machida ◽  
Hitoshi Nakamura ◽  
Koji Okamoto
Keyword(s):  
High Temperature ◽  
Nuclear Power ◽  
Failure Modes ◽  
Low Cycle Fatigue ◽  
External Pressure ◽  
Lessons Learned ◽  
Local Failure ◽  
Severe Accident ◽  
Mitigation Measures ◽  
Loading Mode

As the important lessons learned from the Fukushima-nuclear power plant accident, mitigation of failure consequences and prevention of catastrophic failure became essential against severe accident and excessive earthquake conditions. To improve mitigation measures and accident management, clarification of failure behaviors with locations is premise under design extension conditions such as severe accidents and earthquakes. Design extension conditions induce some different failure modes from design conditions. Furthermore, best estimation for these failure modes are required for preparing countermeasures and management. Therefore, this study focused on identification of failure modes under design extension conditions. To observe ultimate failure behaviors of structures under extreme loadings, new experimental techniques were adopted with simulation materials such as lead and lead-antimony alloy, which has very small yield stress. Postulated failure modes of main components under design extension conditions were investigated according three categories of loading modes. The first loading mode is high temperature and internal pressure. Under this mode, ductile fracture and local failure were investigated. At the structural discontinuities, local failure may become dominant. The second is high temperature and external pressure loading mode. Buckling and fracture were investigated. Buckling occurs however hardly break without additional loads or constraints. The last loading is excessive earthquake. Ratchet deformation, collapse, and fatigue were investigated. Among them, low-cycle fatigue is dominant.

scholarly journals Investigation on Aerosol Pool Scrubbing Model During Severe Accidents

2021 ◽  
Vol 9 ◽  
Author(s):  
L. W. He ◽  
Y. X. Li ◽  
Y. Zhou ◽  
S. Chen ◽  
L. L. Tong ◽  
...  
Keyword(s):  
Bubble Size ◽  
Nuclear Power ◽  
Predictive Ability ◽  
Decontamination Factor ◽  
Light Water Reactor ◽  
Severe Accident ◽  
Spent Fuel ◽  
Analysis Model ◽  
Light Water ◽  
Water Reactor

During a nuclear power plant severe accident, discharging gas mixture into the spent-fuel pool is an alternative containment depressurization measurement through which radioactive aerosols can be scrubbed. However, it is necessary to develop a code for analyzing the decontamination factor of aerosol pool scrubbing. This article has established the analysis model considering key aerosol pool scrubbing mechanisms and introduced the Akita bubble size relationship. In addition, a code for evaluating the decontamination factor of aerosol pool scrubbing was established. The Advanced Containment Experiment and Light Water Reactor Advanced Containment Experiment were simulated with the code considering different bubble sizes of the Akita model and MELCOR default value to verify the suitability of the Akita bubble size model for simulating aerosol pool scrubbing. Furthermore, the simulation results were compared with the results analyzed by MELCOR code and COCOSYS code from literature, and equivalent predictive ability was observed. In addition, a sensitivity analysis on bubble size was conducted, and the contribution of different behaviors and mechanisms has been discussed. Finally, the bubble breakup equation was revised and verified with the conditions of the multi-hole bubbler in the Advanced Containment Experiment and Light Water Reactor Advanced Containment Experiment.

The Experimental Study on High Temperature Air Combustion and CF4 Decomposition

2005 ◽  
Author(s):  
L. Jia ◽  
S. Ma
Keyword(s):  
Heat Transfer ◽  
Experimental Study ◽  
High Temperature ◽  
Air Temperature ◽  
Waste Heat ◽  
Exhaust Gas ◽  
High Temperature Air Combustion ◽  
Factors Influencing ◽  
Regenerative Burner ◽  
Ceramic Honeycomb

For the purpose of decomposing the processing gases CF4 from semiconductor manufacturers, ceramic honeycomb regenerative burner system is suggested by using the principle of HTAC. A simulated high temperature air combustion furnace has been used to determine the features of HTAC flames and the results of the decomposition of CF4. The preheat air temperature of it is above 900 °C. The exhaust gas released into the atmosphere is lower than 150 °C. Moreover, the efficiency of recovery of waste heat is higher than 70%, the NOx level in exhaust gas is less than 300 mg/m3 and the distribution of temperature in the furnace is nearly uniform. The factors influencing heat transfer, temperature profile in chamber and NOx emission were discussed. Also some CF4 can be decomposed in this system. Experimental results indicated that the destruction removal efficiency (DRE) of CF4 increases with the increasing of concentration of H2O in some scale, and will not keep climbing when the concentration reach a point. DRE of CF4 decreases with the decreasing of concentration of CF4 under condition of other factors unchanged.

scholarly journals Development of Instrument Transmitter Protecting Device against High-Temperature Condition during Severe Accidents

2014 ◽  
Vol 2014 ◽  
pp. 1-7 ◽  
Author(s):  
Min Yoo ◽  
Sung Min Shin ◽  
Hyun Gook Kang
Keyword(s):  
Heat Transfer ◽  
High Temperature ◽  
Nuclear Power ◽  
Power Plants ◽  
Severe Accident ◽  
Transfer Model ◽  
Potential Threat ◽  
Severe Accidents ◽  
Development Of Instrument ◽  
Novel Concept

Reliable information through instrumentation systems is essential in mitigating severe accidents such as the one that occurred at the Fukushima Daiichi nuclear power plant. There are five elements which might pose a potential threat to the reliability of parameter detection at nuclear power plants during a severe accident: high temperature, high pressure, high humidity, high radiation, and missiles generated during the evolution of a severe accident. Of these, high temperature apparently poses the most serious threat, since thin shielding can get rid of pressure, humidity, radiation (specifically, alpha and beta radiations), and missile effects. In view of this fact, our study focused on designing an instrument transmitter protecting device that can eliminate the high-temperature effect on transmitters to maintain their functional integrity. We present herein a novel concept for designing such a device in terms of heat transfer model that takes into account various heat transfer mechanisms associated with the device.

The Formulation of Material Characteristics of Austenitic Stainless Steels at Extremely High Temperature

2017 ◽  
Author(s):  
Kenta Shimomura ◽  
Takashi Onizawa ◽  
Shoichi Kato ◽  
Masanori Ando ◽  
Takashi Wakai
Keyword(s):  
High Temperature ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Structural Integrity ◽  
Austenitic Stainless Steels ◽  
Severe Accident ◽  
Integrity Assessment ◽  
Severe Accidents ◽  
Material Characteristics

This paper describes the formulation of material characteristics of austenitic stainless steels at extremely high temperature which meets in some kinds of severe accidents of nuclear power plants. After the severe accident in Fukushima dai-ichi nuclear power plants, it has been supposed to be very important not only to prevent the occurrence of abnormal conditions, i.e. from the first to the third layer safety, but also to prevent the expansion of the accident conditions, i.e. the fourth layer safety[1] [2]. In order to evaluate the structural integrity under the severe accident condition, material characteristics which can be used in the numerical analyses, such as finite element analysis, were required [3] [4]. However, there were no material characteristics applicable to the structural integrity assessment at extremely high temperature. Therefore, a series of tensile and creep tests was performed for austenitic stainless at extremely high temperature which meets in some kinds of severe accidents of nuclear power plants, namely up to 1000 °C. Based on the acquired data from the tests, monotonic stress-strain equation and creep rupture equation applicable to the structural analysis at extremely high temperature, up to 1000 °C were formulated. As a result, these formulae make it possible to conduct the structural integrity assessment using numerical analysis techniques, such as finite element method.

A review of exploratory studies at RSC Energia into high-temperature voltage conversion systems for an electrically-propelled transportation spacecraft

2021 ◽  
pp. 46-68
Author(s):  
Viktor V. SINYAVSKIY ◽  
Stanislav R. TROITSKIY
Keyword(s):  
High Temperature ◽  
Power Supply ◽  
Nuclear Power ◽  
Low Voltage ◽  
Experimental Studies ◽  
Voltage Converter ◽  
Experimental Proof ◽  
Power Sources ◽  
Conversion Frequency ◽  
Voltage Conversion

The paper presents studies conducted at RSC Energia that looked into the feasibility of developing a high-temperature high-power voltage converter of a low-voltage space power supply system for an electrically-propelled space transportation spacecraft based on classical voltage conversion systems using electroplasma gates (key elements) and high-temperature transformers. It provides electrical schematics fro voltage conversion systems. It presents results of experimental studies on lab prototypes of electroplasma gates with electrical power of tens and hundreds kilowatts, which have demonstrated that in principle it is possible to use them as a basis for development of high-temperature radiation-proof voltage converters with mass fraction of ~0.01 kg/A for such power sources as nuclear power-supply systems with sub-megawatt and megawatt thermionic converter reactor. It provides a rationale for selecting the voltage conversion frequency in systems without forced cooling. It provides an experimental proof that in principle it is possible to implement high-temperature transformers with 600°С operating temperature level, and gives recommendations on selection of materials. Key words: space power system, electrically-propelled transportation spacecraft, voltage converter, electroplasma gate, high-temperature transformer, conversion frequency.

Study on Flow Structure in a Supersonic Steam Injector

2018 ◽  
Author(s):  
Yuki Kamata ◽  
Masaya Fujishiro ◽  
Akiko Kaneko ◽  
Yutaka Abe
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Flow Structure ◽  
Nuclear Power ◽  
Severe Accident ◽  
Analysis Model ◽  
Operating Characteristics ◽  
Two Phase ◽  
High Heat Transfer ◽  
Discharge Pressure

Steam injector (SI) are attracting attention as countermeasures against severe-accident in nuclear reactors. It is a static jet pump which operates using driving force to draw steam and water by internal pressure being reduced by direct contact condensation of these two fluids. In addition, capability of SI as a heat exchanger with high heat-transfer is expected. The absence of a drive unit such as an external power supply and rotating machine is significant characteristic of SI, and it can be expected to suppress the cost of installation and maintenance. It is also possible to produce a discharge pressure higher than the inlet pressure. From these facts, SI is expected to be applied as a static safety system that can cool the reactor core even if power lose at the nuclear power plant. Although SI has been used for steam engines since long ago, the mechanism of its operation has not yet been clarified. Thus, elucidation of the mechanism of operation of SI is indispensable for introduction to a nuclear power plant. A one-dimensional analytical model which predicts the operating characteristics assuming full condensation and evaluated discharge pressure is constructed (Narabayashi et al., 1996). In addition, from detailed observation, it was confirmed by that there is a boundary of luminance in the diffuser section (Abe et al., 2012). This is considered as the boundary where the two-phase flow condenses. However, this phenomenon is not considered in the current analysis model. The aim of this research is to clarify the flow structure in the diffuser section of SI. For that purpose, the state of the diffuser section of the transparent SI test part was observed with a highspeed camera, and the pressure at each point in it was measured simultaneously. The boundary of the luminance is confirmed to approach the throat as closing the back-pressure valve. In addition to this boundary, it was confirmed that the bright region intermittently propagated downstream. This phenomenon is supposed to be caused by pressure increasing, and the propagations assumed as a pressure wave moving at the sound speed. Thus, void fraction is estimated by calculating this propagation speed with image processing. Furthermore, experiments were carried out using three types of large, medium and small test parts, respectively. From the above results, the internal flow structure in the SI diffuser section was discussed.

Severe Accident Condition Modeling in PWR Environment: Creep Rupture Modeling

2013 ◽  
Author(s):  
Frederick W. Brust ◽  
R. Iyengar ◽  
M. Benson ◽  
Howard Rathbun
Keyword(s):  
High Temperatures ◽  
Nuclear Power ◽  
Reactor Core ◽  
Creep Rupture ◽  
Severe Accident ◽  
Time To Failure ◽  
Analysis Model ◽  
Pressurized Water ◽  
Condition Modeling ◽  
Core Cooling

A problem of interest in the nuclear power industry involves the response of pressurized water reactor (PWR) pressure boundary components under long-term station blackout (SBO) conditions. SBO is a particularly challenging event to nuclear safety, since all alternating current power required for core cooling is lost. If unmitigated, such a scenario will eventually lead to the reactor core being uncovered. Thermal-hydraulic (T-H), computational fluid dynamics, and structural combined creep/plasticity analyses of this scenario have been conducted and are presented here. In this severe accident scenario, high temperatures can occur, and impart this thermal energy to the surrounding structures, including the reactor vessel, nozzles, reactor coolant system (RCS) hot leg piping and S/G tubes. At such high temperatures and pressures, creep rupture of RCS piping and/or steam generator (S/G) tubes becomes possible. The intent of this paper is to present a finite element based analysis model that can be used to evaluate the time to failure of the nozzle-weld-pipe configuration.

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