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.

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.

Improvement of SAMPSON Code for BWR Severe Accident Analysis

2013 ◽  
Author(s):  
Maolong Liu ◽  
Yuki Ishiwatari ◽  
Koji Okamoto
Keyword(s):  
Stainless Steel ◽  
Oxidation Reaction ◽  
Reactor Core ◽  
Severe Accident ◽  
Accident Analysis ◽  
Light Water Reactors ◽  
Fuel Cladding ◽  
Fuel Rod ◽  
Water Reactors ◽  
The Impact

The SAMPSON code has been developed in the IMPACT project in Japan to investigate severe accident phenomena for light water reactors. It integrates various analysis modules into a single code. The authors improved the fuel rod heat-up module of SAMPSON code by modeling the oxidation reaction of various core structures, including Zircaloy, stainless steel and B4C. And the creep failures of the Zircaloy fuel cladding and stainless steel monitoring guide tubes of the source range monitor (SRM) in the reactor core was also modeled for severe accident analysis.

Application of Thermal Hydraulic and Severe Accident Code SOCRAT/V3 to Bottom Water Reflood Experiment QUENCH-LOCA-1

2016 ◽  
Vol 2 (2) ◽  
Author(s):  
Alexander Vasiliev ◽  
Juri Stuckert
Keyword(s):  
Nuclear Power ◽  
Hydrogen Generation ◽  
Cooling System ◽  
Reactor Core ◽  
Severe Accident ◽  
Loss Of Coolant Accident ◽  
Post Test ◽  
Core Cooling ◽  
Modeling Code ◽  
Code Quality

This study aims to (1) use the thermal hydraulic and severe fuel damage (SFD) best-estimate computer modeling code SOCRAT/V3 for post-test calculation of QUENCH-LOCA-1 experiment and (2) estimate the SOCRAT code quality of modeling. The new QUENCH-LOCA bundle tests with different cladding materials will simulate a representative scenario for a loss-of-coolant-accident (LOCA) nuclear power plant (NPP) accident sequence in which the overheated (up to 1050°C) reactor core would be reflooded from the bottom by the emergency core cooling system (ECCS). The test QUENCH-LOCA-1 was successfully performed at the KIT, Karlsruhe, Germany, on February 2, 2012, and was the first test for this series after the commissioning test QUENCH-LOCA-0 conducted earlier. The SOCRAT/V3-calculated results describing thermal hydraulic, hydrogen generation, and thermomechanical behavior including rods ballooning and burst are in reasonable agreement with the experimental data. The results demonstrate the SOCRAT code’s ability for realistic calculation of complicated LOCA scenarios.

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.

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 Nuclear policy-nuclear fuels for peaceful, safer co-generation and environmentally benign applications

2020 ◽  
Vol 9 (3) ◽  
pp. 724
Author(s):  
Syazwani Mohd Fadzil ◽  
Shafi Qureshi ◽  
Sekhar Basu ◽  
K. Kasturirangan ◽  
Anil Kakodkar ◽  
...  
Keyword(s):  
High Temperature ◽  
Nuclear Power ◽  
Reactor Core ◽  
Negative Temperature ◽  
Hydrogen Gas ◽  
Hydraulic Diameter ◽  
Nuclear Fuels ◽  
Spent Fuel ◽  
Fuel Temperature ◽  
Nuclear Policy

Here, safer nuclear fuels which can sustain in the high temperature and fluence environment of the reactor core are investigated to utilize nuclear energy peacefully. At Nuclear Fuel Complex in Hyderabad, nuclear fuels are being manufactured which are best suited for high temperature and fluence environment of the reactor core even in accidental scenarios. In this paper, nuclear fuels manufactured at NFC, Hyderabad are presented. The developed nuclear fuels have higher equivalent hydraulic diameter and breeding capability to produce U^233. Nuclear fuels having higher equivalent hydraulic diameter reduce the reactor core temperature substantially. These fuels have negative temperature coefficient of reactivity. Thus, in case of an accident, the fuel temperature never exceeds the safety limit. Therefore, the thermal heat available across the secondary of a heat exchanger can be utilized for different industrial processes. This allows the development of key technologies, such as safer co-generation of electricity and Hydrogen. The Three-Stage Indian Nuclear Power Program has been explained for nuclear disarmament. The product Hydrogen gas has been utilized in many ways for different applications. Moreover, the processing of iron ore with the energy obtained from the IHX secondary side, eliminates the burning of coals and CO2 emissions into the environment. Several radioisotopes have been developed for medical applications from spent fuel.  

Development of Transient and Safety Analysis Method for ATF With SiC

2016 ◽  
Author(s):  
Fumie Sebe ◽  
Masato Yamada ◽  
Yutaka Takeuchi ◽  
Kazuo Kakiuchi ◽  
Kazunari Okonogi
Keyword(s):  
High Temperature ◽  
Oxidation Reaction ◽  
Transient Analysis ◽  
Safety Analysis ◽  
Hydrogen Gas ◽  
Lessons Learned ◽  
Temperature Behavior ◽  
Severe Accident ◽  
Water Steam ◽  
Accident Conditions

Based on lessons learned from the Fukushima Daiichi nuclear power plant accident, pursuit of accident tolerant fuel (ATF) has been discussed by many institutions in the world. Toshiba identified a silicon carbide (SiC) ceramic as the most promising material for accident tolerant fuel. Since SiC has less active characteristics in the presence of high temperature water steam (H2O) and is expected to be tolerant of severe accident conditions. Moreover, SiC has a smaller neutron absorption cross-section which is advantageous feature in terms of neutron economy. Zirconium alloys (Zry) are one of the main structural materials in LWR core. In high temperature H2O environment under severe accident conditions, Zry rapidly reacts with H2O and oxidation reaction accompanied by release of hydrogen gas occurs. Since SiC may inhibit the progress of oxidation reaction compared to Zry metal alloys, hydrogen and heat generation is expected to decrease in the case of core uncovered accident conditions. In order to confirm the advantage of SiC over Zry as core materials, transient analysis and safety analysis are carried out. For transient analysis, analyses of temperature behavior of cladding at plant transient condition are carried out with best-estimate transient analysis code. This analysis confirmed the effect of physical properties differences between SiC and Zry on cladding temperature behavior. Moreover to indicate the effectiveness of SiC under the core uncovered condition with oxidation reaction, safety analysis with latest “MAAP” code is carried out and the whole plant behavior during severe accident sequence is simulated. This analysis showed the effectiveness of SiC to mitigate the oxidation reaction. As the result of these analyses, the advantage of SiC over Zry can be perceived. And also, future challenges of SiC application as ATF can be clarified through these analyses.

Thermal Resistance of Listeria spp. in Milk

1991 ◽  
Vol 54 (1) ◽  
pp. 12-14 ◽  
Author(s):  
JOE G. BRADSHAW ◽  
JAMES T. PEELER ◽  
ROBERT M. TWEDT
Keyword(s):  
High Temperature ◽  
Thermal Resistance ◽  
Heat Resistance ◽  
Raw Milk ◽  
The Other ◽  
Time Processing ◽  
Listeria Ivanovii ◽  
High Temperature Short Time ◽  
The Difference ◽  
Short Time

The thermal resistance of one strain each of Listeria ivanovii, L. seeligeri, and L. welshimeri and three L. monocytogenes strains was determined in raw and sterile milk. Listeria spp. suspended in milk at concentrations of 1 × 105 cells/ml were heated at temperatures ranging from 52.2 to 71.1°C for various contact times. The heat resistance of L. monocytogenes appeared somewhat greater than that of the other Listeria spp. in both milks, but the difference was not statistically significant (α = 0.05). High-temperature, short-time processing is adequate for pasteurization of raw milk.

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.

scholarly journals Note on Dr. Burdon Sanderson’s latest views of ferments and germs

1878 ◽  
Vol 26 (179-184) ◽  
pp. 353-356
Keyword(s):  
High Temperature ◽  
Bacillus Anthracis ◽  
Royal Society ◽  
The Other ◽  
Spontaneous Generation ◽  
Outer Line ◽  
Living Things ◽  
Characteristic Behaviour ◽  
The Difference ◽  
The One

While writing the paper which the Council of the Royal Society has recently done me the honour of accepting for the Philosophical Transactions, the abstract of a lecture delivered by Dr. Burdon Sanderson to the association of Medical Officers of Health was placed in my hands. The teem in which the author’s name is justly held will certainly give eight and currency to the views enunciated in this lecture. Speaking: ferments Dr. Sanderson says :—“ In defining the nature of fermentition we are in a dilemma, out of which there is no escape except by compromise. A. ferment is not an organism, because it has no structure. It is not a chemical body, because when it acts upon other bodies it maintains its own molecular integrity. On the whole, it resembles an organism such more than it resembles a chemical body, for its characteristic behaviour is such as, if it had a structure, would prove it to be living. Ten years ago the opponents of spontaneous generation were called Pansperusts, because it was supposed that in the so-called generation equivoca, in very case in which Bacteria appeared to spring out of nothing, the result as referable to the influence of unseen but actually existing germs. The assearches of the last few years have carried us beyond this stage. . . . the outer line of defence, represented by the aphoristic expression omne ivum ex ovo , has been for some time abandoned. The ground which the orthodox biologist holds now, as against the heterodox, is not that every bacterium must have been born of another Bacterium, but that every Bacterium must have been born of something which emanated from another bacterium, that something not being assumed to be endowed with structure in the morphological or anatomical sense, but only in the molecular chemical sense. It is admitted by all, even by Professor Tyndall, that, far as structure is concerned, the germinal or life-producing matter out which Bacteria originate exhibits no characters which, can be appreciated by the microscope; and other researches have proved that the Seminal matter is capable of resisting destructive influences, particularly those of high temperature, which are absolutely fatal to the Bacteria themselves. Germs have given place to things which are ultramicr scopical—to molecular aggregates—of which all we can say is, what we have already said about the ferments, that they occupy the border between living and non-living things.” As directed against “ germs ” the argument that the “ germinal matter is capable of resisting destructive influences which are fatal to the themselves, will, I think, be found on consideration to lack validity Nobody is better acquainted than Dr. Sanderson with the two forms under which the contagium of splenic fever appears. He knows that the one fugitive and readily destroyed, the other persistent and destroyed will difficulty. Now the recent researches of Koch, which have been verified by Cohn, prove conclusively that the difference here referred to is bast upon the fact that the fugitive contagium is the developed organism Bacillus anthracis, while the persistent contagium is the spore of tin organism. Dallinger’s excellent observations also establish a difference between the death-temperatures of monad germs and of adult monads while I need not do more than refer to the forthcoming Part of till Philosophical Transactions for illustrations of the extraordinary differences of the same nature which my recent researches have brought to light.

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