Development of Technologies on Innovative-Simplified Nuclear Power Plant Using High-Efficiency Steam Injector: Part 14—Cooling Characteristics of the Particulate Core Material Debris Accumulated in Horizontal Narrow Gap

2006 ◽  
Author(s):  
Y. Hirao ◽  
G. Su ◽  
K. Sugiyama ◽  
T. Narabayashi ◽  
M. Mori ◽  
...  
Keyword(s):  
Heat Supply ◽  
Nuclear Power ◽  
High Efficiency ◽  
Hot Spot ◽  
Research Work ◽  
Lower Surface ◽  
Injection System ◽  
Core Material ◽  
Metal Debris ◽  
The Core

When LOCA occurs in proposed nuclear reactor systems, the coolability of the core would be kept by the SI core injection system and therefore the probability of the core meltdown is negligible small. In this research work, we make it clear that the coolability of the RPV bottom is secured even if a part of the core should melt and a substantial amount of debris should be deposited on the lower plenum. In this report, we examined experimentally the coolability of the RPV bottom that a Zircaloy-based loose debris layer is deposited on. We set up a heat supply section made by SUS304 on the loose debris layer and measured the heat flux released into the loose debris bed and the temperature at the lower surface of the heat supply section. In addition, we measured the temperature distribution at the bottom of the loose debris bed. It became clear in this study that the coolability depends on the amount of coolant supplied, and the hot spot would not occur when coolant is supplied. Even if a hotspot should occur in the oxidization of loose metal debris accompanied with rapid heat generation. It is found that when a small amount of coolant can be supplied, it disappears because of a high capillary force of oxidized loose debris. So it is confirmed that the soundness of RPV is basically maintained.

Experiments and Analytical Simulation on Steam Injector Driven Passive Core Injection System for Innovative-Simplified Nuclear Power Plant

2008 ◽  
Author(s):  
Shuichi Ohmori ◽  
Tadashi Narabayashi ◽  
Michitsugu Mori ◽  
Fumitoshi Watanabe
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
High Pressures ◽  
High Efficiency ◽  
Cooling System ◽  
Steam Pressure ◽  
Injection System ◽  
The Core ◽  
Core Cooling

A steam injector (SI) is a simple, compact and passive pump and also acts as a high-performance direct-contact compact heater. We are developing an innovative idea by applying SI system for core injection system in emergency core cooling systems (ECCS) to further improve the safety of nuclear power plants. The passive core injection system (PCIS) driven by high-efficiency SI is a system that, in an accident such as a LOCA (loss of coolant accident), attains discharge pressure higher than the supply steam pressure to inject water into the reactor by operating the SI, by supplying water from a pool in a containment vessel and the steam from a reactor pressure vessel (RPV). The SI, passive equipment, is used to replace large rotating machines such as pumps and motors, eliminating the failure probabilities of such active equipment. When the water and steam supply valves open, the SI-driven PCIS (SI-PCIS) will automatically start to inject water into the core to keep the core covered with water. The SI-PCIS works for the range of steam pressure conditions from atmosphere pressure through high pressures, in which the analytical simulations of SI were carried out based on the plenty amount of experimental data using reduced scale SI. We further simulated and evaluated the core cooling and water injection performance of SI-PCIS in BWR using RETRAN-3D code for the case of small LOCA. A reactor, such as ESBWR, equipped with the passive safety system by gravity-driven cooling system (GDCS) and the depressurization valves (DPVs) should be inevitable to lead to large LOCA even for the case of small LOCA by forcibly opening the DPVs to inject water from the GDCS pool due to that the GDCS water head is up to ∼0.2MPa. On the contrary, our simulation exhibited that SI-PCIS could save the reactors from leading to large LOCA by discharge of the water into a core for the cases of small LOCA or DPV unexpectedly open. In addition, we conducted the analytical simulations of SI, which grew in size for the actual nuclear power plant. A part of this report are fruits of research which is carried out by Tokyo Electric Power Company (TEPCO), Toshiba corporation, and seven universities in Japan, funded from the Ministry of Economy, Trade and Industry (METI) of Japan as the national public research-funded program.

Epstein frame investigation on soft magnetic properties of Fe-based amorphous strips

2021 ◽  
Vol 63 (10) ◽  
pp. 604-609
Author(s):  
Yanxing Xing ◽  
Shaoxiong Zhou ◽  
Wenzhi Chen ◽  
Bangshao Dong ◽  
Yaqiang Dong ◽  
...  
Keyword(s):  
High Efficiency ◽  
Core Loss ◽  
Core Material ◽  
Test Results ◽  
The Core ◽  
Distribution Transformers ◽  
Core Losses ◽  
Power Frequency ◽  
Field Annealing ◽  
Frame Method

Fe-based amorphous strip (AM strip) is a core material for high-efficiency distribution transformers and contributes to saving energy loss in electricity distribution. The core loss and apparent power for 2605SA1 amorphous strips at power frequency are studied using the Epstein frame method. Longitudinal magnetic field annealing and the influence of measuring modes on test results are investigated in detail. Improved test efficiency and higher accuracy in test results for amorphous ribbons are demonstrated and it is found that the number of strips and the lap joint methods affect the test results greatly. The waveform of the secondary induction voltage becomes sinusoidal with the increase of strip number. The values of core loss and apparent power become stable once the total number of strips is larger than 20. The coefficient of eddy current loss (e) also affects the correction of testing core losses. The test results could be improved at a smaller value of e when the waveform of the secondary induction voltage becomes deformed from sinusoidal due to a lower number of strips (below 20). The measured results were found to be reproducible when the strip number of each layer was one or two. However, the core loss and the apparent power increased along with the increase in the number of strips in each layer. Moreover, demagnetisation showed no effects on the test results when using the Epstein frame method.

scholarly journals Microencapsulation of Camelina sativa Oil Using Selected Soluble Fractions of Dietary Fiber as the Wall Material

Foods ◽  
2019 ◽  
Vol 8 (12) ◽  
pp. 681 ◽  
Author(s):  
Kanclerz ◽  
Drozińska ◽  
Kurek
Keyword(s):  
Dietary Fiber ◽  
High Efficiency ◽  
Gum Arabic ◽  
Camelina Sativa ◽  
Core Material ◽  
Wall Material ◽  
The Core ◽  
Wall Materials ◽  
Low Efficiency

The aim of the study was to prove the usefulness of microencapsulation of Camelina sativa oil regarding its vulnerability to oxidation caused by oxygen, temperature, and other factors. Pectin, inulin, gum arabic, and β-glucan, each of them mixed with maltodextrin, were used as wall materials and their appropriability to reduce oxidation of the core material was examined. Microcapsules were prepared by spray drying, which is the most commonly used and very effective method. The research confirmed results known from literature, that gum arabic and inulin are most proper wall materials, because they ensure small oxidation increase during storage (4.59 and 5.92 eq/kg after seven days respectively) and also provide high efficiency of process (83.93% and 91.74%, respectively). Pectin turned out to be the least appropriate polysaccharide because it is not able to assure sufficient protection for the core material, in this case Camelina sativa oil, due to low efficiency (61.36%) and high oxidation (16.11 eq/kg after seven days). β-glucan occurred to be the coating material with relatively high encapsulation efficiency (79.26%) but high humidity (4.97%) which could negatively influence the storage of microcapsules. The use of polysaccharides in microencapsulation, except performing the role of wall material, has the advantage of increasing the amount of dietary fiber in human diet.

scholarly journals Ferrites and Nanocrystalline Alloys Applied to DC–DC Converters for Renewable Energies

2022 ◽  
Vol 12 (2) ◽  
pp. 709
Author(s):  
Dante Ruiz ◽  
Jorge Ortíz ◽  
Edgar Moreno ◽  
Claudio Fuerte ◽  
Vicente Venegas ◽  
...  
Keyword(s):  
Power Density ◽  
High Performance ◽  
Low Cost ◽  
Research Work ◽  
Design Procedure ◽  
Renewable Energies ◽  
Nanocrystalline Alloys ◽  
Core Material ◽  
The Core ◽  
Technical Specifications

The medium frequency transformer (MFT) with nanocrystalline alloys is quintessential in new DC–DC converters involved in various front-end applications. The center piece to achieve high-performance, efficient MFTs is the core. There are various options of core materials; however, no deep information is available about which material characteristics and design procedure combo are best to get high performance MFTs while operating at maximal power density. To provide new insights about interrelation between the selection of the core material with the compliance technical specifications, differently to other proposals, this research work aims to design and build, with the same methodology, two MFT prototypes at 20 kHz, with nanocrystalline and ferrite cores, to highlight power density, and overall performance and cost, as matching design criteria. As the experimental results show, a nanocrystalline core has the highest power density (36.91 kW/L), designed at 0.8 T to obtain low losses at 20 kHz, achieving an efficiency of 99.7%. The power density in the ferrite MFT is 56.4% lower than in the nanocrystalline MFT. However, regarding construction cost, the ferrite MFT is 46% lower, providing this a trend towards low-cost DC–DC converters. Finally, high power density in MFTs increases the power density of power DC–DC converters, which have relevant applications in fuel cell-supplied systems, renewable energies, electric vehicles, and solid-state transformers.

A Numerical Simulation of Poroelastic Cylinder Decompression Problem on CUDA in an Axisymmetric Domain

2021 ◽  
Author(s):  
Salavat Ishbulatov ◽  
Viktoriya Yarushina ◽  
Yury Podladchikov
Keyword(s):  
Laboratory Experiments ◽  
High Efficiency ◽  
Core Sample ◽  
Analytical Models ◽  
Core Material ◽  
Physical Domain ◽  
Dynamic Boundary Conditions ◽  
The Core ◽  
Homogeneous Media ◽  
Fully Coupled

<p>The reliability of geomechanical and petrophysical laboratory experiments depends on coring operation. One of the steps where the core material undergoes critical loads is decompression during the core retrieval operation. Currently, a few numerical and analytical models simulate that process only with critical simplifications. The analytical solution considers only homogeneous media that neglects micro defects. FEM methods calculate slower than FDM up to several orders, simulating lifting processes with dynamic boundary conditions.</p><p>We present an axisymmetric cylindrical model of fully coupled fluid flow and elastic deformation solution by pseudo-transient numerical method. Calculation in the physical domain allows for high efficiency of parallelization on GPU, making it possible to simulate with high resolution of loading a core sample.</p>

Experiments and Analytical Simulation Work on an Innovative Steam-Injector-Driven Passive Core Injection Cooling System

2007 ◽  
Author(s):  
Michitsugu Mori ◽  
Tadashi Narabayashi ◽  
Shuichi Ohmori ◽  
Fumitoshi Watanabe
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
High Pressures ◽  
Cooling System ◽  
Steam Pressure ◽  
Injection System ◽  
The Core ◽  
Coolant Injection ◽  
Core Cooling

A Steam Injector (SI) is a simple, compact, passive pump which also functions as a high-performance direct-contact compact heater. We are developing this innovative concept by applying the SI system to core injection systems in Emergency Core Cooling Systems (ECCS) to further improve the safety of nuclear power plants. Passive ECCS in nuclear power plants would be inherently very safe and would prevent serious accidents by keeping the core covered with water (Severe Accident-Free Concept). The Passive Core Coolant Injection System driven by a high-efficiency SI is one that, in an accident such as a loss of coolant accident (LOCA), attains a higher discharge pressure than the supply steam pressure used to inject water into the reactor by operating the SI using water stored in the pool as the water supply source and steam contained in the reactor as the source of pressurization energy. The passive SI equipment would replace large, rotating machines such as pumps and motors, so eliminating the possibility of such equipment failing. In this Si-driven Passive Core Coolant Injection System (SI-PCIS), redundancy will be provided to ensure that the water and steam supply valves to the SI open reliably, and when the valves open, the SI will automatically start to inject water into the core to keep the core covered with water. The SI used in SI-PCIS works for a range of steam pressure conditions, from atmosphere pressure through to high pressures, as confirmed by analytical simulations which were done based on comprehensive experimental data obtained using reduced scale SI. We did further simulations and evaluations of the core cooling and coolant injection performance of SI-PCIS in BWR using RETRAN-3D code, developed using EPRI and other utilities, for the case of small LOCA. Reactors equipped with passive safety systems — the gravity-driven core cooling/injection system (GDCS) and depressurization valves (DPV) — would inevitably end up having large LOCA, even if they are initially small LOCA, as depressurization valves are forcibly opened in order to inject coolant from the GDCS pool to the GDCS water head at up to ∼0.2MPa. On the other hand, our simulation demonstrated that SI-PCIS could prevent large LOCA occurring in reactors by having by coolant discharged into the core in the event of small LOCA or when DPV unexpectedly open.

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.

scholarly journals Methods and Model Development for Coupled RELAP5/PARCS Analysis of the Atucha-II Nuclear Power Plant

2011 ◽  
Vol 2011 ◽  
pp. 1-16
Author(s):  
Andrew M. Ward ◽  
Benjamin S. Collins ◽  
Marcelo Madariaga ◽  
Yunlin Xu ◽  
Thomas J. Downar
Keyword(s):  
Steady State ◽  
Cross Sections ◽  
Nuclear Power ◽  
Model Development ◽  
Lattice Models ◽  
Transient Behavior ◽  
Injection System ◽  
The Core ◽  
Control Rods ◽  
Good Agreement

In order to analyze the steady state and transient behavior of CNA-II, several tasks were required. Methods and models were developed in several areas. HELIOS lattice models were developed and benchmarked against WIMS/MCNP5 results generated by NA-SA. Cross-sections for the coupled RELAP5/PARCS calculation were extracted from HELIOS within the GenPMAXS framework. The validation of both HELIOS and PARCS was performed primarily by comparisons to WIMS/PUMA and MCNP for idealized models. Special methods were developed to model the control rods and boron injection systems of CNA-II. The insertion of the rods is oblique, and a special routine was added to PARCS to treat this effect. CFD results combined with specialized mapping routines were used to model the boron injection system. In all cases there was good agreement in the results which provided confidence in the neutronics methods and modeling. A coupled code benchmark between U of M and U of Pisa is ongoing and results are still preliminary. Under a LOCA transient, the best estimate behavior of the core appears to be acceptable.

scholarly journals Post-test Analyses of the CMMR-4 Test

2021 ◽  
Author(s):  
Takuya Yamashita ◽  
Hiroshi Madokoro ◽  
Ikken Sato
Keyword(s):  
Nuclear Power ◽  
Gas Permeability ◽  
Plasma Heating ◽  
Heating System ◽  
Boiling Water Reactors ◽  
Core Material ◽  
The Core ◽  
Post Test ◽  
Ceramic Fuel ◽  
Water Reactors

Abstract Understanding the final distribution of core materials and their characteristics is important for decommissioning the Fukushima Daiichi Nuclear Power Station (1F). Such characteristics depend on the accident progression in each unit. However, BWR accident progression involves great uncertainty. This uncertainty, which was clarified by MAAP-MELCOR Crosswalk, cannot be resolved with existing knowledge and was thus addressed in this work through core material melting and relocation (CMMR) tests. For the test bundle, ZrO2 pellets were installed instead of UO2 pellets. A plasma heating system was used for the tests. In the CMMR-4 test, useful information was obtained on the core state just before slumping. The presence of macroscopic gas permeability of the core approaching ceramic fuel melting was confirmed, and the fuel columns remained standing, suggesting that the collapse of fuel columns, which is likely in the reactor condition, would not allow effective relocation of the hottest fuel away from the bottom of the core. This information will help us comprehend core degradation in boiling water reactors, similar to those in 1F. In addition, useful information on abrasive water suspension jet (AWSJ) cutting for debris-containing boride was obtained in the process of dismantling the test bundle. When the mixing debris that contains oxide, metal, and boride material is cut, AWSJ may be repelled by the boride in the debris, which may cut unexpected parts, thus generating a large amount of waste in cutting the boride part in the targeted debris. This information will help the decommissioning of 1F.

Development of Technologies on Innovative-Simplified Nuclear Power Plant Using High-Efficiency Steam Injectors: Part 9—System Outline and Endurance Test of Low-Pressure Steam Injectors

2006 ◽  
Author(s):  
Shuichi Ohmori ◽  
Michitsugu Mori ◽  
Shoji Goto ◽  
Tadashi Narabayashi ◽  
Chikako Iwaki ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Direct Contact ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
High Efficiency ◽  
Low Pressure ◽  
Injection System ◽  
System A

A Steam Injector (SI) is a simple, compact and passive pump and also acts as a high-performance direct-contact compact heater. This provides SI with capability to serve also as a direct-contact feedwater heater that heats up feedwater by using extracted steam from the turbine. We are developing technology for “Innovative Simplified Nuclear Power Plants” in order to further improve the economy and safety of nuclear power plants. Our technology development aims to significantly simplify equipment and reduce physical quantities by applying “High-Efficiency SI”, which are applicable to a wide range of operation regimes beyond the performance and applicable range of existing SIs and enables unprecedented multistage and parallel operation, to the low-pressure feedwater heaters and Emergency Core Cooling Systems (ECCS) of nuclear power plants, as well as achieve high inherent safety to prevent severe accidents by keeping the core covered with water (a Severe Accident-Free Concept). The innovative-simplified nuclear power plant consists of a simplified feedwater heating system, a passive core injection system and a passive containment cooling system. This report describes the results of the endurance and performance tests of low-pressure SIs for feedwater heaters with Jet-deaerator and core injection system. A part of this report are fruits of research which is carried out by Tokyo Electric Power Company (TEPCO), Toshiba, and 7 Universities in Japan, funded from the Ministry of Economy, Trade and Industry (METI) of Japan as the national public research-funded program.

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