Development of New Organic Iodine Filter for FCVS

2020 ◽  
Author(s):  
Kazuo Tominaga ◽  
Sohei Fukui ◽  
Motoi Tanaka ◽  
Tomoharu Hashimoto ◽  
Ryuichi Tayama
Keyword(s):  
Nuclear Power ◽  
Gas Adsorption ◽  
Current System ◽  
High Heat ◽  
Small Scale ◽  
Alkaline Water ◽  
Organic Iodine ◽  
Silver Zeolite ◽  
Water Reactors ◽  
Inorganic Iodine

Abstract All of the boiling water reactors in Japan were required to install filtered containment venting system (FCVS) for restart after the accident of Fukushima Daiichi Nuclear Power Plant. FCVS is composed of alkaline water and a metal fiber filter (MFF). Alkaline water has the function of reducing aerosol and inorganic iodine (I2). MFF is constituted for the purpose of removing the aerosol which was not able to be removed with alkaline water. With the above system, the released aerosol can be removed with DF1000 and inorganic iodine DF100. On the other hand, organic iodine that cannot be removed by alkaline water and MFF is removed by silver zeolite added downstream of the FCVS because of its high scattering property. Silver zeolite has particle of silver in the pores of the zeolite. Organic iodine is removed by chemical bonding with the silver. In the current system, the silver zeolite is DF50. Hitachi has developed the organic iodine removal system focusing on removing liquid chemicals for the purpose of improving the performance andmaintain ability, and reducing the price. In this presentation, we report the ionic liquid (IL) that has high heat resistance, high radiation resistance and high gas adsorption among liquid chemicals. In the small-scale tests, we found that IL has a higher performance than DF250, and improved maintainability by liquid property that is able to discharge with scrubber water at the same time, and have got the prospect of reducing manufacturing costs.

Recirculation Pump Speed Control System Upgrades in Nuclear Boiling Water Reactors

2008 ◽  
Author(s):  
James W. Morgan
Keyword(s):  
Control System ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Flow Control ◽  
Nuclear Power ◽  
Boiling Water ◽  
Boiling Water Reactors ◽  
Pump Speed ◽  
Water Reactors ◽  
Generator Sets

The nuclear power industry is faced with determining what to do with equipment and instrumentation reaching obsolescence and selecting the appropriate approach for upgrading the affected equipment. One of the systems in a nuclear power plant that has been a source of poor reliability in terms of replacement parts and control performance is the reactor recirculation pump speed/ flow control system for boiling water reactors (BWR). All of the operating BWR-3 and BWR-4’s use motor-generator sets, with a fluid coupled speed changer, to control the speed of the recirculation water pumps over the entire speed range of the pumps. These systems historically have had high maintenance costs, relative low efficiency, and relatively inaccurate speed control creating unwanted unit de-rates. BWR-5 and BWR-6 recirculation flow control schemes, which use flow control valves in conjunction with two-speed pumps, are also subject to upgrades for improved performance and reliability. These systems can be improved by installing solid-state adjustable speed drives (ASD), also known as variable frequency drives (VFD), in place of the motor-generator sets and the flow control valves. Several system configurations and ASD designs have been considered for optimal reliability and return on investment. This paper will discuss a highly reliable system and ASD design that is being developed for nuclear power plant reactor recirculation water pump controls. Design considerations discussed include ASD topology, controls architecture, accident, transient and hydraulic analyses, potential reactor internals modifications, installation, demolition and economic benefits.

Pressure Distributions in the Tip Clearance Region of an Unshrouded Axial Turbine as Affecting the Problem of Tip Burnout

1987 ◽  
Author(s):  
Jeffery P. Bindon
Keyword(s):  
Separation Bubble ◽  
Low Pressure ◽  
High Heat ◽  
Tip Clearance ◽  
Small Scale ◽  
Narrow Strip ◽  
Pressure Surface ◽  
High Heat Transfer ◽  
Axial Turbines ◽  
Secondary Vortices

The pressure distribution in the tip clearance region of a 2D turbine cascade was examined with reference to unknown factors which cause high heat transfer rates and burnout along the edge of the pressure surface of unshrouded cooled axial turbines. Using a special micro-tapping technique, the pressure along a very narrow strip of the blade edge was found to be 2.8 times lower than the cascade outlet pressure. This low pressure, coupled with a thin boundary layer due to the intense acceleration at gap entry, are believed to cause blade burnout. The flow phenomena causing the low pressure are of very small scale and do not appear to have been previously reported. The ultra low pressure is primarily caused by the sharp flow curvature demanded of the leakage flow at gap entry. The curvature is made more severe by the apparent attachement of the flow around the corner instead of immediately separating to increase the radius demanded of the flow. The low pressures are intensified by a depression in the suction corner and by the formation of a separation bubble in the clearance gap. The bubble creates a venturi action. The suction corner depression is due to the mainstream flow moving round the leakage and secondary vortices.

14C Distribution in Atmospheric and Aquatic Environments Around Qinshan Nuclear Power Plant, China

Radiocarbon ◽  
2014 ◽  
Vol 56 (3) ◽  
pp. 1107-1114 ◽  
Author(s):  
Zhongtang Wang ◽  
Dan Hu ◽  
Hong Xu ◽  
Qiuju Guo
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Water Samples ◽  
Nuclear Power ◽  
Dissolved Inorganic Carbon ◽  
Tap Water ◽  
Surface Seawater ◽  
Pressurized Water ◽  
Specific Activities ◽  
Water Reactors

Atmospheric CO2 and aquatic water samples were analyzed to evaluate the environmental 14C enrichment due to operation of the Qinshan nuclear power plant (NPP), where two heavy-water reactors and five pressurized-water reactors are employed. Elevated 14C-specific activities (2–26.7 Bq/kg C) were observed in the short-term air samples collected within a 5-km radius, while samples over 5 km were close to background levels. The 14C-specific activities of dissolved inorganic carbon (DIC) in the surface seawater samples ranged from 196.8 to 206.5 Bq/kg C (average 203.4 Bq/kg C), which are close to the background value. No elevated 14C level in surface seawater was found after 20 years of operation of Qinshan NPP, indicating that the 14C discharged was well diffused. The results of the freshwater samples show that excess 14C-specific activity (average 17.1 Bq/kg C) was found in surface water and well water samples, while no obvious 14C increase was found in drinking water (groundwater and tap water) compared to the background level.

Finite Element Analyses of Residual Stresses in Typical Welded Joints Used in Nuclear Power Plants and Comparisons With Experiments

2010 ◽  
Author(s):  
Dean Deng ◽  
Kazuo Ogawa ◽  
Nobuyoshi Yanagida ◽  
Koichi Saito
Keyword(s):  
Numerical Simulation ◽  
Residual Stress ◽  
Residual Stresses ◽  
Welded Joints ◽  
Nuclear Power ◽  
Power Plants ◽  
Butt Joint ◽  
Welding Residual Stress ◽  
Dissimilar Metal ◽  
Water Reactors

Recent discoveries of stress corrosion cracking (SCC) at nickel-based metals in pressurized water reactors (PWRs) and boiling water reactors (BWRs) have raised concerns about safety and integrity of plant components. It has been recognized that welding residual stress is an important factor causing the issue of SCC in a weldment. In this study, both numerical simulation technology and experimental method were employed to investigate the characteristics of welding residual stress distribution in several typical welded joints, which are used in nuclear power plants. These joints include a thick plate butt-welded Alloy 600 joint, a dissimilar metal J-groove set-in joint and a dissimilar metal girth-butt joint. First of all, numerical simulation technology was used to predict welding residual stresses in these three joints, and the influence of heat source model on welding residual stress was examined. Meanwhile, the influence of other thermal processes such as cladding, buttering and heat treatment on the final residual stresses in the dissimilar metal girth-butt joint was also clarified. Secondly, we also measured the residual stresses in three corresponding mock-ups. Finally, the comparisons of the simulation results and the measured data have shed light on how to effectively simulate welding residual stress in these typical joints.

scholarly journals 14C Release in Various Chemical Forms With Gaseous Effluents From the Paks Nuclear Power Plant

Radiocarbon ◽  
1989 ◽  
Vol 31 (03) ◽  
pp. 754-761 ◽  
Author(s):  
Ede Hertelendi ◽  
György Uchrin ◽  
Peter Ormai
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Liquid Scintillation ◽  
Mean Value ◽  
Scintillation Counting ◽  
Pressurized Water ◽  
The Mean ◽  
Water Reactors

We present results of airborne 14C emission measurements from the Paks PWR nuclear power plant. Long-term release of 14C in the form of carbon dioxide or carbon monoxide and hydrocarbons were simultaneously measured. The results of internal gas-proportional and liquid scintillation counting agree well with theoretical assessments of 14C releases from pressurized water reactors. The mean value of the 14C concentration in discharged air is 130Bqm-3 and the normalized release is equal to 740GBq/GWe · yr. > 95% of 14C released is in the form of hydrocarbons, ca 4% is apportioned to CO2, and <1% to CO. Tree-ring measurements were also made and indicated a minute increase of 14C content in the vicinity of the nuclear power plant.

PM-HIP Research for Pressure Retaining Applications Within the Electric Power Industry

2014 ◽  
Author(s):  
David Gandy ◽  
John Siefert ◽  
Lou Lherbier ◽  
David Novotnak
Keyword(s):  
Electric Power ◽  
Nuclear Power ◽  
Hot Isostatic Pressing ◽  
Electric Power Industry ◽  
Power Industry ◽  
Plate Rolling ◽  
Alloy 690 ◽  
Class 1 ◽  
Water Reactors ◽  
Grade 91

For more than 60 years now, the nuclear power industry has relied on structural and pressure retaining materials generated via established manufacturing practices such as casting, plate rolling-and-welding, forging, drawing, and/or extrusion. During the past three years, EPRI has been leading the development and introduction of another established process, powder metallurgy and hot isostatic pressing (PM/HIP), for pressure retaining applications in the electric power industry. The research includes assessment of two primary alloys: 316L stainless steel and Grade 91 creep-strength enhanced ferritic steels, for introduction into the ASME Boiler and Pressure Vessel Code. Continuing DOE and EPRI research on other structural/pressure retaining alloys such as Alloy 690, SA 508 Class 1, Alloy 625, hard-facing materials, and others are also underway. This research will have a tremendous impact as we move forward over the next few decades on the selection of new alloys and components for advanced light water reactors and small modular reactors. Furthermore, fabrication of high alloy materials/components may require the use of new manufacturing processes to achieve acceptable properties for higher temperature applications such as those in Generation IV applications. Current research by EPRI and DOE will be reviewed and emphasis will be targeted at advanced applications where PM/HIP may be applied in the future.

scholarly journals On the possibility to improve mixed uranium-plutonium fuel in fast reactors

2020 ◽  
Vol 6 (2) ◽  
pp. 131-135
Author(s):  
Vladimir A. Eliseev ◽  
Dmitry A. Klinov ◽  
Noël Camarcat ◽  
David Lemasson ◽  
Clement Mériot ◽  
...  
Keyword(s):  
Isotopic Composition ◽  
Nuclear Power ◽  
Spent Nuclear Fuel ◽  
Light Water Reactors ◽  
Fast Reactors ◽  
Light Water ◽  
Fuel Cycles ◽  
Mox Fuel ◽  
Water Reactors ◽  
Operational Activities

Accumulation of plutonium extracted from the spent nuclear fuel (SNF) of light water reactors is one of the central problems in nuclear power. To reduce out-of-the-reactor Pu inventory, leading nuclear power countries (France, Japan) use plutonium in light water power reactors in the form of MOX fuel, with half of Pu fissioning in this fuel. The rest of Pu cannot be reused easily and efficiently in light water reactors because of the high content of even isotopes. Plutonium for which there are no potential consumers is accumulated. Unlike thermal reactors, fast reactors take plutonium of any isotopic composition. That makes it possible to improve plutonium isotopic composition and to reduce the fraction of even isotopes to the level that allows reuse of such plutonium in thermal reactors. The idea of changing the isotopic composition of Pu in fast reactors is well-known. The originality of the research lies in applying this idea to combine the fuel cycles of fast and thermal reactors. Pu isotopic composition can be improved by combining certain operational activities in order to supply fuel to thermal and fast reactors. Scientific and technological justification of the possibility will let Russian BN technologies and French MOX fuel technologies work in synergy with thermal reactors.

scholarly journals Reactor Vessel Internals Segmentation Experience using Mechanical Cutting Tools

2013 ◽  
Vol 10 (2) ◽  
pp. 6-10 ◽  
Author(s):  
Petr Pospíšil
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Cutting Tools ◽  
Reactor Vessel ◽  
Boiling Water Reactors ◽  
Abrasive Waterjet ◽  
Pressurized Water ◽  
Water Reactors ◽  
Mechanical Cutting ◽  
Reactor Internals

Abstract Some commercial nuclear power plants have been permanently shut down to date and decommissioned using dismantling methods. Other operating plants have decided to undergo an upgrade process that includes replacement of reactor internals. In both cases, there is a need to perform a segmentation of the reactor vessel internals with proven methods for long term waste disposal. Westinghouse has developed several concepts to dismantle reactor internals based on safe and reliable techniques, including plasma arc cutting (PAC), abrasive waterjet cutting (AWJC), metal disintegration machining (MDM), or mechanical cutting. Mechanical cutting has been used by Westinghouse since 1999 for both Pressurized Water Reactors (PWR’s) and Boiling Water Reactors (BWR’s) and its process has been continuously improved over the years. The complexity of the work requires well designed and reliable tools. Different band saws, disc saws, tube cutters and shearing tools have been developed to cut the reactor internals. All of those equipments are hydraulically driven which is very suitable for submerged applications. Westinghouse experience in mechanical cutting has demonstrated that it is an excellent technique for segmentation of internals. In summary, the purpose of this paper will be to provide an overview of the Westinghouse mechanical segmentation process, based on actual experience from the work that has been completed to date.

Comparison of US and UK Practices and Philosophies for Nuclear Plant Control Room Design

1982 ◽  
Vol 26 (7) ◽  
pp. 659-663
Author(s):  
Les Ainsworth
Keyword(s):  
Nuclear Power ◽  
Plant Design ◽  
Control Room ◽  
Pressurized Water ◽  
Basic Task ◽  
Room Design ◽  
Power Plant Control ◽  
Gas Cooling ◽  
Water Reactors ◽  
Plant Control

The American and British nuclear power programmes have in the past taken divergent routes, with the Americans choosing pressurized water reactors, whilst the British have opted for gas cooling. Although the technology and plant design for these two systems encompasses many fundamental differences, some of which have ramifications for the controllers, the basic task of monitoring and controlling a reactor holds many similarities in both countries. It is therefore instructive to compare and contrast the approaches which have been taken to human factors in nuclear power plant control room design on both sides of the Atlantic.

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